ML062070582

From kanterella
Jump to navigation Jump to search
AN2-07-2006-FINAL-Outline
ML062070582
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 07/17/2006
From: Gody A
Division of Reactor Safety IV
To: Forbes J
Entergy Operations
References
50-368/06-302 50-368/06-302
Download: ML062070582 (30)
v • d • e


Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ANO-2 Date of Examination: 07/17/2006 Examination Level: RO SRO Operating Test Number: 1 Type Administrative Topic Describe activity to be performed Code*

(see note)

Conduct of Operations Ability to perform specific system and integrated plant 2.1.23 RO (3.9) procedures during all modes of plant operation.

N Determine volume of Boric acid and DI water to makeup to RWT.

JPM-ANO-2-JPM-NRC-ADMIN RWT Ability to perform specific system and integrated plant Conduct of Operations procedures during all modes of plant operation.

2.1.23 RO (3.9) M Determine CEDM temperature using OP 2105.009.

Modified-JPM-ANO-2-JPM-NRC-ADMIN XTCEA Knowledge of Surveillance procedures.

Equipment Control 2.2.12 RO (3.0) D/P Review 2P89B Surveillance as RO.

Direct-JPM-ANO-2-NRC-ADMIN-Surveillance review 2P89B Radiation Control Knowledge of the process fro performing a containment purge.

2.3.9 RO (2.5)

N Complete a containment purge release permit New-JPM-ANO-2-JPM-NRC-ADMIN-complete containment purge release permit Emergency Plan NA NA 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 & ROretakes)

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

(P)revious 2 Exams ( 1; randomly selected)

ES-301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 Facility: ANO-2 Date of Examination: 07/17/2006 Examination Level: RO SRO Operating Test Number: 1 Type Administrative Topic Code* Describe activity to be performed (see note)

Determine RPS trip set point due to inoperable MSSV is correct using Technical Specifications.

Conduct of Operations D Ability to apply technical specification for a system 2.1.12 SRO (4.0) Direct- ANO-2-JPM-NRC-ADMIN MSSVINOP Ability to perform specific system and integrated plant Conduct of Operations procedures during all modes of plant operation.

M Approve CEDM temperature calculation using OP 2105.009.

2.1.23 SRO (4.0)

Modified-ANO-2-JPM-NRC-ADMIN XTCEA Equipment Control Knowledge of Surveillance procedures.

D/P Review 2P89B Surveillance.

2.2.12 SRO (3.4)

Direct-ANO-2-JPM-NRC-ADMIN-Surveillance review 2P89B Knowledge of the process fro performing a containment Radiation Control purge.

N Review a containment purge release permit 2.3.9 SRO (3.4)

New-JPM-ANO-2-JPM-NRC-ADMIN-review containment purge release Knowledge of the emergency plan.

Emergency Plan Classify EAL and complete applicable Shift Manager forms.

N 2.4.29 SRO (4.0) New-ANO-2-JPM-NRC-ADMIN-Classify event and Complete SM E-plan forms 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: __ANO UNIT 2__________________ Date of Examination: _07/17/2006____

Exam Level (circle one): RO SRO(I) SRO(U) Operating Test No.: ____1_________

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. ANO-2-JPM-NRC-ELEC EOP 2 A/L/N/S 6 062 A4.01 RO-3.3 SRO-3.1 Electrical Energize 2A2, non-vital 4160VAC bus following a Loss Of Offsite Power
b. ANO-2-JPM-NRC-SIT006 A/L/D/S 2 006 A4.08 RO-4.2 SRO-4.3 Inventory Isolate Safety Injection Tanks with Safety Injection Actuation System actuated
c. ANO-2-JPM-NRC-LTOP L/N/S 3 010 K4.03 RO-3.8 SRO-4.1 Reactor Pressure Control Respond to Annunciator 2K10 C-4 and place low temperature overpressure relief valves inservice
d. ANO-2-JPM-NRC-RCP03 A/L/D/P/S 8 Plant Service Systems 008 A4.01 RO-3.3 SRO-3.1 Restore Component Cooling Water to Reactor Coolant Pumps
e. ANO-2-JPM-NRC-CEA5 A/M/S 1 Reactivity 001 A4.03 RO-4.0 SRO-3.7 Exercise a Control Element Assembly
f. ANO-2-JPM-NRC-FWCS1 D/S 4 Heat Removal 035 A4.01 RO-3.7 SRO- 3.6 Place Feed Water Control System in Automatic
g. ANO-2-JPM-NRC-H2001 C/D 5 028 A4.01 RO-4.0 SRO-4.0 Containment Integrity Manually start Hydrogen analyzer
h. ANO-2-JPM-NRC-ICI01 7 D/P/S 015 A2.02 RO-3.1 SRO-3.5 Instrumentation Remove Incore instrument from scan for Core Operating Limits Supervisory System In- Plant Systems@ ( 3 for RO; 3 for SRO-I; 3 or 2 for SRO- U)
i. ANO-2-JPM-NRC-AACGLS 064 A3.06 RO-3.3 SRO-3.4 A/D 6 Electrical Local start of Station Blackout Diesel
j. ANO-2-JPM-NRC-P36ASD 004 A4.08 RO-3.8 SRO-3.4 D/E/R 2 Inventory Operate Charging Pump 2P36B Locally During Alternate Shutdown
k. ANO-2-JPM-NRC-PRHTR 3 D/E Reactor Pressure 006 A2.01 RO-3.3 SRO-3.6 Control Locally control pressurizer proportional heaters

@ All 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)direct from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 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: __ANO UNIT 2__________________ Date of Examination: _07/17/2006____

Exam Level (circle one): RO / SRO(I) / SRO(U) Operating Test No.: ____1_________

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. ANO-2-JPM-NRC-ELEC EOP 2 A/L/N/S 6 062 A4.01 RO-3.3 SRO-3.1 Electrical Energize 2A2, non-vital 4160VAC bus following a Loss Of Offsite Power
b. ANO-2-JPM-NRC-SIT006 A/L/D/S 2 006 A4.08 RO-4.2 SRO-4.3 Inventory Isolate Safety Injection Tanks with Safety Injection Actuation System actuated
c. ANO-2-JPM-NRC-LTOP L/N/S 3 010 K4.03 RO-3.8 SRO-4.1 Reactor Pressure Control Respond to Annunciator 2K10 C-4 and place low temperature overpressure relief valves inservice
d. ANO-2-JPM-NRC-RCP03 A/L/D/P/S 8 Plant Service Systems 008 A4.01 RO-3.3 SRO-3.1 Restore Component Cooling Water to Reactor Coolant Pumps
e. ANO-2-JPM-NRC-CEA5 A/M/S 1 Reactivity 001 A4.03 RO-4.0 SRO-3.7 Exercise a Control Element Assembly
f. ANO-2-JPM-NRC-FWCS1 D/S 4 Heat Removal 035 A4.01 RO-3.7 SRO- 3.6 Place Feed Water Control System in Automatic
g. ANO-2-JPM-NRC-H2001 C/D 5 028 A4.01 RO-4.0 SRO-4.0 Containment Integrity Manually start Hydrogen analyzer h.

In- Plant Systems@ ( 3 for RO; 3 for SRO-I; 3 or 2 for SRO- U)

i. ANO-2-JPM-NRC-AACGLS 064 A3.06 RO-3.3 SRO-3.4 A/D 6 Electrical Local start of Station Blackout Diesel
j. ANO-2-JPM-NRC-P36ASD 004 A4.08 RO-3.8 SRO-3.4 D/E/R 2 Inventory Operate Charging Pump 2P36B Locally During Alternate Shutdown
k. ANO-2-JPM-NRC-PRHTR 3 D/E Reactor Pressure 006 A2.01 RO-3.3 SRO-3.6 Control Locally control pressurizer proportional heaters

@ All 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)direct from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 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: __ANO UNIT 2__________________ Date of Examination: _7/17/2006____

Exam Level : RO / SRO(I) / SRO(U) Operating Test No.: ____1_________

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.

b.

c. ANO-2-JPM-NRC-LTOP L/N/S 3 010 K4.03 RO-3.8 SRO-4.1 Reactor Pressure Control Respond to Annunciator 2K10 C-4 and place low temperature overpressure relief valves in service
d. ANO-2-JPM-NRC-RCP03 A/L/D/P/S 8 Plant Service Systems 008 A4.01 RO-3.3 SRO-3.1 Restore Component Cooling Water to Reactor Coolant Pumps
e. ANO-2-JPM-NRC-CEA5 A/M/S 1 Reactivity 001 A4.03 RO-4.0 SRO-3.7 Exercise a Control Element Assembly f.

g.

h.

In- Plant Systems@ ( 3 for RO; 3 for SRO-I; 3 or 2 for SRO- U)

i. ANO-2-JPM-NRC-AACGLS 064 A3.06 RO-3.3 SRO-3.4 A/D 6 Electrical Local start of Station Blackout Diesel
j. ANO-2-JPM-NRC-P36ASD 004 A4.08 RO-3.8 SRO-3.4 D/E/R 2 Inventory Operate Charging Pump 2P36B Locally During Alternate Shutdown k.

@ All 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)direct from bank 9 / 8 / 4 (E)mergency or abnormal in-plant 1 / 1 / 1 (L)ow-Power 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: ANO-2 Scenario No.: 1 Op-Test No.: 2006-1 Page 1 Examiners: Operators:

Initial Conditions:

20% MOL, All Engineered Safety Features systems are in standby. Plant startup following a five day stator water cooling outage. 2P27, MFP standby lube oil pump tagged out for maintenance. Green Train Maintenance Week.

Turnover:

20%. 250 EFPD. EOOS indicates Minimal Risk. Plant Startup in progress; OP 2104.004 section 9, raising power above 20%, is the controlling procedure. 2P27, MFP standby lube oil pump tagged out for maintenance. Green Train Maintenance Week.

Event Malf. No. Event Event No. Type* Description 1 Raise Power R (ATC) Raise reactor and turbine power.

above 20%

2 XRCCHAPCNT I (ATC) Pressurizer control channel pressure fails high.

3 CCWFAILBAUTO C (CBOT) Loop II Component Cooling Water pump C Trips CCW2P33CPWR and B Component Cooling Water pump fails to automatically start.

4 RCP2P32AASLK M (ATC) Reactor Coolant System inter-system leak into N (CBOT) Component Cooling Water system resulting in Loss of Coolant Accident and Safety Injection Actuation System (post reactor trip).

5 BUS2H2 M (ALL) 2H2 lockout. Loss of 2 Reactor Coolant Pumps and one condenser circulating water pump.

6 RPSRXAUTO C (ATC) Reactor Protection System fails to automatically trip the reactor on loss of Reactor Coolant Pumps.

7 XMSHDRPRS I (CBOT) Steam Dump and Bypass Control System fails to automatically open bypass and dump valves to control Steam Generator pressure.

8 HPI2P89AFAL C (CBOT) 2P89A, A High Pressure Safety Injection pump fails to auto-start.

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

Appendix D Scenario Outline Form ES-D-1 Facility: ANO-2 Scenario No.: 2 Op-Test No.: 2006-1 Page 1 Examiners: Operators:

Initial Conditions:

100% MOL, All ESF systems in standby. Green Train Maintenance Week.

Turnover:

100%. 250 EFPD. EOOS indicates Minimal Risk. Green Train Maintenance Week. B main chiller is tagged out for oil change out.

Event Malf. No. Event Event No. Type* Description 1 COND2P2AWIND C(CBOT) A Condensate Pump motor winding with rise resulting in manual start of D Condensate Pump and securing A Condensate Pump.

2 XCV2LT4861 I (ATC) Volume Control Tank level instrument fails low resulting in Refueling Water Tank being aligned to Coolant Charging Pump suction.

3 CWS2P3BBOL R (ATC) Trip 2P3B, B Circulating Water Pump, which N (CBOT causes a partial loss of main condenser circulating water flow resulting in a rapid down power to ~ 90%

power.

4 MSSGBLK M (CBOT) B Steam Generator Excess Steam Demand (ESD)

M (ATC) inside containment results in manual reactor trip and control of Reactor Coolant System heat up and Pressurizer pressure post SG blowdown.

5 CEA51STUCK C (ATC) Control Element Assembly #51 stuck on reactor trip results in Emergency Boration.

6 CV1036-2 C (CBOT) B Emergency Feed Water (EFW) Pump to B CV1075-1 Steam Generator valves fail to close from control room resulting in over feeding Steam Generator with ESD cooldown of RCS unnecessarily. Secure B EFW pump.

7 BS2P35BFAL C (CBOT) B Spray Pump fails to start. Can be manually BS2P35AFAULT started.

A Spray Pump cannot be restarted.

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

Appendix D Scenario Outline Form ES-D-1 Facility: ANO-2 Scenario No.: 3 Op-Test No.: 2006-1 Page 1 Examiners: Operators:

Initial Conditions:

100% MOL, All ESF systems in standby. Green Train Maintenance Week.

Turnover:

100%. 250 EFPD. EOOS indicates Minimal Risk. Green Train Maintenance Week.

Tornado watch for Pope County in effect until 10:00pm today. Natural Emergency AOP actions have been completed.

Event Malf. No. Event Event No. Type* Description 1 CV-4816 C(ATC) Letdown flow control valve fails closed. Restore Letdown using 2CV 4817.

2 XFW2FIS0735 I (CBOT) A Main Feedwater Pump (MFP) suction flow transmitter fails low. Recirculation valves (A loop condensate and A MFP) to manual and closed.

3 MFWWPMPBTRP R (ATC) B MFP trips on thrust bearing wear. Emergency N (CBOT borate and reduce power to <90%. Start D Condensate Pump.

4 500LOSE500 M (CBOT) Loss of offsite power due to tornado.

500LOSE161 M (ATC) 5 EDGDG1OIL C (CBOT) #1 Emergency Diesel Generator trips on low lube oil pressure. Loss of Red vital AC power. Loss of 2Y1, 120VAC non-vital power. CBOT energize Vital Red AC busses with Station Blackout diesel generator.

6 Channel 1 PZR I (ATC) Pressurizer control channel 1 for level and pressure level and pressure will fail low due to loss of power. Control manually /

channels lose select opposite channel.

power.

7 CV0332 C (CBOT) Over speed trip of A Emergency Feedwater Pump.

Requires starting B Emergency Feedwater Pump.

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

ES-401 PWR Examination Outline Form ES-401-2 Facility: Arkansas Nuclear One Unit 2 RO/SRO Written Outline Date of Exam: 07/14/2006 RO K/A Category Points SRO - Only Points Tier Group K K K K K K A A A A G Total A2 G* TOTAL 1 2 3 4 5 6 1 2 3 4 *

1. 1 3 3 3 3 3 3 18 4 2 6 Emergency &

Abnormal Plant Evolutions 2 1 2 N/A 1 1 N/A 2 9 3 4 2 1 Tier Totals 5 4 5 4 4 5 27 7 3 10 1 2 2 3 3 3 2 2 2 3 3 3 28 3 2 5 2.

Plant 1 1 1 0 1 1 1 1 1 1 1 10 1 3 2 2 Systems Tier Totals 3 3 4 3 4 3 3 3 4 4 4 38 4 4 8

3. Generic Knowledge and Abilities Catergories 1 2 3 4 1 2 3 4 10 7 3 2 2 3 2 2 1 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO Outline and the SRO only outlines (i.e. except for one category in Tier 3 of the SRO-only outline, the Tier 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 (IR) 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. 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 number, descriptions, importance ratings, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43 NUREG-1021, Revision 9 Page 1 of 13

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier1 /Group1 (RO /SRO)

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

1 2 3 1 2 00007 (BW/E02 & E10; CE/E02) Reactor X Knowledge of the reasons for the following responses as 3.2 1 Trip - Stabilization - Recovery / 1 they apply to the (Reactor Trip Recovery):

EK3.4 RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated 00007 (BW/E02 & E10; CE/E02) Reactor x Generic 3.0 76 Trip - Stabilization - Recovery / 1 2.1.19 Ability to use plant computer to obtain and evaluate parametric information on system or component status.

00008 Pressurizer Vapor Space X Ability to determine and interpret the following as they apply 3.4 2 Accident/3 to the Pressurizer Vapor Space Accident:

AA2.19 PZR spray valve failure, using plant parameters 000009 Small Break LOCA / 3 X Knowledge of the interrelations between the small break 3.0 3 LOCA and the following:

EK2.03 S/Gs 000011 Large Break LOCA / 3 X Ability to determine and interpret the following as they apply 3.3* 4 to a Large Break LOCA:

EA2.02 Consequences to RHR of not resetting safety injection 000015/17 RCP Malfunctions / 4 X Knowledge of the interrelations between the Reactor 2.6 5 Coolant Pump Malfunctions and the following:

AK2.08 CCWS 000022 Loss of Rx Coolant Makeup / 2 X Knowledge of the operational implications of the following 3.0 6 concepts as they apply to Loss of Reactor Coolant Pump Makeup:

AK1.03 Relationship between charging flow and PZR level.

000025 Loss of RHR System / 4 X Knowledge of the operational implications of the following 3.9 7 concepts as they apply to Loss of Residual Heat Removal System:

AK1.01 Loss of RHRS during all modes of operation 000026 Loss of Component Cooling X Generic: 3.1 8 Water / 8 2.4.6 Knowledge symptom based EOP mitigation strategies.

000027 Pressurizer Pressure Control X Generic: 3.4 9 System Malfunction / 3 2.4.11 Knowledge of abnormal condition procedures.

000029 ATWS / 1 X Ability to operate and/or monitor the following as they apply 3.4* 10 to a ATWS:

EA1.01 Charging pumps Ability to determine and interpret the following as they apply 000038 Steam Gen. Tube Rupture / 3 x to a SGTR: 4.8 77 EA2.02 Existence of an S/G tube rupture and its potential consequences Ability to determine and interpret the following as they apply 000038 Steam Gen. Tube Rupture / 3 x to a SGTR: 4.2 78 EA2.12 Status of MSIV activating system NUREG-1021, Revision 9 Page 2 of 13

ES-401 2 Form ES-401-2 000040 (BW/E05; CE/E05; W/E12) X 040 3.2 11 Steam Line Rupture - Excessive Heat Knowledge of the operational implications of the following Transfer / 4 concepts as they apply to Steam Line Rupture:

AK1.04 Nil ductility temperature 000040 (BW/E05; CE/E05; W/E12) X CE/E05 3.9 12 Steam Line Rupture - Excessive Heat Ability to operate and/or monitor the following as they apply Transfer / 4 to the (Excess Steam Demand):

EA1.1 Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

054 000054 (CE/E06) Loss of Main X 4.3 13 Feedwater / 4 Ability to determine and interpret the following as they apply to the Loss of Main Feedwater (MFW):

AA2.01 Occurrence of reactor and/or turbine trip CE/E06 000054 (CE/E06) Loss of Main X 3.5 14 Feedwater / 4 Knowledge of the interrelations between the (Loss of Feedwater) and the following:

EK2.2 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.

000055 Station Blackout / 6 x Generic 3.3 79 2.2.20 Knowledge of the process for managing troubleshooting activities.

000056 Loss of Off-site Power / 6 X Knowledge of the reasons for the following responses as 3.5 15 they apply to the Loss of Offsite Power:

AK3.01 Order and time to initiation of power for the load sequencer 000057 Loss of Vital AC Inst. Bus / 6 X Knowledge of the reasons for the following responses as 4.1 16 they apply to the Loss of Vital AC Instrument Bus:

AK3.01 Actions contained in EOP for loss of vital ac electrical instrument bus.

Ability to operate and/or monitor the following as they apply 000058 Loss of DC Power / 6 X to the Loss of DC Power: 3.1* 17 AA1.02 Static inverter dc input breaker, frequency meter, ac output breaker, and ground fault detector 000062 Loss of Nuclear Svc Water / 4 X Generic 3.2 18 2.1.28 Knowledge of the purpose and function of major system components and controls.

Ability to determine and interpret the following as they apply 000065 Loss of Instrument Air / 8 x to the Loss of Instrument Air: 2.6* 80 AA2.02 Relationship of flow readings to system operation Ability to determine and interpret the following as they apply 000065 Loss of Instrument Air / 8 x to the Loss of Instrument Air: 4.1 81 AA2.05 When to commence plant shutdown if instrument air pressure is decreasing W/E04 LOCA Outside Containment / 3 Not applicable to this Unit.

W/E11 Loss of Emergency Coolant Not applicable to this Unit.

Recirc. / 4 NUREG-1021, Revision 9 Page 3 of 13

ES-401 2 Form ES-401-2 BW/E04; W/E05 Inadequate Heat Transfer Not applicable to this Unit.

- Loss of Secondary Heat Sink / 4 K/A Category Totals: 3 3 3 3 3 3 Group Point Total: 18/6

/ /

4 2 NUREG-1021, Revision 9 Page 4 of 13

ES-401 3 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier1 /Group2 (RO /SRO)

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 X Knowledge of the interrelations between the 3.0* 19 Continuous Rod Withdrawal and the following:

AK2.06 T-ave./ref. deviation meter Ability to determine and interpret the following as they 000003 Dropped Control Rod / 1 x apply to the Dropped Control Rod: 3.8 82 AA2.03 Dropped rod, using in-core/ex-core instrumentation, in-core or loop temperature measurements 000005 Inoperable/Stuck Control Rod / 1 X Ability to operate and/or monitor the following as they 3.4 20 apply to the Inoperable/Stuck Control Rod:

AA1.05 RPI Ability to determine and interpret the following as they 000024 Emergency Boration / 1 x apply to the Emergency Boration: 4.4 83 AA2.02 When use of manual boration valve is needed 000028 Pressurizer Level Malfunction / 2 Not Selected.

000032 Loss of Source Range NI / 7 Not Selected.

000033 Loss of Intermediate Range NI / 7 Not applicable to this Unit.

Knowledge of the reasons for the following responses 000036 (BW/A08) Fuel Handling X as they apply to the Fuel Handling Incidents: 3.7 21 Accident / 8 AK3.03 Guidance contained in EOP for fuel handling incident.

000037 Steam Generator Tube Leak / 3 X Generic 3.4 22 2.2.22 Knowledge of limiting conditions for operations and safety limits.

000051 Loss of Condenser Vacuum / 4 X Generic 3.4 23 2.1.32 Ability to explain and apply all system limits and precautions.

000059 Accidental Liquid RadWaste Not Selected.

Rel. / 9 000060 Accidental Gaseous Radwaste Not Selected.

Rel. / 9 000061 ARM System Alarms / 7 X Knowledge of the operational implications of the 2.5* 24 following concepts as they apply to Area Radiation Monitoring (ARM)

AK1.01 Detector limitations 000067 Plant Fire On-site / 8 Not Selected.

000068 (BW/A06) Control Room Evac. / 8 Not Selected.

Ability to determine and interpret the following as they 000069 (W/E14) Loss of CTMT Integrity / 5 x apply to the Loss of Containment Integrity: 4.3 84 AA2.01 Loss of containment integrity 000074 (W/E06&E07) Inad. Core Not Selected.

Cooling / 4 NUREG-1021, Revision 9 Page 5 of 13

ES-401 3 Form ES-401-2 000076 High Reactor Coolant Activity / 9 X Ability to determine and interpret the following as they 2.8 25 apply to the High Reactor Coolant Activity:

AA2.02 Corrective actions required for high fission product activity in RCS W/EO1 & E02 Rediagnosis & SI Not applicable to this Unit.

Termination / 3 W/E13 Steam Generator Over-pressure / 4 Not applicable to this Unit.

W/E15 Containment Flooding / 5 Not applicable to this Unit.

W/E16 High Containment Radiation / 9 Not applicable to this Unit.

BW/A01 Plant Runback / 1 Not applicable to this Unit.

BW/A02&A03 Loss of NNI-X/Y / 7 Not applicable to this Unit.

BW/A04 Turbine Trip / 4 Not applicable to this Unit.

BW/A05 Emergency Diesel Actuation / 6 Not applicable to this Unit.

BW/A07 Flooding / 8 Not applicable to this Unit.

BW/E03 Inadequate Subcooling Margin / 4 Not applicable to this Unit.

BW/E08; W/E03 LOCA Cooldown - Not applicable to this Unit.

Depress. / 4 BW/E09; CE/A13; W/E09&E10 Natural Not Selected.

Circ. / 4 BW/E13&E14 EOP Rules and Enclosures Not applicable to this Unit.

Knowledge of the operational implications of the CE/A11; W/E08 RCS Overcooling - PTS / 4 X following concepts as they apply to the (RCS 3.1 26 Overcooling):

AK1.1 Components, capacity, and function of emergency systems CE/A16 Excess RCS Leakage / 2 X Knowledge of the reasons for the following responses 2.8 27 as they apply to the (Excess RCS Leakage):

AK3.2 Normal, abnormal and emergency operating procedures associated with (Excess RCS Leakage)

CE/E09 Functional Recovery x Generic 3.3 85 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

K/A Category Point Totals: 2 1 2 1 1 2 9/4

/ /

3 1 NUREG-1021, Revision 9 Page 6 of 13

ES-401 4 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant systems - Tier 2/Group 1 (RO / 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 4 003 Reactor Coolant Pump X Knowledge of RCPS design feature(s) and/or 2.8 28 interlock(s) which provide for the following:

K4.04 Adequate cooling of RCP motor and seals 003 Reactor Coolant Pump x Generic 3.8 86 2.2.29 Knowledge of SRO fuel handling responsibilities.

Ability to predict and/or monitor changes in 004 Chemical and Volume Control X parameters (to prevent exceeding design 3.0 29 limits) associated with operating the CVCS controls including:

A1.06 VCT level Knowledge of bus power supplies to the 005 Residual Heat Removal X following: 2.7* 30 K2.03 RCS Pressure boundary motor operated valves.

Ability to manually operate and/or monitor in 005 Residual Heat Removal X the control room: 2.8* 31 A4.03 RHR temperature, PZR heaters and flow, and nitrogen 006 Emergency Core Cooling X Ability to monitor automatic operation of the 4.2 32 ECCS, including:

A3.08 Automatic transfer of ECCS flowpaths Ability to (a) predict the impacts of the 006 Emergency Core Cooling x following malfunctions or operations on the 3.2* 87 ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.09 Radioactive release from venting RWST to atmosphere 007 Pressurizer Relief/ Quench Tank X Knowledge of the operational implications of 3.1 33 the following concepts as they apply to the PRTS:

K5.02 Method of forming a steam bubble in the PZR.

008 Component Cooling Water X Ability to (a) predict the impacts of the 3.3* 34 following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.05 Effect of loss of instrument and control air on the position of the CCW valves that are airoperated Knowledge of bus power supplies to the 010 Pressurizer Pressure Control X following: 3.0 35 K2.01 PZR heaters.

Knowledge of the effect that a loss or 012 Reactor Protection X malfunction of the RPS will have on the 3.1* 36 following:

K3.03 SDS NUREG-1021, Revision 9 Page 7 of 13

ES-401 4 Form ES-401-2 013 Engineered Safety Features X Knowledge of the physical connections 3.7 37 Actuation and/or cause-effect relationships between the ESFAS and the following systems:

K1.18 Premature reset of ESF actuation 013 Engineered Safety Features X Generic 2.8 38 Actuation 2.1.25 Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data.

Ability to manually operate and/or monitor in 022 Containment Cooling X the control room: 3.2* 39 A4.02 CCS pumps Ability to manually operate and/or monitor in 022 Containment Cooling X the control room: 3.1* 40 A4.04 Valves in the CCS 025 Ice Condenser Not applicable to this unit.

026 Containment Spray X Ability to monitor automatic operation of the 3.9* 41 CSS, including:

A3.02 Verification that cooling water is supplied to the containment spray heat exchanger 026 Containment Spray x Generic 3.8 88 2.4.17 Knowledge of EOP terms and conditions.

Knowledge of the effect that a loss or 039 Main and Reheat Steam X malfunction of the MRSS will have on the 2.5* 42 following:

K3.04 MFW pumps 059 Main Feedwater X Knowledge of the effect that a loss or 3.6 43 malfunction of the MFW System will have on the following:

K3.02 AFW System 059 Main Feedwater X Generic 3.0 44 2.4.14 Knowledge of general guidelines for EOP flowchart use.

061 Auxiliary/Emergency Feedwater X Knowledge of the physical connections 3.6 45 and/or cause-effect relationships between the AFW System and the following systems:

K1.07 Emergency Water Source Ability to (a) predict the impacts of the 061 Auxiliary/Emergency Feedwater x following malfunctions or operations on the 3.4* 89 AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.05 Automatic control malfunction 062 AC Electrical Distribution X Ability to predict and/or monitor changes in 2.5 46 parameters (to prevent exceeding design limits) associated with operating the A.C.

Distribution System controls including:

A1.03 Effect on instrumentation and controls of switching power supplies NUREG-1021, Revision 9 Page 8 of 13

ES-401 4 Form ES-401-2 063 DC Electrical Distribution X Ability to (a) predict the impacts of the 2.5 47 following malfunctions or operations on the D.C. Electrical System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.01 Grounds 064 Emergency Diesel Generator X Knowledge of the effect of a loss or 3.2 48 malfunction of the following will have on the ED/G System:

K6.08 Fuel oil storage tanks 064 Emergency Diesel Generator X Knowledge of the effect of a loss or 2.7 49 malfunction of the following will have on the ED/G System:

K6.07 Air receivers 073 Process Radiation Monitoring X Knowledge of the operational implications of 2.5 50 the following concepts as they apply to the PRM System:

K5.01 Radiation theory, including sources, types, units, and effects 073 Process Radiation Monitoring X Knowledge of the operational implications of 2.5 51 the following concepts as they apply to the PRM System:

K5.02 Radiation intensity changes with source distance 076 Service Water X Knowledge of SWS design feature(s) and/or 2.9 52 interlock(s) which provide for the following:

K4.02 Automatic start features associated with SWS pump controls 078 Instrument Air X Knowledge of IAS design feature(s) and/or 3.2 53 interlock(s) which provide for the following:

K4.02 Cross-over to other air systems 078 Instrument Air X Generic 3.6 54 2.2.13 Knowledge tagging and clearance procedures.

103 Containment X Ability to monitor automatic operation of the 3.9 55 containment system, including:

A3.01 Containment Isolation Ability to (a) predict the impacts of the 103 Containment x following malfunctions or operations on the 3.6* 90 containment system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.04 Containment evacuation (including recognition of the alarm)

K/A Category Point Totals: 2 2 3 3 3 2 2 2 3 3 3 Group Point Total: 28/5

/ /

3 2 NUREG-1021, Revision 9 Page 9 of 13

ES-401 5 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant systems - Tier 2/Group 2 (RO / 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 4 001 Control Rod Drive Not selected 002 Reactor Coolant X Ability to manually operate and/or monitor in the 3.5* 56 control room:

A4.01 RCS leakage calculation program using the computer 011 Pressurizer Level Control X Knowledge of the effect that a loss or malfunction 3.2 57 of the PZR LCS will have on the following:

K3.03 PZR PCS 014 Rod Position Indication Not selected 015 Nuclear Instrumentation X Knowledge of bus power supplies to the 3.3 58 following:

K2.01 NIS channels, components, and interconnections 016 Non-nuclear Ability to (a) predict the impacts of the following Instrumentation x malfunctions or operations on the NNIS and (b) 3.3* 91 based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.03 Interruption of transmitted signal 017 In-core Temperature Monitor X Ability to monitor automatic operation of the ITM 3.6* 59 System, including:

A3.01 Indications of normal, natural, and interrupted circulation of RCS 027 Containment Iodine Not applicable to this unit Removal 028 Hydrogen Recombiner and Purge Control Not selected 029 Containment Purge Not selected 033 Spent Fuel Pool Cooling X Knowledge of the physical connections and/or 2.7* 60 cause-effect relationships between the Spent Fuel Pool Cooling System and the following systems:

K1.05 RWST 034 Fuel Handling Equipment Not selected 035 Steam Generator Knowledge of the effect of a loss or malfunction X of the following will have on the S/GS: 3.2 61 K6.01 MSIVs 041 Steam Dump/Turbine Bypass Control Not selected 045 Main Turbine Generator X Ability to (a) predict the impacts of the following 2.7* 62 malfunctions or operations on the MT/G System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

A2.17 Malfunction of electrohydraulic control 055 Condenser Air Removal Not selected NUREG-1021, Revision 9 Page 10 of 13

ES-401 5 Form ES-401-2 056 Condensate X Generic 2.6 63 2.3.1 Knowledge of 10CFR20 and related facility radiation control requirements.

068 Liquid Radwaste x Generic 4.1 92 2.4.41 Knowledge of the emergency action level thresholds and classifications.

071 Waste Gas Disposal x Generic 3.2 93 2.3.8 Knowledge of process for performing a planned gaseous radioactive release.

072 Area Radiation Monitoring X Knowledge of the operational implications of the 2.7 64 following concepts as they apply to the ARM system:

K5.01 Radiation theory, including sources, types, units, and effect.

075 Circulating Water Not selected 079 Station Air Not selected 086 Fire Protection X Ability to predict and/or monitor changes in 2.9 65 parameters (to prevent exceeding design limits) associated with operating the Fire Protection System controls including:

A1.05 FPS lineups K/A Category Totals: 1 1 1 0 1 1 1 1 1 1 1 Group Point Total: 10/3

/ /

1 2 NUREG-1021, Revision 9 Page 11 of 13

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: Arkansas Nuclear One Unit 2 RO/SRO Written Outline Date of Exam:

07/14/2006 Category K/A # Topic RO SRO-only IR # IR #

2.1.7 3.7 66 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

1.

2.1.22 2.8 67 Conduct of Ability to determine Mode of Operation.

Operations 2.1.27 2.8 68 Knowledge of system purpose and or function 2.1.11 Knowledge of less than one hour technical specification action statements for systems. 3.8 94 2.1.30 Ability to locate and operate components, including local controls. 3.4 95 Subtotal 3 2 2.2.1 3.7 69 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

2.2.26 2.5 70 2.

Knowledge of refueling administrative requirements Equipment Control 2.2.10 Knowledge of the process for determining if the margin of safety, as defined in the basis of 3.3 96 any technical specification is reduced by a proposed change, test, or experiment.

2.2.11 3.4* 97 Knowledge of the process for controlling temporary changes.

Subtotal 2 2 2.3.9 2.5 71 Knowledge of the process for performing a containment purge.

2.3.11 2.7 72 3.

Ability to control radiation releases.

Radiation Control 2.3.6 Knowledge of the requirements for reviewing and approving release permits. 3.1 98 Subtotal 2 1 NUREG-1021, Revision 9 Page 12 of 13

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 2.4.35 3.3 73 Knowledge of local auxiliary operator tasks during emergency operations including system geography and system implications.

2.4.47 3.4 74

4. Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Emergency Procedures/ Plan 2.4.49 4.0 75 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

2.4.36 Knowledge of chemistry / health physics tasks during emergency operations. 2.8 99 2.4.48 Ability to interpret control room indications to verify the status and operation of system, and 3.8 100 understand how operator actions and directives affect plant and system conditions.

Subtotal 3 2 Tier 3 Point Total 10 7 NUREG-1021, Revision 9 Page 13 of 13

METHOD USED FOR RANDOM K/A SAMPLING A commercial random generation program was used to generate both the RO and SRO written sample plan. The program which was specifically designed for Combustion Engineering designed plants was supplied by the Westinghouse Owners Group (WOG). The name of the program is PWR K&A Database by WD Version 2.2.0 June 2004. This program pre-screens the non-CE related EPE/APE K/As and allows all other K/As of 2.5 or greater importance rating to be sampled. The program also provides for manual input for suppression of additional K/As to tailor the sample pool to fit individual unit design.

Several individual K/As were suppressed as well as 3 complete systems due to design applicability to ANO Unit 2. The systems suppressed include 025 - Ice Condenser System, 027 - Containment Iodine Removal System, and EPE/APE 033 -

Loss of Intermediate Range Nuclear Instrumentation. ANO Unit 2 does not have an Ice Condenser in Containment nor do we have intermediate range nuclear instrumentation.

The excore nuclear instrumentation used at ANO covers the full spectrum of power from the subcritical range to 200% power. The Containment Spray system in conjunction with a passive chemical addition system provides the function of post accident containment iodine removal. A complete list of Suppressed K/As has been submitted for review.

The computer program was initially designed for the Draft Revision 9 of NUREG-1021 and did not select the required SRO points. Westinghouse was contacted prior to the 2005 exam and their programmer sent an executable batch file to update the required SRO point totals, however, the program continued to select one less SRO point total than required in the Plant Systems Tier 2 Group 1 and 2. A manual random selection of an additional K/A for both Tier 2 Group 1 and Tier 2 Group 2 of the SRO outline was performed. This allowed us to align our point totals to match the required point totals in Revision 9 of NUREG-1021 for both the RO and SRO outline. The manual sample was performed as described below.

The system numbers for the remaining systems for Tier 2 Group 1 were placed in a can and a system number was randomly selected. The number selected was 061, Auxiliary/Emergency Feedwater. The required SRO categories A2 and G were then placed in the can and A2 was randomly selected. The numbers 1-9 were then placed in the can and the number 5 was randomly selected. This gave a final K/A of 061 A2.05 which was then added to the written outline.

The system numbers for the remaining systems for Tier 2 Group 2 were placed in a can and a system number was randomly selected. The number selected was 071, Waste Gas Disposal. Then the required SRO categories A2 and G were placed in the can and G was randomly selected. The numbers 1-4 were then placed in the can to select the generic category and the number 4 was randomly selected. Finally, the numbers 1-50 were placed in the can and the number 37 was selected. This gave a final K/A of Generic 2.4.37. This was later rejected, as noted on the Record of Rejected K/As Form ES-401-4 due to the overdraw on the generic category 4 K/As, and replaced with generic 2.3.8.

Replacement of the rejected K/As was performed by manual random sample similar to the method described above and is discussed on the Record of Rejected K/As Form ES-401-4.

Suppressed K/As Printed: 03/27/2006 Facility: Arkansas Nuclear One - Unit 2 IMPORTANCE Basis RO / SRO 001 Continuous Rod Withdrawal AK1.14 Interaction of ICS control stations as well as purpose, function, and modes of ICS not applicable to 3.4*/3.7 operation of ICS Unit 2 003 Dropped Control Rod AK1.13 Interaction of ICS control stations as well as purpose, function, and modes of Not applicable to Unit 2. 3.2*/3.6 operation of ICS AK2.03 Metroscope Not applicable to Unit 2 3.1*/3.2*

AK3.01 When ICS logic has failed on a dropped rod, the load must be reduced until flux is Not applicable to Unit 2 3.5*/3.9*

within specified target bank 005 Inoperable/Stuck Control Rod AK2.03 Metroscope Not applicable to Unit 2 3.1*/3.3*

systems.

AA1.03 Metroscope Not applicable to Unit 2 3.4*/3.4*

systems.

AA2.02 Difference between jog and run rod speeds, effect on CRDM of stuck rod Not applicable to Unit 2 2.5*/3.0*

systems.

015 017 Reactor Coolant Pump (RCP) Malfunctions AK1.03 The basis for operating at a reduced power level when one RCP is out of service Not applicable to unit 2 3.0*/4.0*

design/operation.

AK3.04 Reduction of power to below the steady state power-to-flow limit Not applicable to unit 2 3.1*/3.2*

design/operation.

AK3.05 Shift of T-ave. sensors to the loop with the highest flow Not applicable to unit 2 2.8*/3.0*

design/operation.

AA1.15 High-power/low-flow reactor trip block status lights Not applicable to unit 2 3.5*/3.6*

design/operation.

AA1.16 Low-power reactor trip block status lights Not applicable to unit 2 3.2*/3.5*

design/operation.

AA1.19 Power transfer confirm lamp Not applicable to unit 2 2.9*/3.0*

design/operation.

Page 1 of 5

Suppressed K/As Printed: 03/27/2006 Facility: Arkansas Nuclear One - Unit 2 IMPORTANCE Basis RO / SRO 033 Loss of Intermediate Range Nuclear Instrumentation (complete system suppressed)

AK1.01 Effects of voltage changes on performance Not applicable to Unit 2 2.7/3.0 design.

AK2.01 Power supplies, including proper switch position Not applicable to Unit 2 2.4/2.9 design.

AK2.02 Sensors and detectors Not applicable to Unit 2 2.3/2.6 design.

AK3.01 Termination of startup following loss of intermediate-range instrumentation Not applicable to Unit 2 3.2/3.6 design.

AK3.02 Guidance contained in EOP for loss of intermediate-range instrumentation Not applicable to Unit 2 3.6/3.9 design.

AA1.01 Power-available indicators in cabinets or equipment drawers Not applicable to Unit 2 2.9/3.1 design.

AA1.02 Level trip bypass Not applicable to Unit 2 3.0/3.1 design.

AA1.03 Manual restoration of power Not applicable to Unit 2 3.0*/3.2*

design.

AA2.01 Equivalency between source-range, intermediate-range, and power-range channel Not applicable to Unit 2 3.0/3.5 readings design.

AA2.02 Indications of unreliable intermediate-range channel operation Not applicable to Unit 2 3.3/3.6 design.

AA2.03 Indication of blown fuse Not applicable to Unit 2 2.8/3.1 design.

AA2.04 Satisfactory overlap between source-range, intermediate-range and power-range Not applicable to Unit 2 3.2/3.6 instrumentation design.

AA2.05 Nature of abnormality, from rapid survey of control room data Not applicable to Unit 2 3.0*/3.1?

design.

AA2.06 Cause of failure of an intermediate-range channel Not applicable to Unit 2 2.3/2.8*

design.

AA2.07 Confirmation of reactor trip Not applicable to Unit 2 3.9/4.2 design.

AA2.08 Intermediate range channel operability Not applicable to Unit 2 3.3/3.4 design.

AA2.09 Conditions which allow bypass of an intermediate-range level trip switch Not applicable to Unit 2 3.4*/3.7*

design.

AA2.10 Tech-Spec limits if both intermediate-range channels have failed Not applicable to Unit 2 3.1/3.8 design.

AA2.11 Loss of compensating voltage Not applicable to Unit 2 3.1/3.4 design.

AA2.12 Maximum allowable channel disagreement Not applicable to Unit 2 2.5*/3.1*

design.

AA2.13 Testing required if power lost, then restored Not applicable to Unit 2 2.2*/2.8*

design.

Page 2 of 5

Suppressed K/As Printed: 03/27/2006 Facility: Arkansas Nuclear One - Unit 2 IMPORTANCE Basis RO / SRO 065 Loss of Instrument Air AK3.07 Backup of compressor cooling water Not applicable to unit 2 2.3*/2.5*

system design.

AA1.01 Remote manual loaders Not applicable to unit 2 2.7*/2.5 system design.

AA1.04 Emergency air compressor Not applicable to unit 2 3.5*/3.4*

system design.

AA2.07 Whether backup nitrogen supply is controlling valve position Not applicable to unit 2 2.8*/3.2*

system design.

025 Ice Condenser System (complete system suppressed)

K1.01 Containment ventilation Not Applicable to Unit 2.7*/2.7*

2 Design.

K1.02 Refrigerant systems Not Applicable to Unit 2.7*/2.7*

2 Design.

K1.03 Containment sump system Not Applicable to Unit 3.2*/3.0*

2 Design.

K2.01 Containment ventilation fans and dampers Not Applicable to Unit 2.2*/2.7*

2 Design.

K2.02 Refrigerant systems Not Applicable to Unit 2.0*/2.5*

2 Design.

K2.03 Isolation valves Not Applicable to Unit 2.0*/2.2*

2 Design.

K3.01 Containment Not Applicable to Unit 3.8*/3.8*

2 Design.

K4.01 Glycol expansion tank levels and ice condenser system containment isolation valves Not Applicable to Unit 2.2*/2.5*

2 Design.

K4.02 System control Not Applicable to Unit 2.8*/3.0*

2 Design.

K5.01 Relationships between pressure and temperature Not Applicable to Unit 3.0*/3.4*

2 Design.

K5.02 Heat transfer Not Applicable to Unit 2.6*/2.8*

2 Design.

K5.03 Gas laws Not Applicable to Unit 2.4*/2.8*

2 Design.

K6.01 Upper and lower doors of the ice condenser Not Applicable to Unit 3.4*/3.6*

2 Design.

A1.01 Temperature chart recorders Not Applicable to Unit 3.0*/3.0*

2 Design.

A1.02 Glycol expansion tank level Not Applicable to Unit 2.5*/2.2*

2 Design.

A1.03 Glycol flow to ice condenser air handling units Not Applicable to Unit 2.5*/2.5*

2 Design.

A2.01 Trip of glycol circulation pumps Not Applicable to Unit 2.2*/2.7*

2 Design.

A2.02 High/low floor cooling temperature Not Applicable to Unit 2.7*/2.5*

2 Design.

Page 3 of 5

Suppressed K/As Printed: 03/27/2006 Facility: Arkansas Nuclear One - Unit 2 IMPORTANCE Basis RO / SRO A2.03 Opening of ice condenser doors Not Applicable to Unit 3.0*/3.2*

2 Design.

A2.04 Containment isolation Not Applicable to Unit 3.0*/3.2*

2 Design.

A2.05 Abnormal glycol expansion tank level Not Applicable to Unit 2.5*/2.7*

2 Design.

A2.06 Decreasing ice condenser temperature Not Applicable to Unit 2.5*/2.7*

2 Design.

A3.01 Refrigerant system Not Applicable to Unit 3.0*/3.0*

2 Design.

A3.02 Isolation valves Not Applicable to Unit 3.4*/3.4*

2 Design.

A4.01 Ice condenser isolation valves Not Applicable to Unit 3.0*/2.7*

2 Design.

A4.02 Containment vent fans Not Applicable to Unit 2.7*/2.5*

2 Design.

A4.03 Glycol circulation pumps Not Applicable to Unit 2.2*/2.2*

2 Design.

026 Containment Spray System (CSS)

A2.05 Failure of chemical addition tanks to inject ANO-2 chemical addition 3.7/4.1 is a passive system.

027 Containment Iodine Removal System (CIRS) (complete system suppressed)

K1.01 CSS System not applicable to 3.4*/3.7*

unit 2.

K2.01 Fans System not applicable to 3.1*/3.4*

unit 2.

K5.01 Purpose of charcoal filters System not applicable to 3.1*/3.4*

unit 2.

A2.01 High temperature in the filter system System not applicable to 3.0*/3.3*

unit 2.

A4.01 CIRS controls System not applicable to 3.3*/3.3*

unit 2.

A4.02 Remote operation and handling of iodine filters System not applicable to 2.8*/3.0*

unit 2.

A4.03 CIRS fans System not applicable to 3.3*/3.2*

unit 2.

A4.04 Filter temperature System not applicable to 2.8*/2.9*

unit 2.

Page 4 of 5

Suppressed K/As Printed: 03/27/2006 Facility: Arkansas Nuclear One - Unit 2 IMPORTANCE Basis RO / SRO 086 Fire Protection System (FPS)

A1.02 Fire water storage tank level Not applicable to ANO 3.0*/3.2*

fire system design 103 Containment System K1.03 Shield building vent system Not applicable to Unit 2 3.1*/3.5*

design.

K1.07 Containment vacuum system Not applicable to Unit 2 3.5*/3.7*

design.

K4.01 Vacuum breaker protection Not applicable to Unit 2 3.0*/3.7*

design.

A2.03 Phase A and B isolation Not applicable to unit 2 3.5*/3.8*

design.

A4.02 Excess letdown divert valves to reactor coolant drain tank Not applicable to unit 2 2.1*/2.2*

design.

A4.03 ESF slave relays Not applicable to unit 2 2.7*/2.7*

design.

A4.04 Phase A and phase B resets Not applicable to unit 2 3.5*/3.5*

design.

A4.05 PDP speed controller Not applicable to Unit 2 2.4*/2.2*

design.

A4.09 Containment vacuum system Not applicable to Unit 2 3.1*/3.7*

design.

Generic 2.2 Equipment Control 2.2.3 (multi-unit) Knowledge of the design, procedural, and operational differences between Single unit license. 3.1/3.3 units.

2.2.4 (multi-unit) Ability to explain the variations in control board layouts, systems, Single unit license. 2.8/3.0*

instrumentation and procedural actions between units at a facility.

Page 5 of 5

ES-401 Record of Rejected K/As Form ES-401-4 Tier / Randomly Reason for Rejection Group Selected K/A 008 Pressurizer PORV or PORV isolation valves are not applicable to Unit 2 Vapor Space design. A manual random sample of the remaining A2 statements Accident was performed.

AA2.05 - PORV AA2.09 - PZR spray block valve controls and indicators was isolation selected as a replacement.

  • 008 Pressurizer Vapor Space Unable to develop a credible question for this KA. Re-sampled Accident from the remaining AA2 statements.

AA2.09 - PZR spray AA2.19 - PZR Spray valve failure, using plant parameters was block valve controls selected as a replacement.

and indicators RO Exam Tier 1 027 Pressurizer Group 1 Pressure Control This generic KA was also selected for the 026 Loss of CCW System Malfunction system. A manual random sample of the remaining Generic, Category 4 statements was performed.

Generic 2.4.6 -

Knowledge of Generic 2.4.11 - Knowledge of abnormal condition procedures symptom based EOP was selected as a replacement.

mitigation strategies.

062 Loss of Nuclear The random sample was unusually heavy on the Category 4 Svc Water generic K/As between the RO/SRO exams combined. A manual Generic 2.4.39 - random sample of the remaining 3 Generic categories was Knowledge of the performed for balance.

ROs responsibilities Generic 2.1.28 - Knowledge of the purpose and function of major in emergency plan system components and controls was selected as a replacement.

implementation.

PORVs are not applicable to Unit 2 design for PZR over-pressure RO Exam 010 Pressurizer protection or control. A manual random sample of the remaining Tier 2 Pressure Control K2 statements was performed.

Group 1 K2.03 K2.01 - PZR heaters was selected as a replacement.

061 This knowledge statement is not applicable to unit 2 design. Unit 2 Auxiliary/Emergency does not have a diesel engine driven EFW/AFW pump. A manual Feedwater random sample of the remaining K1 statements for this system was performed.

K1.10 - Diesel fuel oil K1.09 - PRMS was selected as a replacement.

  • 061 Difficulty in developing a credible exam question for this KA. A Auxiliary/Emergency manual random sample of the remaining K1 statements for this Feedwater system was performed.

K1.09 - PRMS K1.07 - Emergency Water Source was selected as a replacement.

NUREG-1021, Revision 9

  • Indicates KA rejected after initial outline submittal. Page 1 of 3

Tier / Randomly Reason for Rejection Group Selected K/A 076 Service Water This knowledge statement is not applicable to unit 2 design. This K4.03 - Automatic description of CCW/SW operation is more related to Westinghouse opening features design. A random sample of the remaining K4 statements for this associated with SWS system was performed.

isolation valves to K4.02 - Automatic start features associated with SWS pump CCW heat controls was selected as a replacement.

exchangers There is no direct tie between the Instrument Air system and 078 Instrument Air control rod programming so a credible question could not be Generic 2.2.33 - created. A random sample of the remaining Generic, Category 2 Knowledge of statements was performed.

control rod Generic 2.2.27 - Knowledge of the refueling process was selected programming as a replacement to be used with the 078 Instrument Air System.

  • 078 Instrument Air Unable to match KA to 10CFR50.41 criteria. Re-sampled Generic Category 2.

Generic 2.2.27 -

Knowledge of the 2.2.13 - Knowledge tagging and clearance procedures was refueling process selected as a replacement.

Tie to KA not clear on initial submittal. Rejected question and

  • 103 Containment KA. No more unused applicable KAs or importance values >2.5 in the A2 category. Randomly sampled categories and then re-A2.05 Emergency sampled KAs in the new category.

Containment Entry A3.01 - Containment Isolation was selected as a replacement.

Tie to KA not clear on initial submittal. Unable to develop a

  • 033 Spent Fuel credible exam question due to system design at ANO-2. Rejected Pool Cooling question and KA. Re-sampled K1 statements.

K1.02 RHRS K1.05 RWST was selected as a replacement.

035 Steam PORVs are not applicable to Unit 2 design for SG/S over-pressure Generator protection or control. A manual random sample of the remaining K6 statements for this system was performed.

K6.02 - secondary PORV K6.01 - MSIVs was selected as a replacement.

056 Condensate RO Exam This K/A was randomly selected in 3 different Tiers between the Generic 2.3.10 - RO and SRO exams. A manual random sample of the remaining Tier 2 Ability to perform Generic, Category 3 statements was performed.

Group 2 procedures to reduce Generic 2.3.1 - Knowledge of 10CFR20 and related facility excessive levels of radiation control requirements was selected as a replacement.

radiation This knowledge statement is not applicable to unit 2 design. There were no other applicable K4 statements for this system with an Importance Rating of > 2.5. A manual random sample of 079 Station Air remaining Group 2 systems was performed and then a random K4.01 - Cross- sample of the remaining K/A statements.

connect with IAS 017 In-Core Temperature Monitor system, A3.01 - Indications of normal, natural, and interrupted circulation of RCS was selected as a replacement.

NUREG-1021, Revision 9

  • Indicates KA rejected after initial outline submittal. Page 2 of 3

Tier / Randomly Reason for Rejection Group Selected K/A 007 Reactor Trip -

This generic knowledge statement was also selected for Tier 3. A Stabilization -

manual random sample of the remaining Generic, Category 1 SRO Exam Recovery statements was performed.

Tier 1 Generic 2.1.11 -

Generic 2.1.19 - Ability to use plant computer to obtain and Group 1 Knowledge of less evaluate parametric information on system or component status than one hour TS was selected as a replacement.

actions 003 Reactor Coolant There is no direct tie between the Reactor Coolant Pump system Pump and this generic knowledge statement so a credible question could not be created. A random sample of the remaining Generic, Generic 2.2.28 -

Category 2 statements was performed.

Knowledge of new and spent fuel Generic 2.2.18 - Knowledge of the process for managing movement maintenance activities during shutdown operations was selected as procedures a replacement to be used with the 003 RCP system.

  • 003 Reactor Coolant Pump Generic 2.2.18 - Question on initial submittal was at too low of responsibility level SRO Exam Knowledge of the (RO). Rejected and re-sampled from Generic Category 2.

Tier 2 process for Generic 2.2.29 - Knowledge of SRO fuel handling responsibilities Group 1 managing was selected as a replacement to be used with the 003 RCP system.

maintenance activities during shutdown operations

  • 026 Containment Spray Question on initial submittal was at too low of responsibility level Generic - 2.4.46 (RO). Rejected and re-sampled from Generic Category 4.

Ability to verify that Generic 2.4.17 - Knowledge of EOP terms and conditions was the alarms are selected as a replacement to be used with the 003 RCP system.

consistent with the plant conditions.

071 Waste Gas The random sample was unusually heavy on the Category 4 Disposal generic K/As between the RO/SRO exams combined. A manual SRO Exam Generic 2.4.37 - random sample of the remaining 3 Generic categories was Tier 2 Knowledge of the performed for balance.

Group 2 lines of authority Generic 2.3.8 - Knowledge of process for performing a planned during an gaseous radioactive release was selected as a replacement.

emergency.

Generic This K/A was randomly selected in 3 different Tiers between the SRO Exam RO and SRO exams. A manual random sample of the Generic 2.3.10 - Ability to Tier 3 Category 1 and 3 statements was performed.

perform procedures (Generic) to reduce excessive Generic 2.1.30 - Ability to locate and operate components, levels of radiation including local controls was selected as a replacement.

NUREG-1021, Revision 9

  • Indicates KA rejected after initial outline submittal. Page 3 of 3
Retrieved from "https://kanterella.com/w/index.php?title=ML062070582&oldid=2583694"