ML062900323

10/2006 - Final Outlines

ML062900323
Person / Time
Site:	Cooper
Issue date:	10/02/2006
From:	- No Known Affiliation
To:	Office of Nuclear Reactor Regulation
References
ES-401, Rev 9, ES-401-1
Download: ML062900323 (30)

Text

ES-401, Rev. 9 BWR Examination Outline Form ES-401-1 Facility: Date of Exam:

RO K/A Category Points SRO-Only Points Tier Group K K K K K K A A A A G A2 G* Total 1 2 3 4 5 6 1 2 3 4

• Total

1. 1 3 4 3 3 4 3 20 4 3 7 Emergency &

Abnormal 2 1 1 1 N/A 1 2 N/A 1 7 1 2 3 Plant Evolutions Tier Totals 4 5 4 4 6 4 27 5 5 10 1 3 2 2 2 2 3 2 2 2 3 3 26 3 2 5 2.

Plant 2 1 1 1 1 1 1 1 1 1 2 1 12 1 2 3 Systems Tier Totals 4 3 3 3 3 4 3 3 3 5 4 38 4 4 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 K/A 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 ATier Totals@ in each K/A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by "1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the 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 ES-401, Attachment 2, for guidance regarding the elimination of inappropriate K/A 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 shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A Catalog, but the topics must be relevant to the applicable evolution or system.

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics= importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals

(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401, Rev. 9 2 Form ES-401-1 ES-401 BWR Examination OutlineForm ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced X AA2.01 Ability to determine and/or interpret the 3.5 1 Core Flow Circulation / 1 & 4 following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :Power/flow map (CFR: 41.10 / 43.5

/ 45.13) 295003 Partial or Complete Loss of AC / 6 X 2.1.14 Knowledge of system status criteria which 2.5 2 require the notification of plant personnel. (CFR:

43.5 / 45.12) 295004 Partial or Total Loss of DC Pwr / 6 X AA2.02 Ability to determine and/or interpret the 3.5 3 following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :Extent of partial or complete loss of D.C. power (CFR: 41.10 / 43.5 /

45.13) 2.9 295005 Main Turbine Generator Trip / 3 X AK2.02 Knowledge of the interrelationships between 4 the MAIN TURBINE GENERATOR TRIP and the following: Feedwater temperature (CFR: 41.7 to 45.8) 295006 SCRAM / 1 X AA1.06 Ability to operate and/or monitor the 3.5 5 following as they apply to SCRAM : CRD hydraulic system. (CFR: 41.7 / 45.6) 295016 Control Room Abandonment / 7 X AK2.01 Knowledge of the interrelations between 4.4 6 CONTROL ROOM ABANDONMENT and the following: Remote shutdown panel: Plant-Specific (CFR: 41.7 / 45.8) 295018 Partial or Total Loss of CCW / 8 X AK1.01 Knowledge of the operational implications of 3.5 7 the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Effects on component/system operations.

(CFR: 41.8 to 41.10) 295019 Partial or Total Loss of Inst. Air / 8 X AK3.03 Knowledge of the reasons for the following 3.2 8 responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Service air isolations: Plant-Specific (CFR: 41.5 / 45.6) 295021 Loss of Shutdown Cooling / 4 X AK1.02 Knowledge of the operational implications of 3.6 9 the following concepts as they apply to LOSS OF SHUTDOWN COOLING : (CFR: 41.8 to 41.10)

Thermal Stratification 295023 Refueling Acc / 8 X AK3.02 Knowledge of the reasons for the following 3.4 10 responses as they apply to REFUELING ACCIDENTS : Interlocks associated with fuel handling equipment.(CFR: 41.5 / 45.6) 295024 High Drywell Pressure / 5 X EK2.18 Knowledge of the interrelations between 3.3 11 HIGH DRYWELL PRESSURE and the following:

Ventilation (CFR: 41.7 / 45.8) 295025 High Reactor Pressure / 3 X EA1.05 Ability to operate and/or monitor the 3.7 12 following as they apply to HIGH REACTOR PRESSURE: RCIC: Plant-Specific (CFR: 41.7 /

45.6) 295026 Suppression Pool High Water Temp. X EK1.01 Knowledge of the operational implications of 3.0 13

/5 the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE : (CFR: 41.8 to 41.10) Pump NPSH 295027 High Containment Temperature / 5 Note 1 295028 High Drywell Temperature / 5 X EA1.04 Ability to operate and/or monitor the 3.9 14 following as they apply to HIGH DRYWELL TEMPERATURE : Drywell pressure (CFR: 41.7 /

45.6) 295030 Low Suppression Pool Wtr Lvl / 5 X EK3.06 Knowledge of the reasons for the following 3.6 15 responses as they apply to LOW SUPPRESSION POOL WATER LEVEL: Reactor SCRAM (CFR:

41.5 / 45.6) 295030 Low Suppression Pool Wtr Lvl / 5 X G2.1.33 Ability to recognize indications for system 3.4 16 operating parameters which are entry-level conditions for technical specifications. (CFR: 43.2 /

43.3 / 45.3) 295031 Reactor Low Water Level / 2 X G2.4.4 Ability to recognize abnormal indications for 4.0 17 system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 295037 SCRAM Condition Present X EA2.07 Ability to determine and/or interpret the 4.0 18 and Power Above APRM Downscale following as they apply to SCRAM CONDITION or Unknown / 1 PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Containment conditions/isolations (CFR: 41.10 / 43.5 / 45.13) 295038 High Off-site Release Rate / 9 X EA2.03 Ability to determine and/or interpret the 3.5 19 following as they apply to HIGH OFF-SITE RELEASE RATE : Radiation levels (CFR: 41.10 /

43.5 / 45.13) 600000 Plant Fire On Site / 8 X AK2.01Knowledge of the interrelations between 2.6 20 PLANT FIRE ON SITE and the following: AK2.01 Sensors / detectors and valves K/A Category Totals: 3 4 3 3 4 3 Group Point Total: 20

ES-401, Rev. 9 2 Form ES-401-1 ES-401 BWR Examination OutlineForm ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 295001 Partial or Complete Loss of Forced X AA2.03 Ability to determine and/or interpret the 3.3 S76 Core Flow Circulation / 1 & 4 following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :

Actual core flow (CFR: 41.10 / 43.5 / 45.13) 295003 Partial or Complete Loss of AC / 6 X G2.1.33 Ability to recognize indications for system 4.0 S77 operating parameters which are entry-level conditions for technical specifications. (CFR: 43.2 /

43.3 / 45.3) 295004 Partial or Total Loss of DC Pwr / 6 295005 Main Turbine Generator Trip / 3 2.4.30 Knowledge of which events related to system 295006 SCRAM / 1 X operations/status should bereported to outside 3.6 S79 agencies. (CFR: 43.5 / 45.11) 295016 Control Room Abandonment / 7 AA2.01 Ability to determine and/or interpret the 295018 Partial or Total Loss of CCW / 8 X following as they apply to PARTIAL OR COMPLETE 3.4 S78 LOSS OF COMPONENT COOLING WATER :

Component temperatures (CFR: 41.10 / 43.5 /

45.13) 295019 Partial or Total Loss of Inst. Air / 8 295021 Loss of Shutdown Cooling / 4 295023 Refueling Acc / 8 295024 High Drywell Pressure / 5 295025 High Reactor Pressure / 3 295026 Suppression Pool High Water Temp. EA2.01 Ability to determine and/or interpret the

/5 X following as they apply to SUPPRESSION POOL 4.2 S80 HIGH WATER TEMPERATURE: Suppression pool water temperature (CFR: 41.10 / 43.5 / 45.13) 295027 High Containment Temperature / 5 Note 1 295028 High Drywell Temperature / 5 295030 Low Suppression Pool Wtr Lvl / 5 295031 Reactor Low Water Level / 2 295037 SCRAM Condition Present X G2.4.4 Ability to recognize abnormal indications for 4.3 S81 and Power Above APRM Downscale system operating parameters which are entry-level or Unknown / 1 conditions for emergency and abnormal operating procedures. (CFR: 41.10 / 43.2 / 45.6) 295038 High Off-site Release Rate / 9 X EA2.03 Ability to determine and/or interpret the 4.3 S82 following as they apply to HIGH OFF-SITE RELEASE RATE : HRadiation levels (CFR: 41.10 /

43.5 / 45.13) 600000 Plant Fire On Site / 8 K/A Category Totals: 4 3 Group Point Total: 7

ES-401, Rev. 9 3 Form ES-401-1 ES-401 BWR Examination OutlineForm ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 X AK1.02 Knowledge of the operational implications of the 3. 21 following concepts as they apply to LOW REACTOR 0 WATER LEVEL : Recirculation pump net positive suction head: Plant-Specific (CFR: 41.8 to 41.10) 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 Note 1 AA2.02 Ability to determine and/or interpret the following 295012 High Drywell Temperature / 5 X as they apply to HIGH DRYWELL TEMPERATURE : 3. 22 Drywell pressure (CFR: 41.10 / 43.5 / 45.13) 9 295013 High Suppression Pool Temp. /

5 AK3.01 Knowledge of the reasons for the following 295014 Inadvertent Reactivity Addition / X responses as they apply to INADVERTENT REACTIVITY 4. 23 1 ADDITION: Reactor SCRAM (CFR: 41.5 / 45.6) 1 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 &

7 295022 Loss of CRD Pumps / 1 X G2.1.30 Ability to locate and operate components / 3. 24 including local controls. (CFR: 41.7 / 45.7) 9 295029 High Suppression Pool Wtr Lvl /

5 295032 High Secondary Containment X EK2.06 Knowledge of the interrelations between HIGH 3. 25 Area Temperature / 5 SECONDARY CONTAINMENT AREA TEMPERATURE 3 and the following: Area temperature monitoring system (CFR: 41.7 / 45.8) 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation High Radiation / 9 EA1.02 Ability to operate and/or monitor the following as 295035 Secondary Containment High X they apply to SECONDARY CONTAINMENT HIGH 3. 26 Differential Pressure / 5 DIFFERENTIAL PRESSURE: SBGT/FRVS (CFR: 41.7 / 8 45.6)

EA2.03 Ability to determine and/or interpret the following 295036 Secondary Containment High X as they apply to Secondary Containment High Sump/Area 3. 27 Sump/Area Water Level / 5 Water Level. Cause of the high water level. ((CFR: 4 41.10/43.5/45.13) 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 1 1 1 1 2 1 Group Point Total: 7

ES-401, Rev. 9 3 Form ES-401-1 ES-401 BWR Examination OutlineForm ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 ( SRO)

E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR #

1 2 3 1 2 295002 Loss of Main Condenser Vac / 3 295007 High Reactor Pressure / 3 295008 High Reactor Water Level / 2 295009 Low Reactor Water Level / 2 295010 High Drywell Pressure / 5 295011 High Containment Temp / 5 Note 1 295012 High Drywell Temperature / 5 295013 High Suppression Pool Temp. / 5 295014 Inadvertent Reactivity Addition / 1 X AA2.03 Ability to determine and/or interpret the 4. S83 following as they apply to INADVERTENT 3 REACTIVITY ADDITION : Cause of reactivity addition (CFR: 41.10 / 43.5 / 45.13) 295015 Incomplete SCRAM / 1 295017 High Off-site Release Rate / 9 295020 Inadvertent Cont. Isolation / 5 & 7 295022 Loss of CRD Pumps / 1 295029 High Suppression Pool Wtr Lvl / 5 295032 High Secondary Containment Area Temperature / 5 295033 High Secondary Containment Area Radiation Levels / 9 295034 Secondary Containment Ventilation X 295034.G2.2.25 Knowledge of bases in technical 3. S84 High Radiation / 9 specifications for limiting conditions for operations and 7 safety limits. (CFR: 43.2) 295035 Secondary Containment High X 295035.G2.4.4 Ability to recognize abnormal 4. S85 Differential Pressure / 5 indications for system operating parameters which are 3 entry-level conditions for emergency and abnormal operating procedures.

295036 Secondary Containment High Sump/Area Water Level / 5 500000 High CTMT Hydrogen Conc. / 5 K/A Category Point Totals: 1 2 Group Point Total: 3

ES-401, Rev. 9 4 Form ES-401-1 ES-401BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (RO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection X K6.04 Knowledge of the effect that a loss or 3.3 28 Mode malfunction of the following will have on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) : Keep Fill System(CFR: 41.7 /

45.7) 205000 Shutdown Cooling X K6.04 Knowledge of the effect that a loss or 3.6 29 malfunction of the following will have on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) : Reactor Water Level (CFR: 41.7 / 45.7) 206000 HPCI X G2.2.22 Knowledge of limiting conditions for 3.4 30 operations and safety limits. (CFR: 43.2 /

45.2) 207000 Isolation (Emergency) Note 2 Condenser A2.02 Ability to (a) predict the impacts of 209001 LPCS X the following on the LOW PRESSURE 3.2 31 CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures (CFR: 41.5 / 45.6) 209002 HPCS Note 3 211000 SLC X A3.06 Ability to monitor automatic operations 4.0 32 of the STANDBY LIQUID CONTROL SYSTEM including: RWCU: Plant-Specific (CFR: 41.7 / 45.7) 211000 SLC X K2.01 Knowledge of electrical power supplies 2.9 33 to the following: SBLC pumps (CFR: 41.7)

K5.02 Knowledge of the operational 212000 RPS X implications of the following concepts as they 3.3 34 apply to REACTOR PROTECTION SYSTEM :

Specific logic arrangements (CFR: 41.5 /

45.3)

A4.04 Ability to manually operate and/or 215003 IRM X monitor in the control room: IRM back 3.7 35 panel switches, meters, and indicating lights (CFR: 41.7 / 45.5 to 45.8) 215004 Source Range Monitor X K1.06 Knowledge of the physical connections 2.8 36 and/or cause effect relationships between SOURCE RANGE MONITOR (SRM)

SYSTEM and the following: Reactor vessel (CFR: 41.2 to 41.9 / 45.7 to 45.8) 215005 APRM / LPRM X A1.03 Ability to predict and/or monitor 3.6 37 changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including:

Control rod block status (CFR: 41.5 / 45.5)

217000 RCIC X G2.4.31 Knowledge of annunciators alarms 3.3 38 and indications / and use of the response instructions.(CFR: 41.10 / 45.3) 218000 ADS X K4.04 Knowledge of AUTOMATIC 3.5 39 DEPRESSURIZATION SYSTEM designfeature(s) and/or interlocks which provide for the following: Insures adequate air supply to ADS valves: Plant-Specific (CFR:

41.7) 218000 ADS X K2.01Knowledge of electrical power supplies 3.1 40 to the following: ADS logic (CFR: 41.7) 223002 PCIS/Nuclear Steam X A4.03 Ability to manually operate and/or 3.6 41 Supply Shutoff monitor in the control room: Reset system isolations (CFR: 41.7 / 45.5 to 45.8) 223002 PCIS/Nuclear Steam X K1.19 Knowledge of the physical connections 2.7 42 Supply Shutoff and/or cause effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following: Component cooling water systems (CFR: 41.2 to 41.9 / 45.7 to 45.8) 239002 SRVs X A1.04 Ability to predict and/or monitor 3.8 43 changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including: Reactor pressure (CFR:

41.5 / 45.5) 259002 Reactor Water Level X G2.4.49 Ability to perform without reference to 4.0 44 Control procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 / 43.2 / 45.6) 261000 SGTS X K3.01 Knowledge of the effect that a loss or 3.3 45 malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following:

Secondary containment and environment differential pressure (CFR: 41.7 /45.6) 261000 SGTS X K1.08 Knowledge of the physical connections 2.8 46 and/or causeeffect relationships between STANDBY GAS TREATMENT SYSTEM and the following: Process radiation monitoring system (CFR: 41.2 to 41.9 / 45.7 to 45.8) 262001 AC Electrical X K5.01 Knowledge of the operational 3.1 47 Distribution implications of the following concepts as they apply to A.C. ELECTRICAL DISTRIBUTION:

Principle involved with paralleling two A.C.

sources (CFR: 41.5 / 45.3) 262002 UPS (AC/DC) X K6.01 Knowledge of the effect that a loss or 2.7 48 malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) : A.C. electrical power (CFR: 41.7

/ 45.7) 263000 DC Electrical X A3.01 Ability to monitor automatic operations 3.2 49 Distribution of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating lights (CFR: 41.7 / 45.7) 263000 DC Electrical X K4.01 Knowledge of D.C. ELECTRICAL 3.1 50 Distribution DISTRIBUTION design feature(s) and/or interlocks which provide for the following: Manual/ automatic transfers of

control: Plant-Specific (CFR: 41.7) 264000 EDGs X K3.01 Knowledge of the effect that a loss or 4.2 51 malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: Emergency core cooling systems CFR: 41.7 / 45.4) 300000 Instrument Air X A4.01 Ability to manually operate and / or 2.6 52 monitor in the control room: Pressure gauges (CFR: 41.7 / 45.5 to 45.8) 400000 Component Cooling X A2.02 Ability to (a) predict the impacts of the 2.8 53 Water following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: High/low surge tank level (CFR: 41.5 / 45.6)

K/A Category Point Totals: 3 2 2 2 2 3 2 2 2 3 3 Group Point Total: 26

ES-401, Rev. 9 4 Form ES-401-1 ES-401BWR Examination Outline Form ES-401-1 Plant Systems - Tier 2/Group 1 (SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 203000 RHR/LPCI: Injection Mode 205000 Shutdown Cooling X G2.2.22 Knowledge of limiting conditions for 4.1 S86 operations and safety limits. (CFR: 43.2 /

45.2) 206000 HPCI 207000 Isolation (Emergency) Note 2 Condenser 209001 LPCS 209002 HPCS Note 3 211000 SLC X A2.05 Ability to (a) predict the impacts of the 3.4 S87 following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of SBLC tank heaters (CFR: 41.5 / 45.6) 212000 RPS X G2.2.25 Knowledge of bases in technical 2.9 S88 specifications for limiting conditions for operations and safety limits.

(CFR: 43.2) 215003 IRM 215004 Source Range Monitor X A2.03 Ability to (a) predict the impacts of the 3.3 S89 following on the SOURCE RANGE MONITOR (SRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck detector (CFR: 41.5 / 45.6) 215005 APRM / LPRM 217000 RCIC 218000 ADS 223002 PCIS/Nuclear Steam Supply Shutoff 239002 SRVs 259002 Reactor Water Level Control 261000 SGTS 262001 AC Electrical X 262001.A2.10 Ability to (a) predict the impacts 3.4 S90 Distribution of the following on the A.C. ELECTRICAL DISTRIBUTION ; and (b) based on those predictions, use procedures to correct,

control, or mitigate the consequences of those abnormal conditions or operations: Exceeding current limitations (CFR: 41.5 / 45.6) 262002 UPS (AC/DC) 263000 DC Electrical Distribution 264000 EDGs 300000 Instrument Air 400000 Component Cooling Water K/A Category Point Totals: 3 2 Group Point Total: 5

ES-401, Rev. 9 5 Form ES-401-1 ES-401BWR Examination OutlineForm ES-401-1 Plant Systems - Tier 2/Group 2 (RO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic X A3.10 Ability to monitor automatic 3.0 54 operations of the CONTROL ROD DRIVE HYDRAULIC SYSTEM including:

Lights and alarms (CFR: 41.7 / 45.7) 201002 RMCS X A1.03 Ability to predict and/or monitor 3.0 55 changes in parameters associated with operating the REACTOR MANUAL CONTROL SYSTEM controls including:

Rod movement sequence lights (CFR:

41.5 / 45.5) 201003 Control Rod and Drive X A4.02 Ability to manually operate and/or 3.5 56 Mechanism monitor in the control room: CRD mechanism position:

(CFR: 41.7 / 45.5 to 45.8) 201004 RSCS Note 4 201005 RCIS Note 5 201006 RWM 202001 Recirculation X A2.06 Ability to (a) predict the impacts of 3.6 57 the following on the RECIRCULATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Inadvertent recirculation flow decrease (CFR: 41.5 / 45.6) 202002 Recirculation Flow Control X K4.02 Knowledge of RECIRCULATION 3.0 58 FLOW CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Recirculation pump speed control: Plant-Specific (CFR: 41.7) 204000 RWCU X G2.4.49 Ability to perform without 4.0 59 reference to procedures those actions that require immediate operation of system components and controls. (CFR:

41.10 / 43.2 / 45.6) 214000 RPIS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst. X K3.16 Knowledge of the effect that a 3.0 60 loss or malfunction of the NUCLEAR BOILER Instrumentation will have on following: Main turbine (CFR: 41.7 / 45.4) 219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray

Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment X K1.09 Knowledge of the physical 2.8 61 connections and/or cause effect relationships between FUEL HANDLING EQUIPMENT and the following: Fuel pool ventilation: Plant-Specific (CFR:

41.2 to 41.9 / 45.7 to 45.8) 239001 Main and Reheat Steam Note 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate 259001 Reactor Feedwater X A4.04 Ability to manually operate and/or 3.1 63 monitor in the control room: System valves (CFR: 41.7 / 45.5 to 45.8) 268000 Radwaste 271000 Offgas X K5.08 Knowledge of the operational 2.5 64 implications of the following concepts as they apply to OFFGAS SYSTEM :

Charcoal absorption of fission product gases (CFR: 41.7 / 45.4) 272000 Radiation Monitoring 286000 Fire Protection X K2.03 Knowledge of electrical power 2.5 65 supplies to the following: Fire detection system: Plant-Specific (CFR: 41.7) 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC X K6.04 Knowledge of the effect that a loss 2.6 62 or malfunction of the following will have on the CONTROL ROOM HVAC : Fire protection (CFR: 41.7 / 45.7) 290002 Reactor Vessel Internals K/A Category Point Totals: 1 1 1 1 1 1 1 1 1 2 1 Group Point Total: 12

ES-401, Rev. 9 5 Form ES-401-1 ES-401BWR Examination OutlineForm ES-401-1 Plant Systems - Tier 2/Group 2 ( SRO)

System # / Name K K K K K K A A A A G K/A Topic(s) IR #

1 2 3 4 5 6 1 2 3 4 201001 CRD Hydraulic 201002 RMCS 201003 Control Rod and Drive Mechanism 201004 RSCS Note: 4 201005 RCIS Note: 5 201006 RWM 202001 Recirculation 202002 Recirculation Flow Control 204000 RWCU 214000 RPIS 215001 Traversing In-core Probe 215002 RBM 216000 Nuclear Boiler Inst.

219000 RHR/LPCI: Torus/Pool Cooling Mode 223001 Primary CTMT and Aux.

226001 RHR/LPCI: CTMT Spray Mode 230000 RHR/LPCI: Torus/Pool Spray Mode 233000 Fuel Pool Cooling/Cleanup 234000 Fuel Handling Equipment 239001 Main and Reheat Steam 239003 MSIV Leakage Control 241000 Reactor/Turbine Pressure Regulator 245000 Main Turbine Gen. / Aux.

256000 Reactor Condensate 259001 Reactor Feedwater 268000 Radwaste X G2.1.33 Ability to recognize indications for 3.3 S91 system operating parameters which are entry-level conditions for technical specifications.c(CFR: 43.2 / 43.3 / 45.3) 271000 Offgas 272000 Radiation Monitoring

286000 Fire Protection X G2.2.22 Knowledge of limiting conditions 4.0 S92 for operations and safety limits.

(CFR: 43.2 / 45.2) 288000 Plant Ventilation 290001 Secondary CTMT 290003 Control Room HVAC 290002 Reactor Vessel Internals X A2.05 Ability to (a) predict the impacts of 3.7 S93 the following on the REACTOR VESSEL INTERNALS ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

HExceeding Thermal Limits CFR: 41.5 /

45.6)

K/A Category Point Totals: 1 2 Group Point Total: 3

ES-401, Rev. 9 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Date of Exam:

Category K/A # Topic RO SRO-Only IR # IR #

2.1.32 Ability to explain and apply system limits and precautions. (CFR: 3.4 66 41.10 / 43.2 / 45.12) 1.

Conduct 2.1.30 Ability to locate and operate components / including local 3.9 67 of Operations controls. (CFR: 41.7 / 45.7) 2.1.16 Ability to operate plant phone / paging system / and two-way 2.9 68 radio. (CFR: 41.10 / 45.12) 2.1.34 Ability to maintain primary and secondary plant chemistry within 2.9 S94 allowable limits. (CFR: 41.10 / 43.5 / 45.12)

Subtotal 3 1 2.2.13 Knowledge of tagging and clearance procedures. (CFR: 41.10 / 3.6 69 45.13)

2. 2.2.1 2.2.1 Ability to perform pre-startup procedures for the facility / 3.7 70 Equipment including operating those controls associated with plant Control equipment that could affect reactivity. (CFR: 45.1) 2.2.18 Knowledge of the process for managing maintenance activities 3.6 S95 during shutdown operations. (CFR: 43.5 / 45.13) 2.2.32 Knowledge of the effects of alterations on core configuration. 3.3 S96 (CFR: 43.6)

Subtotal 2 2 2.3.11 Ability to control radiation releases. (CFR: 45.9 / 45.10) 2.7 71 2.3.4 Knowledge of radiation exposure limits and contamination 2.5 72

3. control / including permissible levels in excess of those Radiation Control authorized. (CFR: 43.4 / 45.10) 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control 3.0 S97 requirements. (CFR: 41.12 / 43.4. 45.9 / 45.10) 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are 2.9 S98 outside the control room (e.g. / waste disposal and handling systems). (CFR: 43.4 / 45.10)

Subtotal 2 2 2.4.39 Knowledge of the RO's responsibilities in emergency plan 3.3 73 implementation. (CFR: 45.11) 4.

Emergency 2.4.34 Knowledge of RO tasks performed outside the main control 3.8 74 Procedures / room Plan during emergency operations including system geography and system implications. (CFR: 43.5 / 45.13) 2.4.48 Ability to interpret control room indications to verify the status 3.5 75 and operation of system / and understand how operator action s and directives affect plant and system conditions. (CFR: 43.5 /

45.12) 2.4.11 Knowledge of abnormal condition procedures. (CFR: 41.10 / 3.6 S99 43.5 / 45.13) 2.4.3 Ability to identify post-accident instrumentation. (CFR: 41.6 / 3.8 S100 45.4)

Subtotal 3 2

Tier 3 Point Total 10 7 Note 1: Cooper Nuclear Station does not have a Mark III containment.

Note 2: Cooper Nuclear Station does not have an isolation condenser.

Note 3: Cooper Nuclear Station does not have a High Pressure Core Spray System (HPCS).

Note 4: Cooper Nuclear Station has abandoned the Rod Sequence Control System (RSCS).

Note 5: Cooper Nuclear Station does not have a Rod Control and Information System (RCIS).

Note 6: Cooper Nuclear Station does not have an automated MSIV leakage control system.

ES 301 Administrative Topics Outline Form ES 301-1 Facility: __Cooper Nuclear Station___________ Date of Examination: __10-02-06__

Examination Level: RO O SRO G Operating Test Number: ___1____

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

Conduct of Operations R, N Perform Jet Pump Operability Surveillance SKL034-20-XX Conduct of Operations R, N Reactor Recirc Pump Startup. Procedure 2.2.68.1 Attachment 1 Surveillance Testing - Review 6.HPCI.201 HPCI Valve Operability Equipment Control R, N Test (IST) (SKL034-50-49)

Radiation Control R, N Determine ALARA requirements for two workers.

Emergency Plan N/A N/A NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (# 3 for ROs; # 4 for SROs & RO retakes)

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

(P)revious 2 exams (# 1; randomly selected)

ES-301, Page 22 of 27

ES 301 Control Room/In Plant Systems Outline Form ES 301-2 Facility: Cooper Nuclear Station Date of Examination: 10/02/2006 Exam Level: RO O SRO-I G SRO-U G Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a. Manually Start SLC Injection per 2.2.74 A,E,P,S 1

b. Manually Start up the RCIC System (SKL034-20-21) NRC Developing A,E,S 2

c. Perform ADS Manual Valve Actuation Surveillance (SKL034-20-77) A,D,M,S 3

d. Perform the Latching and rolling of the Main Turbine per 2.2.77 A,N,S 4

e. Startup Suppression Pool Cooling Mode Of RHR NRC Developing A,N,S 5

f. Transfer 4160 VAC Bus 1G From DG2 To 4160 VAC Bus 1B D,S 6

g. Perform the Panel 9-5 section of 6-RWM-301 NRC Developing L,N,S 7

h. Perform the Control Room Operator Actions for a fire per 5.4FIRE E,N,S 8 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 5.3ALT-STRATEGY -SENSITIVE INFORMATION D,E,R 4

j. Startup RPS Motor Generator Set N,C 7

k. Manually Vent the Scram Air Header, per 5.8.3 NRC Developing E,N,R 1

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank # 9/# 8/# 4 (E)mergency or abnormal in-plant $ 1/$ 1/$ 1 (L)ow-Power / Shutdown $ 1/$ 1/$ 1 (N)ew or (M)odified from bank including 1(A) $ 2/$ 2/$ 1 (P)revious 2 exams # 3 / # 3 / # 2 (randomly selected)

(R)CA $ 1/$ 1/$ 1 (S)imulator ES-301, Page 23 of 27

ES 301 Administrative Topics Outline Form ES 301-1 Facility: _Cooper Nuclear Station_____ Date of Examination: _10-02-06____

Examination Level: RO G SRO O Operating Test Number: ___1_____

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

Conduct of Operations R,N Mode change requirements reviewed. Procedure 2.1.1 Conduct of Operations R,N Review Reactor Recirc Idle Loop Startup.

Equipment Control R,D Develop, Verify & Implement Tagouts (2) SKL034-50-XX Radiation Control R,D Determine the Rad Exposure during Emergency Emergency Plan S,D EAL Tabletop NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (# 3 for ROs; # 4 for SROs & RO retakes)

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

(P)revious 2 exams (# 1; randomly selected)

ES-301, Page 22 of 27

ES 301 Control Room/In Plant Systems Outline Form ES 301-2 Facility: Cooper Nuclear Station Date of Examination: 10/02/2006 Exam Level: RO G SRO-I G SRO-U O Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a. Manually Start SLC Injection per 2.2.74 A,E,P,S 1

b. N/A

c. N/A

d. Perform the Latching and rolling of the Main Turbine per 2.2.77 A,N,S 4

e. N/A

f. N/A

g. Perform the Panel 9-5 section of 6-RWM-301 L,N,S 7

h. N/A In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. N/A

j. Startup RPS Motor Generator Set N,C 7

k. Manually Vent the Scram Air Header, per 5.8.3 E,N,R 1

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank # 9/# 8/# 4 (E)mergency or abnormal in-plant $ 1/$ 1/$ 1 (L)ow-Power / Shutdown $ 1/$ 1/$ 1 (N)ew or (M)odified from bank including 1(A) $ 2/$ 2/$ 1 (P)revious 2 exams # 3 / # 3 / # 2 (randomly selected)

(R)CA $ 1/$ 1/$ 1 (S)imulator ES-301, Page 23 of 27

ES 301 Control Room/In Plant Systems Outline Form ES 301-2 Facility: Cooper Nuclear Station Date of Examination: 10/02/2006 Exam Level: RO G SRO-I O SRO-U G Operating Test No.: 1 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System / JPM Title Type Code*

Function

a. Manually Start SLC Injection per 2.2.74 A,E,P,S 1

b. Manually Start up the RCIC System (SKL034-20-21) A,M,E,S 2

c. Perform ADS Manual Valve Actuation Surveillance (SKL034-20-77) A,D,M,S 3

d. Perform the Latching and rolling of the Main Turbine per 2.2.77 A,N,S 4

e. Startup Suppression Pool Cooling Mode Of RHR A,N,S 5

f. Transfer 4160 VAC Bus 1G From DG2 To 4160 VAC Bus 1B D,S 6

g. Perform the Panel 9-5 section of 6-RWM-301 L,N,S 7

h. N/A In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. 5.3ALT-STRATEGY -SENSITIVE INFORMATION D,E,R 4

j. Startup RPS Motor Generator Set N,C 7

k. Manually Vent the Scram Air Header, per 5.8.3 E,N,R 1

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 (C)ontrol room (D)irect from bank # 9/# 8/# 4 (E)mergency or abnormal in-plant $ 1/$ 1/$ 1 (L)ow-Power / Shutdown $ 1/$ 1/$ 1 (N)ew or (M)odified from bank including 1(A) $ 2/$ 2/$ 1 (P)revious 2 exams # 3 / # 3 / # 2 (randomly selected)

(R)CA $ 1/$ 1/$ 1 (S)imulator ES-301, Page 23 of 27

Appendix D Scenario Outline Form ES D-1 Facility: _CNS___________ Scenario No.: _1__ Op-Test No.: _______

Examiners: ________________________ Operators: ___________________________

Initial Conditions: __The plant is operating at 100% power near the end of the current fuel cycle when the crew takes the shift. The plant is in a normal configuration with the B REC Heat Exchanger in Standby, and the A REC Heat Exchanger in service.

Turnover: The plant is operating at 100% power near the end of the current fuel cycle. The plant is in a normal configuration with the B REC Heat Exchanger in Standby, and the A REC Heat Exchanger in service. Procedure 2.2.65.1 is to be used to swap heat exchangers. REC HX B is in Standby in accordance with Section 19, and an Operator is standing by in R-931-REC HX area. River temperature is 65. REC temperature will be locally controlled after the HX swap.

Event Malf. Event Event No. No. Type* Description 1 N/A N Swap REC Heat Exchangers.

2 1 I LPRM fails downscale.

Condensate Booster Pump failure.

3 N/A C Rapidly decrease Reactor power using Recirculation HPCI inadvertently starts. (Damaged so that it will not start if 4 2 C needed.)

All Bypass Valves open. Reactor reaches level 8. Bypass 5 3 M valves close. Reactor does not scram (ATWS).

6 N/A C HPCI fails to start. Feedwater pumps fail to restart.

7 N/A N Emergency Depressurization

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES D-1 SCENARIO

SUMMARY

The crew is instructed to swap the in service heat exchanger. After the evolution is complete, a single LPRM fails downscale. The operators must bypass the LPRM in the APRM channel, but the APRM remains operable. The operators receive a fire alarm, and a call that the B Condensate Booster Pump is smoking. The operators commence a rapid power reduction and remove the pump from service. The power reduction is with Recirc only, and no rods will need to be moved immediately. There is no fire, and the pump stops smoking after it is de-energized.

After the plant has stabilized, and the fire is not a threat, HPCI inadvertently starts. The operators respond by securing the system. However, one of the critical breakers is damaged so that the system will not restart. The bypass valves go full open, causing Reactor level to reach

8. Most of the control rods do not scram due to channel bowing. HPCI and the A Feedwater Pump will not start, but RCIC starts and injects. However, the Reactor is at approximately 25%

power, so the crew is forced to Emergency Depressurize. When the crew Emergency Depressurizes, The Scenario ends with RCIC controlling level, boric acid injected, and rods driven in.

CRITICAL TASK

1. Insert Control Rods by RMCS or by Scram

2. Emergency Depressurize to allow low pressure systems to recover level.

Appendix D Scenario Outline Form ES D-1 Facility: Cooper Nuclear Station Scenario No.: 2 No.: _______

Examiners: ___________________________ Operators:___________________________

Initial Conditions: The plant is operating at 100% power at the End of Cycle. The plant is in a 7 day LCO due to 1A SLC pump being out of service to replace the discharge relief valve that has failed open.

Turnover: Today is not a red light day. Start SW Pump B and secure D SW Pump.

Event Malf. Event Event No. No. Type* Description

1. N/A N Swap Service Water Pumps

2. 1 C Service Water Pump D trip and LCO Feedwater Heater 5A Tube Failure 2.4Ex-Stm; Reduce

3. 2 C,R power to exit the Loss of Feedwater Heating Region.

4. 3 I NBI-LIS-101B failure that causes a 1/2 scram.

Hydrogen leak entry into 2.4GEN-H2; Reduce power to

5. 4 C,R allow repair of H2 Regulator.

6. 5 C CRD Pump B Trip

7. 6 M RR Pump vibration and eventual LOCA and Scram.

8. 6 C Loss of the Startup Transformer and Lockout of 4160 1G, Failure of CS injection valve to auto open. Use of RHR and

9. 7 C CS to restore reactor water level; Control Restore Reactor water level/Cool Containment

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES D-1 SCENARIO

SUMMARY

The plant is operating at 100% power with A SLC pump tagged out when the crew assumes the shift.

The crew performs a swap of the operating Service Water Pumps. When the B SW Pump is started the D SW Pump trips requiring entry into LCO 3.7.2. After the SW pumps are swapped and Tech Specs are addressed, the 5A Feedwater Heater develops a tube leak and causes a lowering feedwater temperature.

After power is lowered, the NBI-LIS-101 B fails low due to a partially open equalizing valve, resulting a 1/2 scram that can be fixed and the 1/2 scram reset. Once the 1/2 scram is reset, a hydrogen leak develops and lowers pressure to approximately 40 psig. Power will be reduced to approximately 85% to allow repairs. When the repairs are in progress, the B CRD Pump trips requiring the RO to startup the standby pump.

The major event starts as a vibration of the A Reactor Recirc Pump requiring the eventual tripping of the pump and entry into 2.4RR. The vibrations will cause a preexisting flaw in the RR pipe to fail resulting in a large RR pump discharge line break that is not isolable. When the turbine is tripped the Startup Transformer locks out and the emergency transformer picks up only 4160 1F. The CS and RHR pumps powered from 4160 1F automatically start but the CS injection fails to automatically open and must be manually opened. The leak is large enough so that RHR alone has insufficient capacity to refill the RPV and the CS valve must be opened to restore level.

The scenario ends when RPV level is being maintained 3" to 54".

CRITICAL TASKS

1. The crew shall restore 41601F to service to provide containment and core cooling.

2. The crew shall align CS injection valve to restore and maintain reactor water level greater than TAF.

Appendix D Scenario Outline Form ES D-1 Facility: _CNS__________ Scenario No.: _3__ Op-Test No.: _______

Examiners: ______________________ Operators: _________________________

Initial Conditions: The plant is operating at 32% power at the end of the fuel cycle with a normal reactor startup in progress. Containment Inerting is in progress. DPIC-835B, Reactor Building differential pressure controller is out of service due to an unknown failure.

Turnover: The plant is operating at 32% power at the end of the fuel cycle with a normal reactor startup in progress. The rod sequence is at RWM group 10/1, Step 5. A power ascension to 40% has been directed, at which point, a MSIV closure surveillance will be performed.

Containment Inerting is in progress Event Malf. Event Event Description No. No. Type

• 1 N/A R Raise power to 40% using Control rods and Reactor Recirc.

2 1 I RWM Failure halts the power ascension During MSIV Surveillance the 86D MSIV Fails As-Is in mid 3 2 C position.

Reactor Building ventilation failure due to the in service controller 4 3 I failing in Auto, causing building P to rise and go positive.

5 4 C Earthquake - Loss of off-site power requiring manual alignment LOCA - Medium Break, requiring use of low pressure systems to 6 5 M recover level.

• (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor

Appendix D Scenario Outline Form ES D-1 SCENARIO

SUMMARY

The crew raises power by pulling rods and raising Recirc flow. While they are pulling rods, the RWM fails INOP. The RWM must bypass or wait for repairs to finish pulling rods. The rod sequence is at RWM group 10/1, Step 5. A power ascension to 40% has been directed, at which point, the BOP operator will complete MSIV IST testing Section 5 for valve 86D only.

Once reactor power is approximately 40% the MSIV surveillance is commenced. The outboard 86D MSIV will fail mid-position and will not reopen. The CRS will order close the inboard MSIV in order to satisfy the TS. After the plant has stabilized, a malfunction of differential pressure controller DPIC-835A on the Reactor Building HVAC system results in high reactor building pressure and entry into EOP 5A Secondary Containment Control.

An earthquake results in a total loss of offsite power and a Medium Break LOCA simultaneously. Emergency Bus 1F locks out and the Diesel Generator Output Breaker 1GS does not automatically close. This forces the operator to manually close the breaker to make all of the equipment on the bus available.

CRITICAL TASK

1. Recovers Reactor Building Ventilation to restore it to a negative.

2. Realigns Electrical Power to supply Critical Busses

3. Restores Reactor level above TAF.