ML101230159

Initial Exam 2010-301 Draft Administrative Documents

ML101230159
Person / Time
Site:	Sequoyah
Issue date:	05/03/2010
From:	NRC/RGN-II
To:	Tennessee Valley Authority
References
50-327/10-301, 50-328/10-301
Download: ML101230159 (32)

Text

ES-401 1 Rev. 9 PWR Examination Outline Form ES-401-2 Facility:

Sequoyah Date of Exam: 2010 I RO KIA 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 3 3 3 3 3 18 3 3 6 Emergency

& 2 1 2 1 2 2 1 9 2 2 4 Abnormal Plant N/A N/A Evolutions Tier Totals 4 5 4 5 5 4 27 5 5 10 1 2 3 2 3 2 2 3 3 3 3 2 28 3 2 5 2. 1 1 0 1 1 1 1 1 1 1 1 10 2 1 3 Plant 2 Systems Tier Totals 3 4 2 4 3 3 4 4 4 4 3 38 5 3 8 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 3 2 2 2 1 2 1. Ensure that at least two topics from everY applicable KIA categorY are sampled within each tier of the RO and SRO-only outlines (Le., except for one categorY in Tier 3 ofthe SRO-only outline, the "Tier Totals" in each KIA 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 KIA statements.

4. Select topics from as many systems and evolutions as possible; sample everY system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those KIAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7. *The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 ofthe KIA Catalog, butthe topics must be relevant to the applicable evolution or system. 8. On the following pages, enter the KIA 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; iffuel 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 ofthe KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.

Limit SRO selections to KIAs that are linked to 10 CFR 55.43. ES-401 1 Rev. 9 PWR Examination Outline Form ES-401-2 Facility:

Sequoyah Date of Exam: 2010 I RO KIA 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 3 3 3 3 3 18 3 3 6 Emergency

& 2 1 2 1 2 2 1 9 2 2 4 Abnormal Plant N/A N/A Evolutions Tier Totals 4 5 4 5 5 4 27 5 5 10 1 2 3 2 3 2 2 3 3 3 3 2 28 3 2 5 2. 1 1 0 1 1 1 1 1 1 1 1 10 2 1 3 Plant 2 Systems Tier Totals 3 4 2 4 3 3 4 4 4 4 3 38 5 3 8 3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 3 2 2 2 1 2 1. Ensure that at least two topics from everY applicable KIA categorY are sampled within each tier of the RO and SRO-only outlines (Le., except for one categorY in Tier 3 ofthe SRO-only outline, the "Tier Totals" in each KIA 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 KIA statements.

4. Select topics from as many systems and evolutions as possible; sample everY system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those KIAs having an importance rating (IR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7. *The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 ofthe KIA Catalog, butthe topics must be relevant to the applicable evolution or system. 8. On the following pages, enter the KIA 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; iffuel 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 ofthe KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.

Limit SRO selections to KIAs that are linked to 10 CFR 55.43.

ES-401 Form ES-401-2 PWR Examination Outline Emerqencv and Abnormal Plant Evolutions

-Tier 1/Group 1((RO 1 SHe) Form '-.../ E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor R 007 [;ALIO 3.7 Trip -Stabilization

-Recovery 1 1 000008 Pressurizer Vapor Space -Accident 1 3 000009 Small Break LOCA 1 3 R Or&(

4-, ;L 000011 Large Break LOCA 13 f\

(;lAc;) 000015/17 RCP Malfunctions 14 R 01 SAl< /. () I 4,4 000022 Loss of Rx Coolant Makeup 1 2 R OJ."J. AA /,07 ;;Z .g 000025 Loss of RHR System 14 R Dc;( 5 AI< 3, {);;L 33 000026 Loss of Component Cooling -Water 1 8 000027 Pressurizer Pressure Control R OOl? AA/.()3 3,c" System Malfunction 1 3 000029 ATWS 1 1 R 0;)1 '{;;<;< ,() " ;2,1 000038 Steam Gen. Tube Rupture 1 3 R 038 AC(, I£:' 3,g 000040 (BW/E05; CE/E05; W/E12) R A I< /.0 .0< Steam Line Rupture -Excessive Heat Transfer 1 4 000054 (CE/E06) Loss of Main R D64AG'J.,J,'6J 4,fo Feedwater 14 000055 Station Blackout 1 6 --000056 Loss of Off-site Power 1 6 R 05b A/<3,O( 3,S" 000057 Loss of Vital AC Inst. Bus 16 f< 057 AA;;Z';£:'

3, () 000058 Loss of DC Power 1 6 --000062 Loss of Nuclear Svc Water 1 4 ob2 AGe< 14, '1 13,g 000065 Loss of Instrument Air 1 8 I< 0(05"" A 1<3,04-3.0 W/E04 LOCA Outside Containment 1 3 R W£(?J1-£A;<. :2 13.b W/E11 Loss of Emergency Coolant Recirc. 1 4 BW/E04; W/E05 Inadequate Heat A < , '3/1 Transfer -Loss of Secondary Heat Sink 1 4 ../ 000077 Generator Voltage and Electric R 0 iG 4,;1 Grid Disturbances 1 6 '/1 '"' T. "j !J Group Point Total: 18/6 ES-401 Form ES-401-2 PWR Examination Outline Emerqencv and Abnormal Plant Evolutions

-Tier 1/Group 1((RO 1 SHe) Form '-.../ E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor R 007 [;ALIO 3.7 Trip -Stabilization

-Recovery 1 1 000008 Pressurizer Vapor Space -Accident 1 3 000009 Small Break LOCA 1 3 R Or&(

4-, ;L 000011 Large Break LOCA 13 f\

(;lAc;) 000015/17 RCP Malfunctions 14 R 01 SAl< /. () I 4,4 000022 Loss of Rx Coolant Makeup 1 2 R OJ."J. AA /,07 ;;Z .g 000025 Loss of RHR System 14 R Dc;( 5 AI< 3, {);;L 33 000026 Loss of Component Cooling -Water 1 8 000027 Pressurizer Pressure Control R OOl? AA/.()3 3,c" System Malfunction 1 3 000029 ATWS 1 1 R 0;)1 '{;;<;< ,() " ;2,1 000038 Steam Gen. Tube Rupture 1 3 R 038 AC(, I£:' 3,g 000040 (BW/E05; CE/E05; W/E12) R A I< /.0 .0< Steam Line Rupture -Excessive Heat Transfer 1 4 000054 (CE/E06) Loss of Main R D64AG'J.,J,'6J 4,fo Feedwater 14 000055 Station Blackout 1 6 --000056 Loss of Off-site Power 1 6 R 05b A/<3,O( 3,S" 000057 Loss of Vital AC Inst. Bus 16 f< 057 AA;;Z';£:'

3, () 000058 Loss of DC Power 1 6 --000062 Loss of Nuclear Svc Water 1 4 ob2 AGe< 14, '1 13,g 000065 Loss of Instrument Air 1 8 I< 0(05"" A 1<3,04-3.0 W/E04 LOCA Outside Containment 1 3 R W£(?J1-£A;<. :2 13.b W/E11 Loss of Emergency Coolant Recirc. 1 4 BW/E04; W/E05 Inadequate Heat A < , '3/1 Transfer -Loss of Secondary Heat Sink 1 4 ../ 000077 Generator Voltage and Electric R 0 iG 4,;1 Grid Disturbances 1 6 '/1 '"' T. "j !J Group Point Total: 18/6 ES-401, REV 9 R 0 T1G1 PWR EXAMINATION OUTLINE KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 007EA1.10 Reactor Trip -Stabilization

-Recovery 3.7 3.7 / 1 o 0 009EK1.01 Small Break LOCA / 3 4.2 4.7 0 011EK2.02 Large Break LOCA / 3 2.6 2.7 00000 0 015AK1.01 RCP Malfunctions / 4 4.4 4.6 000000000 022AA1.07 Loss of Rx Coolant Makeup / 2 2.8 2.7 0 025AK3.02 Loss of RHR System / 4 3.3 3.7 0000000 027AA1.03 Pressurizer Pressure Control System 3.6 3.5 Malfunction / 3 029EK2.06 ATWS / 1 2.9 3.1 DO 038EA2.06 Steam Gen. Tube Rupture / 3 3.8 4.4 0 DOD 040AK1.04 Steam Line Rupture -Excessive Heat 3.2 3.6 0 0 0 0 0 0 0 Transfer / 4 054AG2.1.31 Loss of Main Feedwater / 4 4.6 4.3 0 0 0 0 0 Page 1 of 2 FORM ES-401-2 TOPIC: S/G pressure Natural circulation and cooling, including reflux boiling Pumps Natural circulation in a nuclear reactor power plant Excess letdown containment isolation valve switches and indicators Isolation of RHR low-pressure piping prior to pressure increase above specified level -""""------------

Pressure control when on a steam bubble Breakers, relays, and disconnects.

Shutdown margins and required boron concentrations Nil ductility temperature Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup. 6/25/2009 12:27 PM ES-401, REV 9 R 0 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 007EA1.10 Reactor Trip -Stabilization

-Recovery 3.7 3.7 0 0 o S/G pressure / 1 009EK1.01 Small Break LOCA / 3 011EK2.02 Large Break LOCA / 3 015AK1.01 RCP Malfunctions / 4 022AA1.07 Loss of Rx Coolant Makeup / 2 025AK3.02 Loss of RHR System / 4 4.2 4.7 0 0 0 0 .-=----::::--=_-.---:c-:-------c--:----:---c------:--c:-

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.. ---.. o Natural circulation and cooling, including reflux boiling ---_._---------------------------_._

.. _--2.6 2.7 0 0 0 0 0 0 Pumps 4.4 4.6 000000000 2.8 2.7 0 3.3 3.7 0000000 Natural circulation in a nuclear reactor power plant Excess letdown containment isolation valve switches and indicators Isolation of RHR low-pressure piping prior to pressure increase above specified level -----_ ... _--------------------------

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_ .... _._ ... _---_. 027AA1.03 Pressurizer Pressure Control System 3.6 3.5 Pressure control when on a steam bubble Malfunction / 3 029EK2.06 ATWS / 1 2.9 3.1 DO Breakers, relays, and disconnects.

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3.8 4.4 0 0 0 0 0 0 0 0 Shutdown margins and required boron concentrations 038EA2.06 Steam Gen. Tube Rupture / 3 040AK1.04 Steam Line Rupture -Excessive Heat 3.2 3.6 Transfer / 4 054AG2.1.31 Loss of Main Feedwater / 4 4.6 4.3 DODD DOD 00000 Page 1 of 2 Nil ductility temperature Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup. 6/25/2009 12:27 PM ES-401, REV 9 R 0 KA NAME / SAFETY FUNCTION:

OS6AK3.01 Loss of Off-site Power / 6 OS7AA2.16 Loss of Vital AC Inst. Bus / 6 062AG2.4.9 Loss of Nuclear Svc Water / 4 06SAK3.04 Loss of Instrument Air / 8 077AG2.2.44 Generator Voltage and Electric Grid Disturbances / 6 WE04EA2.2 LOCA Outside Containment / 3 WEOSEK2.2 Inadequate Heat Transfer -Loss of Secondary Heat Sink / 4 T1G1 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G RO SRO 3.S 3.9 0 0 3 3.1 0 0 0 0 0 0 0 0 0 0 3.8 4.2 0 0 0 0 3 3.2 0 000 4.2 4.4 o 3.6 4.2 0 0 0 0 0 3.9 4.2 0 0 0 0 0 0 0 0 Page 2 of 2 FORM ES-401-2 TOPIC: Order and time to initiation of power for the load sequencer Normal and abnormal PZR level for various modes of plant operation Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

Cross-over to backup air supplies Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

6/2S/2009 12:27 PM ES-401, REV 9 R 0 KA NAME / SAFETY FUNCTION:

OS6AK3.01 Loss of Off-site Power / 6 OS7AA2.16 Loss of Vital AC Inst. Bus / 6 062AG2.4.9 Loss of Nuclear Svc Water / 4 T1G1 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G RO SRO 3.S 3.9 0 3 3.1 0 0 0 0 0 0 0 0 0 0 3.8 4.2 0 0 0 0 FORM ES-401-2 TOPIC: Order and time to initiation of power for the load sequencer Normal and abnormal PZR level for various modes of plant operation Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

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06SAK3.04 Loss of Instrument Air / 8 3 3.2 0 Cross-over to backup air supplies 077AG2.2.44 Generator Voltage and Electric Grid Disturbances / 6 WE04EA2.2 LOCA Outside Containment / 3 WEOSEK2.2 Inadequate Heat Transfer -Loss of Secondary Heat Sink / 4 0000 4.2 4.4 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions

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3.6 4.2 0 0 0 0 0 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

3.9 4.2 0 0 0 0 0 0 0 0 Page 2 of 2 Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

6/2S/2009 12:27 PM ES-401 3 Form ES-401-2 r;S-401 PWR Examination Outline ! Form ES-401-2 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 RO I E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR # 1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 R O;L3 AK;;( ,O;;L R,b 000032 Loss of Source Range NI / 7 P. 0"3;2 AK;;(/Ol 0{,1 000033 Loss of Intermediate Range NI /7 0'33' AA;2" J2 ;;,S 000036 (BW/A08) Fuel Handling Accident /8 It; 03&

3) 3"g 000037 Steam Generator Tube Leak / 3 R 031 It A/, 04-5,t, 000051 Loss of Condenser Vacuum /4 000059 Accidental Liquid RadWaste ReI. / 9 000060 Accidental Gaseous Radwaste ReI. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. /8 {( 068 A K3, 1r1-4, I 000069 (W/E14) Loss of CTMT Integrity

/5 . 000074 (W/E06&E07)

Inad. Core Cooling /4 000076 High Reactor Coolant Activity / 9 W/E01 & E02 Rediagnosis

& SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation

/9 BW/A01 Plant Runback /1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin /4 BW/E08; W/E03 LOCA Cooldown -Depress. / 4 R WE (!) 3 GA :2,1 , 3.4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 R WelD GAf3 3,4-BW/E13&E14 EOP Rules and Enclosures CE/A 11; W/E08 RCS Overcooling

-PTS / 4 1< (fiE 08 t;J!.I, J 3.5 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery f". D. int Totals: I ES-401 3 Form ES-401-2 r;S-401 PWR Examination Outline ! Form ES-401-2 Emergency and Abnormal Plant Evolutions

-Tier 1/Group 2 RO I E/APE # / Name / Safety Function K K K A A G K/A Topic(s) IR # 1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 000024 Emergency Boration / 1 000028 Pressurizer Level Malfunction / 2 R O;L3 AK;;( ,O;;L R,b 000032 Loss of Source Range NI / 7 P. 0"3;2 AK;;(/Ol 0{,1 000033 Loss of Intermediate Range NI /7 0'33' AA;2" J2 ;;,S 000036 (BW/A08) Fuel Handling Accident /8 It; 03&

3) 3"g 000037 Steam Generator Tube Leak / 3 R 031 It A/, 04-5,t, 000051 Loss of Condenser Vacuum /4 000059 Accidental Liquid RadWaste ReI. / 9 000060 Accidental Gaseous Radwaste ReI. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. /8 {( 068 A K3, 1r1-4, I 000069 (W/E14) Loss of CTMT Integrity

/5 . 000074 (W/E06&E07)

Inad. Core Cooling /4 000076 High Reactor Coolant Activity / 9 W/E01 & E02 Rediagnosis

& SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation

/9 BW/A01 Plant Runback /1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip / 4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin /4 BW/E08; W/E03 LOCA Cooldown -Depress. / 4 R WE (!) 3 GA :2,1 , 3.4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 R WelD GAf3 3,4-BW/E13&E14 EOP Rules and Enclosures CE/A 11; W/E08 RCS Overcooling

-PTS / 4 1< (fiE 08 t;J!.I, J 3.5 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery f". D. int Totals: I ES-401, REV 9 {<.O KA NAME / SAFETY FUNCTION:

02SAK2.02 Pressurizer Level Malfunction / 2 032AK2.01 Loss of Source Range NI /7 033AA2.12 Loss of Intermediate Range NI / 7 036AG2.4.35 Fuel Handling Accident / S 037AA1.04 Steam Generator Tube Leak / 3 06SAK3.12 Control Room Evac. / S WE03EA2.1 LOCA Cool down -Depress. / 4 WEOSEK1.1 RCS Overcooling

-PTS / 4 WE1 OEA 1.3 Natural Circ. With Seam Void/ 4 T1G2 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 2.6 2.7 DO DO 2.7 3.1 0 DOD DO 2.5 3.1 0 0 0 0 0 0 o 3.S 4.0 DOD 0 3.6 3.9 DO 0 4.1 4.5 0 0 0 0 0 0 0 3.4 4.2 0 0 DOD 3.5 3.S 0 0 o DOD 3.4 3.7 0 0 0 0 0 Page 1 of 1 FORM ES-401-2 TOPIC: Sensors and detectors Power supplies, including proper switch positions Maximum allowable channel disagreement Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects Condensate air ejector exhaust radiation monitor and failure indicator Required sequence of actions for emergency evacuation of control room Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Components, capacity, and function of emergency systems. Desired operating results during abnormal and emergency situations.

6/25/2009 12:27 PM ES-401, REV 9 {<.O T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 02SAK2.02 Pressurizer Level Malfunction / 2 2.6 2.7 D D Sensors and detectors 032AK2.01 Loss of Source Range NI /7 2.7 3.1 D DD Power supplies, including proper switch positions 033AA2.12 Loss of Intermediate Range NI / 7 2.5 3.1 0 0000 o Maximum allc)walble channel disagreement

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036AG2.4.35 Fuel Handling Accident / S 3.S 4.0 037AA1.04 Steam Generator Tube Leak 3 3.6 3.9 06SAK3.12 Control Room Evac. / S 4.1 4.5 WE03EA2.1 LOCA Cool down -Depress. / 4 3.4 4.2 WEOSEK1.1 RCS Overcooling

-PTS / 4 3.5 3.S WE10EA1.3 Natural Circ. With Seam Void/ 4 3.4 3.7 D DD DD Knowledge of local auxiliary operator tasks during D DDD DDD Page 1 of 1 emergency and the resultant operational effects Condensate air ejector exhaust radiation monitor and failure indicator Required sequence of actions for emergency evacuation of control room Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Components, capacity, and function of emergency systems. Desired operating results during abnormal and emergency situations.

6/25/2009 12:27 PM ES-401 5QN 2010 -120 EXaM 4 Form ES-401-2 ES-401 PWR Examination Form ES-401-2 Plant S stems -Tier 2/Group 1 RO -SRG) System # / Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump R f)() 3 j('t;, Of 3/3 004 Chemical and Volume R &04 A 3, oS' 3*1 Control 004 A4;13 3,3 005 Residual Heat Removal K ooS )<5,oS c:(,7 006 Emergency Core Cooling f{ 00 b AI, 11-' 3,(;, 007 Pressurizer Relief/Quench R 13,/) Tank 008 Component Cooling Water R f{ (Jog A /, '1-/ f);< 3.,-3'0 010 Pressurizer Pressure Control K Off) I< 4-, 3,0 012 Reactor Protection R R f) 1;( Ii I, 01 ) 1<::<,01 013 Engineered Safety Features R 0/3 ' 3,(" Actuation 0/3 Xi 01 0:(.7 022 Containment Cooling !Z 0;((;( I< / , 01 3,S 025 Ice Condenser 1\ 0;(5' 1<1-, 021 !j 026 Containment Spray IZ O;:zro 4-.;(. 039 Main and Reheat Steam R O?1 G, Q ,J. 7 44-059 Main Feedwater tJ'S'1' )(3,0:( :3,(" 061 Auxiliary/Emergency oCt! I j,;Zo if" Feedwater 062 AC Electrical Distribution K Ob(;l J< 4, /)::L RS 063 DC Electrical Distribution R 003 AJ ,01 ;ZS 064 Emergency Diesel Generator K K 10&4-A4:IJ-Kb,t)S 073 Process Radiation R fl f. >;L Monitoring 3,7 076 Service Water R 01& A3.0;t 13,7 078 Instrument Air A 'D1g K;;{, )2 3,3 103 Containment IA fZ 103 /l;;(, of A3,ol 3'!ij 3,. r-. n .. Totals: mrmr;: :3 j :) ,3 2 Group Point Total: 28/5 ES-401 5QN 2010 -120 EXaM 4 Form ES-401-2 ES-401 PWR Examination F cC' '" .., Plant S stems -Tier 2/Group 1 RO .sRG) System # / Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump R f)() 3 j('t;, 'I 3/3 004 Chemical and Volume R &04 A 3, oS' 3*1 Control 004 A4;13 3,3 005 Residual Heat Removal IR ooS )<5'05 c:(,7 006 Emerqency Core Coolinq f{ OOb AI,I1-3,(;, 007 Pressurizer Relief/Quench R 13,/) Tank 008 Component Coolina Water R f{ IvogA/,Of-/

fit?l , 03 '3./ 3'0 010 Pressurizer Pressure Control K Off) I< 4-, 3,0 012 Reactor Protection R R f) 1;( Ii I, 01 ) 1<::<,01 013 Engineered Safety Features p R 0/3 ' 3,(" Actuation

() 1.3 Xi ,01 0:(.7 -022 Containment Coolina !Z 0;((;( I< / , 01 3,S 025 Ice Condenser 1\ 0;(5' 1<1-, 021 !j 026 Containment Spray IZ O;:zro 4-.;(. 039 Main and Reheat Steam R 031 G,Q'}*7 144 059 Main Feedwater D'S'1' )(3,0:( :3,(" 061 Auxiliary/Emergency

!\ oCt! I j,;z,o if" Feedwater 062 AC Electrical Distribution K Ob(;l J< 4, /)::L RS 063 DC Electrical Distribution R 003 AJ ,01 064 Emerqency Diesel Generator K K 10&4-A4:IJ-Kb,t)S 073 Process Radiation R

Monitoring 3,7 076 Service Water R 01& A3,0;t 13,7 078 Instrument Air A . D1g K;;{,D2 3,3 103 Containment IA fZ 103 /l;;(1 of A3,ol 3'!ij 3,. Totals: l 3 2 3 l;;t :3 k3 :) 13 2 Group Point Total: 28/5 ES-401, REV 9 1<-0 KA NAME / SAFETY FUNCTION:

003KS.01 Reactor Coolant Pump 004A3.0S Chemical and Volume Control 004A4.13 Chemical and Volume Control OOSKS OS Residual Heat Removal 006A1.14 Emergency Cooling 007K1.03 Pressurizer Relief/Quench Tank 008A1.04 Component Cooling Water 008A2.03 Component Water 010K4.02 Pressurizer Pressure Control 012A1.01 Reactor Protection 012K2.01 Reactor Protection T2G1 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G RO SRO 3.3 3.9 0 0 0 0 0 0 0 3.9 3.9 0 0 0 0 0 0 0 3.3 2.9 00 2.7 3.1 0 0 0 00 3.6 3.9 0 0 3.0 3.2 0 0 0 0 0 0 0 3.1 3.2 0 0 0 3.0 3.2 0 0 0 0 0 0 0 3.0 3.4 0 00 00 2.9 3.4 0 0 0 0 0 0 3.3 3.7 000 00 Page 1 of 3 FORM ES-401-2 TOPIC: The relationship between the RCPS flow rate and the nuclear reactor core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density, difference in loop T-hot pressure) pressure and temperature VCT level control and pressure control Plant response during "solid plant": pressure change due to the relative incompressibility of water Reactor vessel level Surge tank level High/low CCW temperature Prevention of uncovering PZR heaters Trip setpoint adjustment RPS channels, components and interconnections 6/2S/2009 12:27 PM ES-401, REV 9 1<-0 KA NAME / SAFETY FUNCTION:

003KS.01 Reactor Coolant Pump 004A3.0S Chemical and Volume Control 004A4.13 Chemical and Volume Control OOSKS OS Residual Heat Removal 006A1.14 Emergency Cooling 007K1.03 Pressurizer Relief/Quench Tank 008A1.04 Component Cooling Water 008A2.03 Component Water 010K4.02 Pressurizer Pressure Control 012A1.01 Reactor Protection 012K2.01 Reactor Protection T2G1 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G RO SRO 3.3 3.9 0 0 3.9 3.9 0 0000 3.3 2.9 FORM ES-401-2 TOPIC: The relationship between the RCPS flow rate and the nuclear reactor core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density, difference in loop T-hot pressure) pressure and temperature VCT level control and pressure control --------_._------_._._-_.-

2.7 3.1 0 0 3.6 3.9 0 3.0 3.2 0 0 3.1 3.2 3.0 3.2 0 3.0 3.4 0 2.9 3.4 DO 3.3 3.7 000 o 00000 0 DODD DO Page 1 of 3 Plant response during "solid plant": pressure change due to the relative incompressibility of water Reactor vessel level Surge tank level High/low CCW temperature Prevention of uncovering PZR heaters Trip setpoint adjustment RPS channels, components and interconnections 6/2S/2009 12:27 PM ES-401, REV 9 RO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 013K2.01 Engineered Safety Features Actuation 3.6 3.8 0 0000 ESFAS/safeguards equipment control 013K6.01 Engineered Safety Features Actuation 2.7 3.1 Sensors and detectors 022K1.01 Containment Cooling 3.5 3.7 0 0 0 0 0 0 0 SWS/cooling system 025K4.02 Ice Condenser 2.8 3.0 0 0 0 0 0 System control 026K3.02 Containment Spray 4.2 4.3 0 0 0 0 0 0 0 Recirculation spray system 039G2.1.7 Main and Reheat Steam 4.4 4.7 nn 0 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

059K3.02 Main Feedwater 3.6 3.7 AFW system 061G2.1.20 Auxiliary/Emergency Feedwater 4.6 4.6 Ability to execute procedure steps. 062K4.02 AC Electrical Distribution 2.5 2.7 Circuit breaker automatic trips 063A2.01 DC Electrical Distribution 2.5 3.2 Grounds 064A4.12 Emergency Diesel Generator 2.7 2.6 Synchroscope Page 2 of 3 6/25/2009 12:27 PM ES-401, REV 9 RO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 013K2.01 Engineered Safety Features Actuation 3.6 3.8 0000 ESFAS/safeguards equipment control

..... ---------------

013K6.01 Engineered Safety Features Actuation 2.7 3.1 0 0 0 0 0 Sensors and detectors 022K1.01 Containment Cooling 3.5 3.7 DO SWS/cooling system 025K4.02 Ice Condenser 2.8 3.0 DO DOD System control 026K3.02 Containment Spray 4.2 4.3 0 DODD Recirculation spray system 039G2.1.7 Main and Reheat Steam 4.4 4.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

059K3.02 Main Feedwater 3.6 3.7 AFW system 061 G2.1.20 Auxiliary/Emergency Feedwater 4.6 4.6 0 0 0 0 0 0 0 0 0 0 Ability to execute procedure steps. 062K4.02 AC Electrical Distribution 2.5 2.7 0 0 0 0 0 0 0 Circuit breaker automatic trips 2.5 3.2 0 0 0 0 0 0 0 0 0 0

..

Grounds 063A2.01 DC Electrical Distribution

..

Synchroscope Emergency Diesel Generator 2.7 2.6 0 0 0 0 0 0 ODD 0 064A4.12 Page 2 of 3 6/25/2009 12:27 PM ES-401, REV 9 Ro KA NAME / SAFETY FUNCTION:

064K6.08 Emergency Diesel Generator 073A4.02 Process Radiation Monitoring 076A3.02 Service Water 078K2.02 Instrument Air 103A2.04 Containment 103A3.01 Containment T2G1 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.2 3.3 0 0 0 0 0 0 0 0 0 0 3.7 3.7 3.7 3.7 0 0 0 0 3.3 3.5 0 0 0 0 0 0 000 3.5 3.6 3.9 4.2 0 0 0 0 0 0 0 Page 3 of 3 FORM ES-401-2 TOPIC: Fuel oil storage tanks Radiation monitoring system control panel Emergency heat loads Emergency air compressor Containment evacuation (including recognition of the alarm) Containment isolation 6/25/2009 12:27 PM ES-401, REV 9 Ro T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 064K6.08 Emergency Diesel Generator 3.2 3.3 D D D D D D D D D D Fuel oil storage tanks 073A4.02 Process Radiation Monitoring 3.7 3.7 D D D D D D Radiation monitoring system control panel 076A3.02 Service Water 3.7 3.7 D D D D Emergency heat loads 078K2.02 Instrument Air 3.3 3.5 Emergency air compressor 103A2.04 Containment 3.5 3.6 Containment evacuation (including recognition of the alarm) 103A3.01 Containment 3.9 4.2 DDDDDDD Containment isolation Page 3 of 3 6/25/2009 12:27 PM ES-401 5 Form ES-401-2 ES-401 PWR Examination F Plant S stems -Tier 2/Group 2 RO SRe-) System # 1 Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive R 001 /<;<,0S; 6,1 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication R f) It A4. 0/ "33 015 Nuclear Instrumentation

!<, DISA 3, os-;;;, to 016 Non-nuclear Instrumentation R O/b /<5, 'I ;;:),7 017 In-core Temperature Monitor II? DF7 3, I 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge t< ODl1 At, D;;L 34-033 Spent Fuel Pool Cooling R 0'33 G ;; ,4-,;( I 14,0 034 Fuel Handling Equipment R )31-t: t:"oOL Ic;(,b 035 Steam Generator K 035 1<4.01 041 Steam DumpITurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate IZ O!S(;' 1<1,0], ;(,10 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection KIA Category Point Totals: IITL0lIili I I I I I Group Point Total: 10/3 ES-401 5 Form ES-401-2 ES-401 PWR Examination F Plant S stems -Tier 2/Group 2 RO SRe-) System # 1 Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive R 001 /<;<,0S; 6,1 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication R f) It A4. 0/ "33 015 Nuclear Instrumentation

!<, DISA 3, os-;;;, to 016 Non-nuclear Instrumentation R O/b /<5, 'I ;;:),7 017 In-core Temperature Monitor II? DF7 3, I 027 Containment Iodine Removal 028 Hydrogen Recombiner and Purge Control 029 Containment Purge t< ODl1 At, D;;L 34-033 Spent Fuel Pool Cooling R 0'33 G ;; ,4-,;( I 14,0 034 Fuel Handling Equipment R )31-t: t:"oOL Ic;(,b 035 Steam Generator K 035 1<4.01 041 Steam DumpITurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal 056 Condensate IZ O!S(;' 1<1,0], ;(,10 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water 079 Station Air 086 Fire Protection KIA Category Point Totals: IITL0lIili I I I I I Group Point Total: 10/3 ES-401, REV 9 R 0 KA NAME / SAFETY FUNCTION:

001 K2.05 Control Rod Drive 014A4.01 Rod Position Indication 015A3.05 Nuclear Instrumentation 016K5.01 Non-nuclear Instrumentation 017A2.01 In-core Temperature Monitor 029A1.02 Containment Purge 033G2.4.21 Spent Fuel Pool Cooling 034K6.02 Fuel Handling Equipment 035K4.01 Steam Generator 056K1.03 Condensate T2G2 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.1 3.5 0 0 0 0 3.3 3.1 0 0 0 0 0 0 0 0 0 0 2.6 2.7 0 0 0 0 0 0 2.7 2.8 0 0 0 3.1 3.5 0 3.4 3.4 0 000 4.0 4.6 0 0 0 0 0 2.6 3.3 n noD 0 0 0 0 3.6 3.8 0 0 2.6 2.6 000000 0 Page 1 of 1 FORM ES-401-2 TOPIC: MIG sets Rod selection control Recognition of audio output expected for a given plant condition Separation of control and protection circuits Thermocouple open and short circuits Radiation levels Knowledge of the parameters and logic used to assess the status of safety functions Radiation monitoring systems SIG level control MFW 6/25/2009 12:27 PM ES-401, REV 9 R 0 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION:

IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 001 K2.05 Control Rod Drive 3.1 3.5 0 0 0 0 0 MIG sets ---------------

014A4.01 Rod Position Indication 3.3 3.1 0 0 0 0 0 0 0 0 0 0 Rod selection control --------------------------------------------::

015A3.05 Nuclear Instrumentation 2.6 2.7 0 0 0 0 0 0 Recognition of audio output expected for a given plant condition 016K5.01 Non-nuclear Instrumentation 2.7 2.8 0 DO Separation of control and protection circuits 017A2.01 In-core Temperature Monitor 3.1 3.5 =-=------------------

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,-----:--

o 0 0 0 0 0 0 0 0 Thermocouple open and short circuits 029A1.02 Containment Purge 3.4 3.4 DOD Radiation levels 033G2.4.21 Spent Fuel Pool Cooling 4.0 4.6 0 0 0 0 0 Knowledge of the parameters and logic used to assess the status of safety functions 034K6.02 Fuel Handling Equipment 2.6 3.3 Radiation monitoring systems 035K4.01 Steam Generator 3.6 3.8 SIG level control 056K1.03 Condensate 2.6 2.6 000000 o MFW Page 1 of 1 6/25/2009 12:27 PM ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facilit

_" c. c<o/D Category KIA # Topic RO SRO-Only IR # IR # 2.1. 11 Use p/tl././c:t Cz"AcpvAcA -h e va.,tt(?Jfe 3/1 1A-2.1.3;;2 £Xp!t2//./iJ

P:L '5 3,6 1 1. Conduct 2.1.4D f?r,

/¥j ft7 IZ 0 . f J' ?<, of Operations v . \ 2.1. 2.1. I 2.1. Subtotal 3 2.2.0 Proce'sr;.fvJ Yr\;:dUf'!P'j Ch:Vi'O', *Io prOC'r 3J) J 2.2.40 P-ppi.tJ /.fch .. rp,fCs:' 3,4-v 2. 2.2. Equipment

2.2. Control

2.2. I 2.2. Subtotal . . 2.3. t; vadi4.

yyV1;VI.{A'/:rL

.* ;vt(;{ ;;;'t4)lEJ, I.§ C)/1 2.3.7 tCY"1'4 K! f/Jf rrv, 'tJ

/A.hM) M",F S,S 3. 2.3.12 CvJ/?v':J 'fl. flAi'J Jl£l-u/2lt" r IIcH 3*Z Radiation " . Control 2.3. r 2.3. 2.3. Subtotal 1.3. :. 2.4.;<g rv,Jop );/J,.1*C A JJf,Ar.rr;.,.&'}.t..

3,2 2.4.4b V8'Vil v (. /7 () t-)/i ML,l rXriR!4>!?r:f

"'/ f /t?r.f C1'/;;:f'r 4,2 4. . Emergency

2.4. Procedures

/ "-Pian 2.4. 2.4. .11 2.4. I/JA 'V Subtotal ::l . ,J)4. Tier 3 Point Total If) 10 ....*... 7 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 II .. /"

Date of Exam: 'OlD Category KIA # Topic RO SRO-Only IR # IR # 2.1. 11 !J se p / t/././c:t tpU0!J .1c:A -h e vaet ( ?lie 31 iV/A-1. 2.1.3;;2 £X/;/,;?//A.

P'L '5 3,6 Conduct 2.1.4D {;,

IZ UJ 'i:r ?<, of Operations v J \ 2.1. 2.1. I 2.1. Subtotal 3 2.2.0 Prot: es:;.fvJ YP;!tiu t'lP] {" h:vi 'e' ( *Ill jJ foe 'r 3J) I 2.2.40 P-pp4; /.fch .. rr;:;.fCs:'

3,4-u 2. 2.2. 1 Equipment

2.2. Control

2.2. 2.2. Subtotal ;(,. 2.3. t; yy0;V!Jvt'lf'L,,;Vt(;{

C)/1 2.3.7 tt* 1'4 K! "'JP rrv; 'tJ

/.lthM) Mil S,S 3. 2.3. 12. (:y, jeW ,v'l '1/,{

iCk? 'II. flAi'J JI£I-,,/:11]

I r IICH 3*Z Radiation If . Control 2.3. 2.3. 2.3. Subtotal 1.3 '.'. 2.4.;<g Pr JeRel Vt/!.<J f",) rVcP}/J,;2,C A ..f AAr.rr;J(,&'jt. >12 2.4.4b tUJfI/;' .,/j rXriRW>/?r:f "'j I (p../.f C1'/;;1 r 4,2 4. , Emergency

2.4. Procedures

/ "-Pian 2.4. 2.4. 2.4. ,1/ Subtotal ::2 ** Tier 3 Point Total If) . 10 ., .....*. 7 ES-401, REV 9 gD KA NAME / SAFETY FUNCTION: G2.1.19 Conduct of operations . _-G2.1.32 Conduct of operations G2.1.40 Conduct of operations G2.2.40 Equipment Control G2.2.6 Equipment Control G2.3.12 Radiation Control G2.3.5 Radiation Control G2.3.7 Radiation Control G2.4.28 Emergency Procedures/Plans G2.4.46 Emergency Procedures/Plans T3 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.9 3.8 3.8 4.0 2.8 3.9 o 3.4 4.7 0 0 0 0 0 0 0 0 0 0 3.0 3.6 0 0 0 0 0 3.2 3.7 0 0 0 0 0 0 0 0 0 2.9 2.9 3.5 3.6 0 0 0 0 0 0 0 0 3.2 4.1 0 4.2 4.2 DO DOD Page 1 of 1 FORM ES-401-2 TOPIC: Ability to use plant computer to evaluate system or component status . Ability to explain and apply all system limits and precautions. Knowledge of refueling administrative requirements Ability to apply technical specifications for a system. Knowledge of the process for making changes to procedures Knowledge of radiological safety principles pertaining to licensed operator duties Ability to use radiation monitoring systems Ability to comply with radiation work permit requirements during normal or abnormal conditions Knowledge of procedures relating to emergency response to sabotage. to verify that the alarms are consistent with the plant conditions. 6/25/2009 12:27 PM ES-401, REV 9 gD KA NAME / SAFETY FUNCTION: G2.1.19 Conduct of operations T3 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.9 3.8 FORM ES-401-2 TOPIC: Ability to use plant computer to evaluate system or component status . . ____ =c___c=-_==_:=__==__=::_==_==__==:_==_-__:_:__:_::_. 3.8 4.0 0 0 Ability to explain and apply all system limits and Conduct of operations G2.1.32 G2.1.40 Conduct of operations

---

G2.2.40 Equipment Control G2.2.6 Equipment Control -_._-------

G2.3.12 Radiation Control G2.3.5 Radiation Control G2.3.7 Radiation Control G2.4.28 Emergency Procedures/Plans G2.4.46 Emergency Procedures/Plans 2.8 3.9 o 3.4 4.7 0 0 0 0 0 0 0 0 0 0 3.0 3.6 0 0 3.2 3.7 precautions. Knowledge of refueling administrative requirements Ability to apply technical specifications for a system. Knowledge of the process for making changes to procedures

---

_._-Knowledge of radiological safety principles pertaining to licensed operator duties o 0 0 0 0 0 0 Ability to use radiation monitoring systems 2.9 2.9 3.5 3.6 DOD 0 Ability to comply with radiation work permit requirements during normal or abnormal conditions 3.2 4.1 0 0 0 0 Knowledge of procedures relating to emergency 4.2 4.2 Page 1 of 1 response to sabotage.

to verify that the alarms are consistent with the plant conditions. 6/25/2009 12:27 PM ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 EmerQencv and Abnormal Plant Evolutions -Tier 1/Group 1 SRil E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor p O()7 13A ;{, Trip -Stabilization -Recovery 1 1 000008 Pressurizer Vapor Space Accident 1 3 000009 Small Break LOCA 1 3 S 001 4.;;;, 143 000011 Large Break LOCA 13 1$ 011 GA;),01 13*4 000015/17 RCP Malfunctions 1 4 000022 Loss of Rx Coolant Makeup 1 2 000025 Loss of RHR System 14 000026 Loss of Component Cooling ;5 l3.b Water 18 000027 Pressurizer Pressure Control System Malfunction 1 3 000029 A TWS 1 1 000038 Steam Gen. Tube Rupture 1 3 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture -Excessive Heat Transfer 14 000054 (CE/E06) Loss of Main Feedwater 1 4 000055 Station Blackout 1 6 000056 Loss of Off-site Power 1 6 S OSC? AG.;;2, I. 3,g 000057 Loss of Vital AC Inst. Bus 1 6 000058 Loss of DC Power 1 6 000062 Loss of Nuclear Svc Water 1 4 000065 Loss of Instrument Air 18 W/E04 LOCA Outside Containment 1 3 W/E11 Loss of Emergency Coolant Recirc. 1 4 BW/E04; W/E05 Inadequate Heat Transfer -Loss of Secondary Heat Sink 1 4 000077 Generator Voltage and Electric 1)17 A fl. , 4.4-Grid Disturbances 1 6 KIA Category Totals: .3 13 Group Point Total: 18/6 ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 EmerQencv and Abnormal Plant Evolutions -Tier 1/Group 1 SRil E/APE # 1 Name 1 Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000007 (BW/E02&E10; CE/E02) Reactor p O()7 13A ;{, Trip -Stabilization -Recovery 1 1 000008 Pressurizer Vapor Space Accident 1 3 000009 Small Break LOCA 1 3 S 001 4.;;;, 143 000011 Large Break LOCA 13 1$ 011 GA;),01 13*4 000015/17 RCP Malfunctions 1 4 000022 Loss of Rx Coolant Makeup 1 2 000025 Loss of RHR System 14 000026 Loss of Component Cooling ;5 l3.b Water 18 000027 Pressurizer Pressure Control System Malfunction 1 3 000029 A TWS 1 1 000038 Steam Gen. Tube Rupture 1 3 000040 (BW/E05; CE/E05; W/E12) Steam Line Rupture -Excessive Heat Transfer 14 000054 (CE/E06) Loss of Main Feedwater 1 4 000055 Station Blackout 1 6 000056 Loss of Off-site Power 1 6 S OSC? AG.;;2, I. 3,g 000057 Loss of Vital AC Inst. Bus 1 6 000058 Loss of DC Power 1 6 000062 Loss of Nuclear Svc Water 1 4 000065 Loss of Instrument Air 18 W/E04 LOCA Outside Containment 1 3 W/E11 Loss of Emergency Coolant Recirc. 1 4 BW/E04; W/E05 Inadequate Heat Transfer -Loss of Secondary Heat Sink 1 4 000077 Generator Voltage and Electric 1)17 A fl. , 4.4-Grid Disturbances 1 6 KIA Category Totals: .3 13 Group Point Total: 18/6 ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 007EA2.04 Reactor Trip -Stabilization -Recovery 4.6 4.4 1 1 o 0 009EG2.4.20 Small Break LOCA 1 3 3.8 4.3 0 DOD 011EA2.07 Large Break LOCA / 3 3.2 3.4 026AA2.02 Loss of Component Cooling Water 18 2.9 3.6 0 0 0 0 0 0 0 0 0 0 056AG2.1.19 Loss of Off-site Power 1 6 077AG2.2.44 Generator Voltage and Electric Grid Disturbances 1 6 3.9 3.8 4.2 4.4 0 0 0 0 0 0 0 0 Page 1 of 1 FORM ES-401-2 TOPIC: If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP Knowledge of operational implications of EOP warnings, cautions and notes. That equipment necessary for functioning of critical pump water seals is operable The cause of possible CCW loss Ability to use plant computer to evaluate system or component status. Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 6/25/2009 12:27 PM ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 007EA2.04 Reactor Trip -Stabilization -Recovery 1 1 4.6 4.4 DO 009EG2.4.20 Small Break LOCA 1 3 3.8 4.3 0 011EA2.07 Large Break LOCA / 3 3.2 3.4 026AA2.02 Loss of Component Cooling Water 18 2.9 3.6 0 0 0 0 0 0 0 0 0 0 056AG2.1.19 Loss of Off-site Power 1 6 3.9 3.8 ------"----- 077AG2.2.44 Generator Voltage and Electric Grid Disturbances 1 6 4.2 4.4 0 0 0 0 0 0 0 0 Page 1 of 1 FORM ES-401-2 TOPIC: If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP Knowledge of operational implications of EOP warnings, cautions and notes. That equipment necessary for functioning of critical pump water seals is operable The cause of possible CCW loss Ability to use plant computer to evaluate system or component status. Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 6/25/2009 12:27 PM ES-401 3 ES-401 PWR Examination Outline Form E'" fI)" ,., Emerqency and Abnormal Plant Evolutions -Tier 1/Group 2 fRO"/ SRO E/APE # / Name / Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 t DD :; +. 3S 11. 0 000024 Emergency Boration / 1 5 o f21 AAo? 0 I 4.1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI /7 033 AA,;:?, 0'3> 3,/ 000036 (BW/A08) Fuel Handling Accident /8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum /4 000059 Accidental Liquid RadWaste ReI. / 9 000060 Accidental Gaseous Radwaste ReI. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W /E 14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 I's f}/1 EG r:2;), 41-000076 Hiqh Reactor Coolant Activity / 9 W/E01 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation /9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip /4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin /4 BW/E08; W/E03 LOCA Cooldown -Depress. /4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling -PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery t:I:E r-. fA r'. "'.' T. Total: 9/ ES-401 3 ES-401 PWR Examination Outline Form E'" fI)" ,., Emerqency and Abnormal Plant Evolutions -Tier 1/Group 2 fRO"/ SRO E/APE # / Name / Safety Function K K K A A G KIA Topic(s) IR # 1 2 3 1 2 000001 Continuous Rod Withdrawal / 1 000003 Dropped Control Rod / 1 000005 Inoperable/Stuck Control Rod / 1 t DD :; +. 3S 11. 0 000024 Emergency Boration / 1 5 o f21 AAo? 0 I 4.1 000028 Pressurizer Level Malfunction / 2 000032 Loss of Source Range NI / 7 000033 Loss of Intermediate Range NI /7 033 AA,;:?, 0'3> 3,/ 000036 (BW/A08) Fuel Handling Accident /8 000037 Steam Generator Tube Leak / 3 000051 Loss of Condenser Vacuum /4 000059 Accidental Liquid RadWaste ReI. / 9 000060 Accidental Gaseous Radwaste ReI. / 9 000061 ARM System Alarms / 7 000067 Plant Fire On-site / 8 000068 (BW/A06) Control Room Evac. / 8 000069 (W /E 14) Loss of CTMT Integrity / 5 000074 (W/E06&E07) Inad. Core Cooling / 4 I's f}/1 EG r:2;), 41-000076 Hiqh Reactor Coolant Activity / 9 W/E01 & E02 Rediagnosis & SI Termination / 3 W/E13 Steam Generator Over-pressure / 4 W/E15 Containment Flooding / 5 W/E16 High Containment Radiation /9 BW/A01 Plant Runback / 1 BW/A02&A03 Loss of NNI-X/Y / 7 BW/A04 Turbine Trip /4 BW/A05 Emergency Diesel Actuation / 6 BW/A07 Flooding / 8 BW/E03 Inadequate Subcooling Margin /4 BW/E08; W/E03 LOCA Cooldown -Depress. /4 BW/E09; CE/A13; W/E09&E10 Natural Circ. / 4 BW/E13&E14 EOP Rules and Enclosures CE/A11; W/E08 RCS Overcooling -PTS / 4 CE/A16 Excess RCS Leakage / 2 CE/E09 Functional Recovery t:I:E r-. fA r'. "'.' T. Total: 9/ ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 005AG2.4.35 Inoperable/Stuck Control Rod 1 1 3.8 4.0 0 000000 024AA2.01 Emergency Boration 1 1 3.8 4.1 0 0 0 0 0 033AA2.03 Loss of Intermediate Range NI 17 2.8 3.1 0 0 0 0 074EG2.2.44 Inad. Core Cooling 1 4 4.2 4.4 Page 1 of 1 FORM ES-401-2 TOPIC: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects Whether boron flow and/or MOVs are malfunctioning from plant conditions Indication of blown fuse Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions 6/25/2009 12:27 PM ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 005AG2.4.35 Inoperable/Stuck Control Rod 1 1 3.8 4.0 0 0 0 0 0 0 0 0 FORM ES-401-2 TOPIC: Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects ---------------------------- 024AA2.01 Emergency Boration 1 1 033AA2.03 Loss of Intermediate Range NI 17 --074EG2.2.44 Inad. Core Cooling 1 4 3.8 4.1 DOD 0 ------------------ 2.8 3.1 0 0 0 0 0 Whether boron flow and/or MOVs are malfunctioning from plant conditions Indication of blown fuse ---------------------

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4.2 4.4 0 Ability to interpret control room indications to verify the Page 1 of 1 status and operation of a system, and understand how operator actions and directives affect plant and system conditions 6/25/2009 12:27 PM ES*401 4 ES-401 PWR Examination Form ES-401-2 Plant S stems -Tier 2/Group 1 fl SRO System # / Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal Is I:> G; 1,7 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection

1 DJ;2 6ii.;<, 4. // 14,2 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam lS 03J A;(,OiiL 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator I¢ 064 A 3,1 073 Process Radiation 073 !1 g,2 Monitoring , 076 Service Water 078 Instrument Air 103 Containment KIA Category Point Totals

3 :L Group Point Total: I 28/5 ES*401 4 ES-401 PWR Examination Form ES-401-2 Plant S stems -Tier 2/Group 1 fl SRO System # / Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 003 Reactor Coolant Pump 004 Chemical and Volume Control 005 Residual Heat Removal Is I:> G; 1,7 006 Emergency Core Cooling 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection

1 DJ;2 6ii.;<, 4. // 14,2 013 Engineered Safety Features Actuation 022 Containment Cooling 025 Ice Condenser 026 Containment Spray 039 Main and Reheat Steam lS 03J A;(,OiiL 059 Main Feedwater 061 Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator I¢ 064 A 3,1 073 Process Radiation 073 !1 g,2 Monitoring , 076 Service Water 078 Instrument Air 103 Containment KIA Category Point Totals

:3 :L Group Point Total: I 28/5 ES-401, REV 9 KA NAME / SAFETY FUNCTION:

005G2.2.40 Residual Heat Removal 012G2.4.11 Reactor Protection 039A2.02 Main and Reheat Steam 064A2.16 Emergency Diesel Generator 073A2.02 Process Radiation Monitoring SRO T2G1 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.4 4.7 0 0 0 0 0 0 0 0 4.0 4.2 0 0 0 0 0 0 0 0 0 2.4 2.7 0 0 0 3.3 3.7 2.7 3.2 0 0 0 0 0 0 0 0 Page 1 of 1 FORM ES-401-2 TOPIC: Ability to apply technical specifications for a system. Knowledge of abnormal condition procedures. Decrease in turbine load as it relates to steam escaping from relief valves Loss of offsite power during full-load testing of ED/G Detector failure 6/25/2009 12:27 PM ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC: RO SRO 005G2.2.40 Residual Heat Removal 3.4 4.7 0 0 Ability to apply technical specifications for a system. -:c----__ 012G2.4.11 Reactor Protection 4.0 4.2 0 0 0 0 0 0 0 0 0 Knowledge of abnormal condition procedures. 039A2.02 Main and Reheat Steam 2.4 2.7 0 0 0 0 0

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_.

064A2.16 Emergency Diesel Generator 3.3 3.7 0 0 0 0 0 0 073A2.02 Process Radiation Monitoring 2.7 3.2 00000 Page 1 of 1 Decrease in turbine load as it relates to steam escaping from relief valves Loss of offsite power during full-load testing of ED/G Detector failure 6/25/2009 12:27 PM ES-401 5 Form ES-401-2 PWR Examination F Plant S stems -Tier 2/Group 2 SRO System # 1 Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner 1$ I 40 and Purge Control -029 Containment Purge 033 Spent Fuel Pool Cooling 034 Fuel Handling Equipment 035 Steam Generator 041 Steam DumpITurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal ,5 OS)' GQd,'JO 3,2 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water IrS' . 07t; Ad, D/ 3.';;{ 079 Station Air 086 Fire Protection I Lt' UI n ,int Totals: l.:l I p Point Total: 10/3 ES-401 5 Form ES-401-2 PWR Examination Cv-'"tu Plant S stems -Tier 2/Group 2 SRO System # 1 Name K K K K K K A A A A G KIA Topic(s) IR # 1 2 3 4 5 6 1 2 3 4 001 Control Rod Drive 002 Reactor Coolant 011 Pressurizer Level Control 014 Rod Position Indication 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation 017 In-core Temperature Monitor 027 Containment Iodine Removal 028 Hydrogen Recombiner $ I 40 and Purge Control -029 Containment Purge 033 Spent Fuel Pool Coolinq 034 Fuel Handling Equipment 035 Steam Generator 041 Steam DumpITurbine Bypass Control 045 Main Turbine Generator 055 Condenser Air Removal ,5 OS)' 3,2 056 Condensate 068 Liquid Radwaste 071 Waste Gas Disposal 072 Area Radiation Monitoring 075 Circulating Water IrS' . 07t; Ad, D/ 3.)" 079 Station Air 086 Fire Protection fA . .. tals: l.:l I Group Point Total: 10/3 ES-401, REV 9 KA NAME / SAFETY FUNCTION: 028A2.03 Hydrogen Recombiner and Purge Control 055G2.1.20 Condenser Air Removal 075A2.01 Circulating SRO T2G2 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.4 4.0 000 4.6 4.6 0 0 0 0 0 0 3.0 3.2 0 0 0 0 0 0 Page 1 of 1 FORM ES-401-2 TOPIC: The hydrogen air concentration in excess of limit flame propagation or detonation with resulting equipment damage in containment Ability to execute procedure steps. Loss of intake structure 6/25/2009 12:27 PM ES-401, REV 9 KA NAME / SAFETY FUNCTION: 028A2.03 Hydrogen Recombiner and Purge Control SRO T2G2 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.4 4.0 FORM ES-401-2 TOPIC: The hydrogen air concentration in excess of limit flame propagation or detonation with resulting equipment damage in containment 055G2.1.20 Condenser Air Removal 4.6 4.6 D D D D D D Ability to execute procedure steps. 075A2.01 Circulating 3.0 3.2 D D D D D D intake structure Page 1 of 1 6/25/2009 12:27 PM ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 I Facility: SQ,,! (Sf2-D £Kt{lY') Date of Exam: Feb ;(01 Category KIA # Topic RO SRO-Only IR # IR # 2.1.;2& !'VO (fcA'ZNldJ! efl1,) AJ/Jt 36 1. 2.1.;<1 >tj(} )e*"" liYlO [ III [;5 . j-J:.rs, SW \ 4,0 v I Conduct 2.1. of Operations 2.1. 2.1. 2.1. Subtotal I . ... ... 2.2. 102 5t1 ItVvic..c!.. I JrTU.d/.,,, ddl 4-,/ 2.2.11 Pl 1'"i'!t7/V'Lt:'-i-'7f'; Y;q P. "" *Ire 6,% v 2. 2.2. Equipment

2.2. Control

2.2. 2.2. Subtotal . . .........

L ...... 2.3.5 (SC ff 0{,1 'I>'--' 2.3. 3. 2.3. Radiation Control 2.3. 2.3. 2.3. Subtotal I',' .; . .1 2.4. b t7t/}7V. p-ttJ\"h In" ,§ ji pt?f ,rl-'Yl.5 ( /tfi",j'c r

' rlYtiI#f!J/t 4,7 2.4. 1;2 0w Y(b.1 J i}y'(iil/Jrit'1h dAl-I1;'/&!

Ip

• ,-3 4. f if U Emergency

2.4. Procedures

I Plan 2.4. 2.4. ,l; 2.4. Ai/fir V Subtotal foIJ;; ";;;L Tier 3 Point Total ... ' .......*...

• • • • I.' 10 I *. ***... 7 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 I SON (Sf2-D {'"raM) Date of Exam: Feb ;(olD I Category KIA # Topic RO SRO-Onlv IR # IR # 2.1.;2&

jJvo Pt-d.A ("'c/fZE'.?lr1 I AJA 36 1. 2.1.;<1 >t;(} )e*"" l:'YlI7. [ III [;5 . j-J:.rs, SW \ 4,0 v I Conduct 2.1. of Operations 2.1. 2.1. 2.1. Subtotal I*** '. . ....* .. 2.2. 102 5t1 ,rvedltVv1(..e. jiJrTUdt.,/j v-<L 4-,/ 2.2.11 pl ,,"f'!t7A/' Cl"?f'; .P. "'" 'ee 6,% {/ 2. 2.2. Equipment

2.2. Control

2.2. 2.2. Subtotal ........ .:;;L. 2.3.5 lAse fI-0{,1 '"," 2.3. 3. 2.3. Radiation Control 2.3. 2.3. 2.3. Subtotal I .i. '.; .1 2.4. b r;;?}7V,rHtJ\*% /",L10/-! pt?f '}'Yl.5 ffp1jA:f*0 r:' 4,7 2.4. 1;2 0-v/ Y (b.IJI}y'(jil/ )rft'ih u/4.1 ;"11 /f. .,1 '14 ;/ "'3 4. if U Emergency

2.4. Procedures

I Plan 2.4. 2.4. ,II 2.4. Ai/fit 'V Subtotal "f{ C. foil;; Tier 3 Point Total ./'7 10 *. .**... 7 ES-401, REV 9 KA NAME / SAFETY FUNCTION: G2.1.26 Conduct of operations G2.1.29 Conduct of operations G2.2.12 Equipment Control G2.2.17 Equipment Control G2.3.5 Radiation Control G2.4.12 Emergency Procedures/Plans G2.4.6 Emergency Procedures/Plans SRO T3 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.4 3.6 0 0 0 0 0 0 4.1 4.0 0 0 0 0 0 0 0 0 3.7 4.1 0 0 0 0 2.6 3.8 0 0 2.9 2.9 4.0 4.3 0 0 0 0 0 0 0 0 0 0 3.7 4.7 Page 1 of 1 FORM ES-401-2 TOPIC: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. Knowledge of surveillance procedures. Knowledge of the process for managing maintenance activities during power operations. Ability to use radiation monitoring systems Knowledge of general operating crew responsibilities during emergency operations. Knowledge symptom based EOP mitigation strategies. 6/25/2009 12:27 PM ES-401, REV 9 KA NAME / SAFETY FUNCTION: G2.1.26 Conduct of operations G2.1.29 Conduct of operations G2.2.12 Equipment Control G2.2.17 Equipment Control SRO T3 PWR EXAMINATION OUTLINE IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G RO SRO 3.4 3.6 0 0 0 0 0 0 4.1 4.0 00000 3.7 4.1 0000 2.6 3.8 0 FORM ES-401-2 TOPIC: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc. Knowledge of surveillance procedures. Knowledge of the process for managing maintenance activities during power operations. G2.3.5 Radiation Control 2.9 2.9 Ability to use radiation monitoring systems G2.4.12 Emergency Procedures/Plans 4.0 4.3 0 0 0 0 0 0 0 0 0 0 G2.4.6 Emergency Procedures/Plans 3.7 4.7 Page 1 of 1 Knowledge of general operating crew responsibilities during emergency operations. Knowledge symptom based EOP mitigation strategies. 6/25/2009 12:27 PM ES*301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah 1 & 2 Date of Examination: 2/16/2010 Examination Level: RO D SRO X Operating Test Number: 2010301 Administrative Topic Type Describe activity to be performed (see Note) Code* N,R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. (3.9/4.2) Boric Acid storage tank level calculation (JPM #2.1.a SRO 09-AP) D,R 2.1.7 Ability to evaluate plant performance and make Conduct of Operations operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (4.4/4.7) Determine Mn. Turb. Startup and Loading Times (JPM2.1.b) D,R 2.2.12 Knowledge of surveillance procedures Equipment Control Review a Surveillance for approval (3.7/4.1) (JPM410) D,R,P 2.3.6 Ability to approve release permits Radiation Control Approval of a Waste Gas Decay tank Release (2.0/3.8) (JPM A-3) D,R 2.4.38 Ability to take actions called for in the facility Emergency emergency plan, including supporting or acting as Procedures/Plan emergency coordinator if required. (2.4/4.4) Classify Rep: LOCA with Fuel Fail Loss of Cntmt (JPM 019 AP2 ) 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 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah 1 & 2 Date of Examination: 2/16/2010 Examination Level: RO D SRO X Operating Test Number: 2010301 Administrative Topic Type Describe activity to be performed (see Note) Code* N,R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. (3.9/4.2) Boric Acid storage tank level calculation (JPM #2.1.a SRO 09-AP) D,R 2.1.7 Ability to evaluate plant performance and make Conduct of Operations operational judgments based on operating characteristics, reactor behavior, and instrument interpretation (4.4/4.7) Determine Mn. Turb. Startup and Loading Times (JPM2.1.b) D,R 2.2.12 Knowledge of surveillance procedures Equipment Control Review a Surveillance for approval (3.7/4.1) (JPM410) D,R,P 2.3.6 Ability to approve release permits Radiation Control Approval of a Waste Gas Decay tank Release (2.0/3.8) (JPM A-3) D,R 2.4.38 Ability to take actions called for in the facility Emergency emergency plan, including supporting or acting as Procedures/Plan emergency coordinator if required. (2.4/4.4) Classify Rep: LOCA with Fuel Fail Loss of Cntmt (JPM 019 AP2 ) 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)

SRO 2.1.a This JPM has the candidate if the "C" boric acid storage tank, which is in standby is operable such that it can be aligned to Unit 1 to replace the "A" BAST which is being removed from service. The candidate will determine that "C" BAST is not operable and makes recommendation to make tank operable before allowing "A" to be removed from service. This is a New JPM that can be performed in the simulator or in the classroom. SRO 2.1.b This JPM has the candidate determine the Main Turbine Startup and Loading times for Unit 1 turbine from cold shutdown conditions during a Unit startup. This is a Bank JPM that can be performed in the simulator or classroom. SRO 2.2 This JPM has the candidate review a surveillance test for determining whether or not the adjust RCP seal injection supply controlled leakage. This is a Bank JPM that can be performed on the simulator or in the classroom. SRO 2.3 This JPM has the candidate review a radioactive gas decay tank release to determine if release permit is accurate and can take place as written. The candidate will determine who needs to approve the permit and actions needed due to having a radiation monitor out of service. This is a Bank JPM that can be performed in the simulator or in the classroom. This JPM is a Repeat from the 2009 NRC exam and was randomly selected from a group of Radiation Control JPMs. SRO 2.4 This JPM has the candidate determine the correct Emergency Classification based on the data provided and make the initial contacts. This is a time critical, Bank JPM that can be performed in the simulator or in the classroom. SRO 2.1.a This JPM has the candidate if the "C" boric acid storage tank, which is in standby is operable such that it can be aligned to Unit 1 to replace the "A" BAST which is being removed from service. The candidate will determine that "C" BAST is not operable and makes recommendation to make tank operable before allowing "A" to be removed from service. This is a New JPM that can be performed in the simulator or in the classroom. SRO 2.1.b This JPM has the candidate determine the Main Turbine Startup and Loading times for Unit 1 turbine from cold shutdown conditions during a Unit startup. This is a Bank JPM that can be performed in the simulator or classroom. SRO 2.2 This JPM has the candidate review a surveillance test for determining whether or not the adjust RCP seal injection supply controlled leakage. This is a Bank JPM that can be performed on the simulator or in the classroom. SRO 2.3 This JPM has the candidate review a radioactive gas decay tank release to determine if release permit is accurate and can take place as written. The candidate will determine who needs to approve the permit and actions needed due to having a radiation monitor out of service. This is a Bank JPM that can be performed in the simulator or in the classroom. This JPM is a Repeat from the 2009 NRC exam and was randomly selected from a group of Radiation Control JPMs. SRO 2.4 This JPM has the candidate determine the correct Emergency Classification based on the data provided and make the initial contacts. This is a time critical, Bank JPM that can be performed in the simulator or in the classroom. ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah 1 & 2 Date of Examination: 2/16/2010 Examination Level: RO X SRO D Operating Test Number: 2010301 Administrative Topic Type Describe activity to be performed (see Note) Code* N, R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. (3.9/4.2) Boric Acid storage tank level calculation (JPM 2.1.a RO) D,R 2.1.26 Knowledge of industrial safety procedures (such Conduct of Operations as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen. (3.4/3.6) Containment Formaldehyde Stay Time Calculation (JPM 123) D,S 2.2.12 Knowledge of surveillance procedures (3.7/4.1) Equipment Control Perform Reactor Coolant System water inventory (JPM 43-1) Radiation Control N,R 2.4.13 Knowledge of crew roles and responsibilities Emergency during EOP usage. (4.0/4.6) Procedures/Plan Calculating maximum reactor vessel vent time Per EA-0-7 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 (2: 1) (P)revious 2 exams (:::; 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah 1 & 2 Date of Examination: 2/16/2010 Examination Level: RO X SRO D Operating Test Number: 2010301 Administrative Topic Type Describe activity to be performed (see Note) Code* N, R 2.1.25 Ability to interpret reference materials, such as Conduct of Operations graphs, curves, tables, etc. (3.9/4.2) Boric Acid storage tank level calculation (JPM 2.1.a RO) D,R 2.1.26 Knowledge of industrial safety procedures (such Conduct of Operations as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen. (3.4/3.6) Containment Formaldehyde Stay Time Calculation (JPM 123) D,S 2.2.12 Knowledge of surveillance procedures (3.7/4.1) Equipment Control Perform Reactor Coolant System water inventory (JPM 43-1) Radiation Control N,R 2.4.13 Knowledge of crew roles and responsibilities Emergency during EOP usage. (4.0/4.6) Procedures/Plan Calculating maximum reactor vessel vent time Per EA-0-7 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 (2: 1) (P)revious 2 exams (:::; 1; randomly selected)

RO 2.1.a This JPM that has the candidate determine if "C" Boric Acid storage tank can be aligned to Unit 1 so that "A" Boric Acid storage tank can be removed from service. This is a New JPM can be performed in the simulator or classroom. RO 2.1.b This JPM has the candidate determine the stay time in containment based on the formaldehyde concentration and determine the respiratory protection requirements. This is a Bank JPM that can be performed in the simulator or classroom. RO 2.2 This JPM has the candidate determine the Reactor Coolant System Water inventory (leak rate). This is a Bank JPM that should be performed in the simulator (using plant computer). RO 2.4 This JPM has the candidate determine the allowable Reactor Vessel head venting time to prevent CNMT Hydrogen from exceeding 3% per EA-O-7. This is a New JPM that can be performed in the classroom. RO 2.1.a This JPM that has the candidate determine if "C" Boric Acid storage tank can be aligned to Unit 1 so that "A" Boric Acid storage tank can be removed from service. This is a New JPM can be performed in the simulator or classroom. RO 2.1.b This JPM has the candidate determine the stay time in containment based on the formaldehyde concentration and determine the respiratory protection requirements. This is a Bank JPM that can be performed in the simulator or classroom. RO 2.2 This JPM has the candidate determine the Reactor Coolant System Water inventory (leak rate). This is a Bank JPM that should be performed in the simulator (using plant computer). RO 2.4 This JPM has the candidate determine the allowable Reactor Vessel head venting time to prevent CNMT Hydrogen from exceeding 3% per EA-O-7. This is a New JPM that can be performed in the classroom. ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: __ Sequoyah Nuclear Station Date of Examination: _02/16/2010 -Exam Level: RO X SRo-ID SRO-U 0 Operating Test No.: 2010301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function a. 005 Inoperable/Stuck Control Rod (AA2.03 3.5/4.4) D, S, A, L 1 001 AP-1 Emergency 80ration (Stuck Rods) b. 011 Large 8reak LOCA (EA 1.11 4.2/4.2) D, S, A, EN 2 013AP1 Transfer to Hot Leg Recirc. c. 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve M,S,A 3 Stuck Open) (AA2.19 3.4/3.6) 070 AP Respond to Failed Open PZR Spray valve d. 005 Residual Heat Removal System (A4.01 3.6/3.4) D,S,L 4P 152 Swap RHR pumps (A train to 8 train) with level in the PZR. e. WE05 Loss of Secondary Heat Sink (EA2.1 3.4/4.4) N, S 4S 034-1 Establish MFW per EA-2-2 f. 103 Containment System (A1.01 3.7/4.1) D, S 5 065 Re-establishment of CNMT pressure g. 064 Emergency Diesel Generators (A4.06 3.9/3.9) D,S,A 6 077-1AP Perform DG load test on 18-8 DG (with high crankcase pressure)

h. 015 Nuclear Instrumentation System (A2.02 3.1/3.5) M,S, 7 021 AP Respond to a failure of N-41 In-Plant Systems(gl (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 078 Instrument Air System D,E 8 096 Respond to Loss of Control Air System j. 005 Residual Heat Removal System D,E,R 4P 044 Venting A-A RHR pump due to cavitation

k. 062 AC Electrical Distribution System D,E,A 6 091AP Transfer Controls to Aux Mode per AOP-C.04, Att 3 All RO and SRO-I control room (and in-plant) systems must be different and seNe different safety functions; all 5 SRO-U systems must seNe different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes I Criteria for RO I SRO-II SRO-U ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:

__ Sequoyah Nuclear Station Date of Examination: _02/16/2010 -Exam Level: RO X SRo-ID SRO-U 0 Operating Test No.: 2010301 Control Room Systemsfg/ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function a. 005 Inoperable/Stuck Control Rod (AA2.03 3.5/4.4) D, S, A, L 1 001 AP-1 Emergency 80ration (Stuck Rods) b. 011 Large 8reak LOCA (EA 1.11 4.2/4.2) D, S, A, EN 2 013AP1 Transfer to Hot Leg Recirc. c. 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve M,S,A 3 Stuck Open) (AA2.19 3.4/3.6) 070 AP Respond to Failed Open PZR Spray valve d. 005 Residual Heat Removal System (A4.01 3.6/3.4) D,S,L 4P 152 Swap RHR pumps (A train to 8 train) with level in the PZR. e. WE05 Loss of Secondary Heat Sink (EA2.1 3.4/4.4) N, S 4S 034-1 Establish MFW per EA-2-2 f. 103 Containment System (A1.01 3.7/4.1) D, S 5 065 Re-establishment of CNMT pressure g. 064 Emergency Diesel Generators (A4.06 3.9/3.9) D,S,A 6 077-1AP Perform DG load test on 18-8 DG (with high crankcase pressure)

h. 015 Nuclear Instrumentation System (A2.02 3.1/3.5) M,S, 7 021 AP Respond to a failure of N-41 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 078 Instrument Air System D,E 8 096 Respond to Loss of Control Air System j. 005 Residual Heat Removal System D,E,R 4P 044 Venting A-A RHR pump due to cavitation

k. 062 AC Electrical Distribution System D,E,A 6 091AP Transfer Controls to Aux Mode per AOP-C.04, Att 3 All RO and SRO-I control room (and in-plant) systems must be different and seNe different safety functions; all 5 SRO-U systems must seNe different safety functions; in-plant systems and functions may overlap those tested in the control room.

• Type Codes I Criteria for RO I SRO-II SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6/4-6 / 2-3 s9/s8/s4 ;::1/;::1/;::1 -/ -/ ;::1 (control room system) ;::1/;::1/;::1

2/;

:2/;::1 s 3 / s 3 / s 2 (randomly selected)

1/;::1/;::1 JPM A. Candidate has to determine that emergency boration is required due to 2 stuck rods following a Rx trip, with the normal emergency boration valve failing to open, requiring the candidate to align alternate boration through the charging pump suction. This is a Bank, Low Power, Alternate path, JPM. JPM B. Candidate is directed to make alignment change for Transfer to Hot Leg Recirculation per ES-1.4. following a trip due to large break LOCA. The Hot leg recirculation valve will fail to open requiring RHR to be aligned to Cold Leg i1\iection with High head pumps aligned to Hot Legs. This is a Bank, Low Power, Alternate path, JPM. JPM C. Candidate will be required to respond to failed PZR pressure instrument which causes the PZR spray valves to with one valve sticking open. With a PZR spray valve failing to close, candidate will follow actions of AOP-1.4 bnd trip the RX and trip at least two RCPs to stop mitigate the depressurization to prevent SI actuation.

This is a-Modified Bank, Alternate Path JPM. Original JPM JPM D. Plant is in Mode 4, and Candidate is directed to transfer RHR pumps from B train to A train. This is a Bank, Low Power JPM. JPM E. Candidate is directed to establish a Secondary Heat Sink using Main Feed Water System following a Rx Trip. MFW will be required due to a failure of all AFW pumps. This is a New JPM. JPM F. Candidate is directed to vent excess pressure from CNMT and then establish Ventilation alignment to re-establish CNMT vacuum atmosphere. This is a Bank JPM. JPM G. Candidate is to perform a quick start ofEDG A-A and load the EDG. While the candidate is loading the EDG, a high crankcase condition will develop requiring a manual emergency trip of the EDG. This is a Bank, Alternate Path JPM. JPM H. Candidate will respond to failed Nuclear Instrument (N41) High. Control Rods will be stepping in at maximum rate, Candidate will take Rod bank selector switch to Manual and proceed to remove failed channel from service. This is a Modified Bank, JPM. Original JPM (021) had N-41 failing low from -45%, this JPM has N-41 failing High, requiring immediate manual action to stop control rod movement prior to removing channel from service. JPM 1. Candidate is to restore Control Air pressure following a Loss of Off-site power. Candidate will restore Control and Service Air compressors using EA-32-r. This will require a Local Manual start. This a Bank JPM. JPM J. Candidate is directed to vent the lA-A RHR pump due to pump cavitating during mid-loop operation. This venting is done locally and is required to be performed to return the RHR pump to service. This is a Bank JPM and is performed in the RCA. JPM K. Candidate is directed to perform checklist 3 of AOP-C.04 Shutdown from Auxiliary Control Room, following an event which requires Control Room Abandonment. This a Bank, Alternate Path JPM. (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6/4-6 / 2-3 s9/s8/s4 ;::1/;::1/;::1 -/ -/ ;::1 (control room system) ;::1/;::1/;::1

2/;

:2/;::1 s 3 / s 3 / s 2 (randomly selected)

1/;::1/;::1 JPM A. Candidate has to determine that emergency boration is required due to 2 stuck rods following a Rx trip, with the normal emergency boration valve failing to open, requiring the candidate to align alternate boration through the charging pump suction. This is a Bank, Low Power, Alternate path, JPM. JPM B. Candidate is directed to make alignment change for Transfer to Hot Leg Recirculation per ES-1.4. following a trip due to large break LOCA. The Hot leg recirculation valve will fail to open requiring RHR to be aligned to Cold Leg i1\iection with High head pumps aligned to Hot Legs. This is a Bank, Low Power, Alternate path, JPM. JPM C. Candidate will be required to respond to failed PZR pressure instrument which causes the PZR spray valves to with one valve sticking open. With a PZR spray valve failing to close, candidate will follow actions of AOP-1.4 bnd trip the RX and trip at least two RCPs to stop mitigate the depressurization to prevent SI actuation.

This is a-Modified Bank, Alternate Path JPM. Original JPM JPM D. Plant is in Mode 4, and Candidate is directed to transfer RHR pumps from B train to A train. This is a Bank, Low Power JPM. JPM E. Candidate is directed to establish a Secondary Heat Sink using Main Feed Water System following a Rx Trip. MFW will be required due to a failure of all AFW pumps. This is a New JPM. JPM F. Candidate is directed to vent excess pressure from CNMT and then establish Ventilation alignment to re-establish CNMT vacuum atmosphere. This is a Bank JPM. JPM G. Candidate is to perform a quick start ofEDG A-A and load the EDG. While the candidate is loading the EDG, a high crankcase condition will develop requiring a manual emergency trip of the EDG. This is a Bank, Alternate Path JPM. JPM H. Candidate will respond to failed Nuclear Instrument (N41) High. Control Rods will be stepping in at maximum rate, Candidate will take Rod bank selector switch to Manual and proceed to remove failed channel from service. This is a Modified Bank, JPM. Original JPM (021) had N-41 failing low from -45%, this JPM has N-41 failing High, requiring immediate manual action to stop control rod movement prior to removing channel from service. JPM 1. Candidate is to restore Control Air pressure following a Loss of Off-site power. Candidate will restore Control and Service Air compressors using EA-32-r. This will require a Local Manual start. This a Bank JPM. JPM J. Candidate is directed to vent the lA-A RHR pump due to pump cavitating during mid-loop operation. This venting is done locally and is required to be performed to return the RHR pump to service. This is a Bank JPM and is performed in the RCA. JPM K. Candidate is directed to perform checklist 3 of AOP-C.04 Shutdown from Auxiliary Control Room, following an event which requires Control Room Abandonment. This a Bank, Alternate Path JPM. ES*301 Control Room/ln*Plant Systems Outline Form ES*301*2 Facility: Seauovah Nuclear Plant Date of Examination: _02/16/2010 -Exam Level: RO D SRO-I X SRO-U D Operating Test No.: 2010301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function a. 005 Inoperable/Stuck Control Rod (AA2.03 3.5/4.4) D, S,A,L 1 001 AP-1 Emergency Boration (Stuck Rods) b. 011 Large Break LOCA (EA1.11 4.2/4.2) D, S,A, EN 2 013AP1 Transfer to Hot Leg Recirc. c. 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve N,S,A 3* Stuck Open) (AA2.19 3.4/3.6) 070 AP Respond to Failed Open PZR Spray valve d. 005 Residual Heat Removal System (A4.01 3.6/3.4) D, S,L 4P . 152 Swap RHR pumps (A train to B train) with level in the PZR. e. WE05 Loss of Secondary Heat Sink (EA2.1 3.4/4.4) N,S 4S 034-1 Establish MFW per EA-2-2 f. g. 064 Emergency Diesel Generators (A4.06 3.9/3.9) D,S,A 6 077 -1 AP Perform DG load test on 1 B-B DG (with high crankcase pressure)

h. 015 Nuclear Instrumentation System (A2.02 3.1/3.5) M,S, 7 021 AP Respond to a failure of N-41 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 078 Instrument Air System D,E 8 096 Respond to Loss of Control Air System j. 005 Residual Heat Removal System D,E,R 4P 044 Venting A-A RHR pump due to cavitation

k. 062 AC Electrical Distribution System D,E,A 6 091 AP Transfer Controls to Aux Mode per AOP-C.04, AU 3 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 I SRO-II SRO-U ES*301 Control Room/ln*Plant Systems Outline Form ES*301*2 Facility:

Seauovah Nuclear Plant Date of Examination: _02/16/2010 -Exam Level: RO D SRO-I X SRO-U D Operating Test No.: 2010301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF) System / JPM Title Type Code* Safety Function a. 005 Inoperable/Stuck Control Rod (AA2.03 3.5/4.4) D, S,A,L 1 001 AP-1 Emergency Boration (Stuck Rods) b. 011 Large Break LOCA (EA1.11 4.2/4.2) D, S,A, EN 2 013AP1 Transfer to Hot Leg Recirc. c. 008 Pressurizer (PZR) Vapor Space Accident (Relief Valve N,S,A 3* Stuck Open) (AA2.19 3.4/3.6) 070 AP Respond to Failed Open PZR Spray valve d. 005 Residual Heat Removal System (A4.01 3.6/3.4) D, S,L 4P . 152 Swap RHR pumps (A train to B train) with level in the PZR. e. WE05 Loss of Secondary Heat Sink (EA2.1 3.4/4.4) N,S 4S 034-1 Establish MFW per EA-2-2 f. g. 064 Emergency Diesel Generators (A4.06 3.9/3.9) D,S,A 6 077 -1 AP Perform DG load test on 1 B-B DG (with high crankcase pressure)

h. 015 Nuclear Instrumentation System (A2.02 3.1/3.5) M,S, 7 021 AP Respond to a failure of N-41 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U) i. 078 Instrument Air System D,E 8 096 Respond to Loss of Control Air System j. 005 Residual Heat Removal System D,E,R 4P 044 Venting A-A RHR pump due to cavitation

k. 062 AC Electrical Distribution System D,E,A 6 091 AP Transfer Controls to Aux Mode per AOP-C.04, AU 3 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 I SRO-II SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 :;;9/:;;8/:;;4

1/;

:1/;::1 -/ -/ ;::1 (control room system) ;::1/;::1/;::1

2/;::2/;::1

3/

;; 3/ :;; 2 (randomly selected)

1/;::1/;::1 JPM A. Candidate has to determine that emergency boration is required due to 2 stuck rods following

?t Rx trip, with the normal emergency boration valve failing to open, requiring the candidate to align alternate boration through the charging pump suction. This is a Bank, Low Power, Alternate path, JPM. . JPM B. Candidate is directed to make alignment change for Transfer to Hot Leg Recirculation per ES-I.4. following a trip due to large break LOCA. The Hot leg recirculation valve will fail to open requiring RHR to be aligned to Cold Leg injection with High head pumps aligned to Hot Legs. This is a Bank, Low Power, Alternate path, JPM. JPM C. Candidate will be required to respond to failed PZR pressure instrument which causes the PZR spray valves to open with one valve sticking open. With a PZR spray valve failing to close, candidate will follow actions of AOP-1.4 and trip the RX and trip at least two RCPs to stop mitigate the depressurization to prevent SI actuation. This is a Modified Bank, Alternate Path JPM. Original JPM JPM D. Plant is in Mode 4, and Candidate is directed to transfer RHR pumps from B train to A train. This is a Bank, Low PowerJPM. JPM E. Candidate is directed to establish a Secondary Heat Sink using Main Feed Water System following a Rx Trip. MFW will be required due to a failure of all AFW pumps. This is a New JPM. JPM G. Candidate is to perform a quick start ofEDG A-A and load the EDG. While the candidate is loading the EDG, a high crankcase condition will develop requiring a manual emergency trip of the EDG. This is a Bank, Alternate Path JPM. JPM H. Candidate will respond to failed Nuclear Instrument (N41) High. Control Rods will be stepping in at maximum rate, Candidate will take Rod bank selector switch to Manual and proceed to remove failed channel from service. This is a Modified Bank, JPM. Original JPM (021) had N-41 failing low from --45%, this JPM has N-4 1 failing High, requiring immediate manual action to stop control rod movement prior to removing channel from service. JPM I. Candidate is to restore Control Air pressure following a Loss of Off-site power. Candidate will restore Control and Service Air compressors using EA-32-1. This will require a Local Manual start. This a Bank JPM. JPM J. Candidate is directed to vent the lA-A RHR pump due to pump cavitating during mid-loop operation. This venting is done locally and is required to be performed to return the RHR pump to service. This is a Bank JPM and is performed in the RCA. JPM K. Candidate is directed to perform checklist 3 of AOP-C.04 Shutdown from Auxiliary Control Room, following an event which requires Control Room Abandonment. This a Bank, Alternate Path JPM. (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power / Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6 / 4-6 / 2-3 :;;9/:;;8/:;;4

1/;

:1/;::1 -/ -/ ;::1 (control room system) ;::1/;::1/;::1

2/;::2/;::1

3/

;; 3/ :;; 2 (randomly selected)

1/;::1/;::1 JPM A. Candidate has to determine that emergency boration is required due to 2 stuck rods following

?t Rx trip, with the normal emergency boration valve failing to open, requiring the candidate to align alternate boration through the charging pump suction. This is a Bank, Low Power, Alternate path, JPM. . JPM B. Candidate is directed to make alignment change for Transfer to Hot Leg Recirculation per ES-I.4. following a trip due to large break LOCA. The Hot leg recirculation valve will fail to open requiring RHR to be aligned to Cold Leg injection with High head pumps aligned to Hot Legs. This is a Bank, Low Power, Alternate path, JPM. JPM C. Candidate will be required to respond to failed PZR pressure instrument which causes the PZR spray valves to open with one valve sticking open. With a PZR spray valve failing to close, candidate will follow actions of AOP-1.4 and trip the RX and trip at least two RCPs to stop mitigate the depressurization to prevent SI actuation. This is a Modified Bank, Alternate Path JPM. Original JPM JPM D. Plant is in Mode 4, and Candidate is directed to transfer RHR pumps from B train to A train. This is a Bank, Low PowerJPM. JPM E. Candidate is directed to establish a Secondary Heat Sink using Main Feed Water System following a Rx Trip. MFW will be required due to a failure of all AFW pumps. This is a New JPM. JPM G. Candidate is to perform a quick start ofEDG A-A and load the EDG. While the candidate is loading the EDG, a high crankcase condition will develop requiring a manual emergency trip of the EDG. This is a Bank, Alternate Path JPM. JPM H. Candidate will respond to failed Nuclear Instrument (N41) High. Control Rods will be stepping in at maximum rate, Candidate will take Rod bank selector switch to Manual and proceed to remove failed channel from service. This is a Modified Bank, JPM. Original JPM (021) had N-41 failing low from --45%, this JPM has N-4 1 failing High, requiring immediate manual action to stop control rod movement prior to removing channel from service. JPM I. Candidate is to restore Control Air pressure following a Loss of Off-site power. Candidate will restore Control and Service Air compressors using EA-32-1. This will require a Local Manual start. This a Bank JPM. JPM J. Candidate is directed to vent the lA-A RHR pump due to pump cavitating during mid-loop operation. This venting is done locally and is required to be performed to return the RHR pump to service. This is a Bank JPM and is performed in the RCA. JPM K. Candidate is directed to perform checklist 3 of AOP-C.04 Shutdown from Auxiliary Control Room, following an event which requires Control Room Abandonment. This a Bank, Alternate Path JPM.}}