ML100410119

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Initial Exam 2009-302 Draft Administrative Documents
ML100410119
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 01/05/2010
From:
NRC/RGN-II
To:
References
IR-09-302
Download: ML100410119 (27)
v • d • e


Text

FACILITY NAME: Harris Nulear Plant Section 1 REPORTNUMBER: ____ ____~O~5~O~OO~4~O~O/~2~OO~9~-3~O~2~

~O~5~OO~O~40~O~~~O~O~9-~30~2~_____

DRAFT ADMINISTRATIVE DOCUMENTS CONTENTS:

~~rft Written Exam sample plan (ES-401-1/2) ciVBraft Administrative Topics Outline (ES-301-1) c¥Oraft Draft Control Room Systems & & Facility Walk-Through Test Outline (ES-301-2)

Location of Electronic Files:

Submitted By: ~~=::::!....-

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FT ES-301 Administrative Topics Outline Form ES-301-1

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Facility: Harris Nuclear Plant Date of Examination: 11/30/2009 Examination Level: RO

  • SRO 0 D Operating Test Number: 05000400/2009302 Administrative Topic Type Describe activity to be performed (see Note) Code*

Response to Voids In the Reactor Vessel Vessel-- Calculate Conduct of Operations Reactor Vessel Maximum Vent Time. (JPM ADM-096)

M,R KIA G2.1.20 20098 NRC RO A1-1 2009B Perform A Manual Shutdown Margin Calculation Conduct of Operations (JPM-ADM-019)

M,R KIA G2.1.25 20098 NRC RO A1-2 2009B Perform OP-111, Att. 3 Low Head SI Standby Lineup Equipment Control Checklist. (JPM ADM-024)

( M,S KIA G2.2.15 20098 NRC RO A2 2009B Using Survey Maps, Simplified Drawings, Plant Maps and valve lists, determine stay times while Radiation Control N,R performing a clearance activity (JPM ADM-1 00)

KIA G2.3.4 20098 NRC RO A3 2009B NOT SELECTED FOR RO Emergency Procedures/Plan ProcedureslPlan N/A 20098 NRC RO A4 2009B 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 (:5 3 for ROs; :5 4 for SROs & RO retakes) (0)

(N)ew or (M)odified from bank (~ 1) (4)

(P)revious 2 exams (:5 1; randomly selected) (0)

20098 NRC RO Admin JPM Summary

(

20098 NRC RO A1 Response to Voids In Reactor Vessel - Calculate Reactor Vessel Maximum Vent Time (JPM ADM-096) Modified KIA EPE E10 EK3.2 - Natural Circulation with Steam Void in Vessel with/without RVLlS RVL/S - Normal, abnormal and emergency operating procedures associated with (Natural Circulation with Steam Void in RVLlS). RO 3.2/ SRO 3.7 Vessel withlwithout RVL/S).

The plant was at 100 percent power when a Small Break LOCA occurred. The RCP have been secured, SI terminated, RVLlS Upper Range reads 90 percent, PZR level is 96 percent, RCS Pressure is 1750 psig, Containment Temp is 181°F and Containment Hydrogen concentration is 1.72 percent. Based on these conditions a void has formed in the Reactor Vessel. The crew has completed preparing the Containment for Reactor Vessel venting.

The candidate is required to calculate the maximum time the Reactor Vessel should be vented, using Attachment 1 of FRP-I-3, Response To Voids In Reactor Vessel. The calculation should be determined to the second to allow timing using the MCR timer.

This JPM has been modified by changing all values provided as initial conditions. The answer is now over one minute different than the original version with a tolerance

~ 5 seconds.

of .:!:.

20098 NRC RO A1 Perform A Manual Shutdown Margin Calculation (JPM-ADMIN-019)

Modified KIA G2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc.

( (CFR: 41.10/43.5/45.12) RO 3.9 SRO 4.2 The plant is operating at 50% power and the CRS will direct the candidate to complete OST-1036, Shutdown Margin Calculation Modes 1-5, Section 7.3, for the current plant conditions.

This JPM has been modified by changing all values provided as initial conditions and using current cycle curves will yield a different value of Shutdown margin.

20098 NRC RO A2 (Common) - Perform OP-111 , Att. 3 Low Head SI Standby Lineup Checklist. (JPM ADM-024) Modified KIA G2.2. 15 - Knowledge of the process for controlling equipment configuration or status.

(CFR: 41.10/43.3/45.13) RO 3.9 SRO 4.3 NOTE: The Simulator will be utilized for this Admin JPM.

The initial setup will have the plant in Mode 4 with the RHR System aligned for ECCS operation. Several of the components will be intentionally mispositioned for this lineup. The applicant will be assigned to perform an independent verification of OP-111, Residual Heat Removal System, Attachment 3 "Low Head Safety Injection Standby Lineup Checklist".

They will be expected to identify all errors with the lineup and list the mis-positioned controls and valves in the remarks section of the procedure.

This JPM has been modified by selecting different mis-aligned components and not using any of the components that were originally selected.

20098 NRC RO Admin JPM Summary (continued) 2009B 20098 NRC RO A3 (Common) - Using Survey Maps, Simplified Drawings, Plant Maps and 2009B ADM-1 00) New valve lists, determine stay times while performing a clearance activity (JPM ADM-100)

KIA G2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12/43.4 //45.10) 45.10) RO 3.2 SRO 3.7 The applicant will be supplied a survey map of a location in the RAB and a clearance mission to complete in this radioactive area. The location also contains one or more hot spots. They must determine the individual stay times for themselves and another Auxiliary Operator (AO) without exceeding the annual administrative dose limits. They will be provided Survey Maps, Simplified plant drawings to locate valves, Plant Maps of the area and a plant valve list to determine the location of the valves they will be hanging a clearance on. The The given information will supply the accumulated annual whole body doses for themselves and the other AO who has recently worked for another utility. They must perform their calculations based on Progress Energy Administrative Dose Limits.

20098 NRC RO A4 - NOT SELECTED FOR RO 2009B

(

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

( Facility: Harris Nuclear Plant Date of Examination: 11130/2009 Examination Level: RO ROO D SRO

  • Operating Test Number: 05000400/2009302 Administrative Topic Type Describe activity to be performed (see Note) Code*

During a 1055 loss of shutdown cooling, determine the time Conduct of Operations that the RCS will reach core boiling and core boil-off N,R conditions (JPM ADM-OOS)

ADM-005)

KIA G2.1.20 20098 NRC SRO A1-1 Determine Boric Acid Addition Following Control Room Evacuation (JPM IP-049)

Conduct of Operations M,R KIA G2.1.25 20098 NRC SRO A1-2 Perform OP-111, Att. AU. 3 Low Head SI Standby Lineup

( Equipment Control M,S Checklist. (JPM ADM-024)

KIA G2.2.15 20098 NRC SRO A2 Using Survey Maps, Simplified Drawings, Plant Maps and valve lists, determine stay times while Radiation Control N,R performing a clearance activity (JPM ADM-100) ADM-1 00)

KIA G2.3.4 20098 NRC SRO A3 Given a set of conditions, Classify an Event Emergency Procedures/Plan (JPM ADM-099)

N,R KIA G2.4.41 20098 NRC SRO A4 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 (S(:S: 3 for ROs; S :s: 4 for SROs & RO retakes) (0)

(N)ew or (M)odified from bank (;:: (~ 1) (5)

(P)revious 2 exams (S (:S: 1; randomly selected) (0)

20098 NRC SRO Admin JPM Summary 2009B

(

20098 NRC SRO A1 During a loss of shutdown cooling, determine the time that the RCS will reach core boiling and core boil-off conditions. (SRO JPM ADM-005) New KIA G2.1.20 - Ability to interpret and execute procedure steps.

(CFR: 41.10/43.5/45.12) RO 4.6 SRO 4.6 The applicant will be provided with initial plant conditions. A plant shutdown for refueling is in progress with the Reactor Vessel head off when a loss of RHR has occurred. The crew is implementing AOP-020, Loss of RCS Inventory or Residual Heat Removal While Shutdown.

The SRO applicants must first determine which of the four plant curves to use (H-X-B through H-X-11) and then calculate the time the RCS will reach core boiling and core boil-off.

20098 NRC SRO A1 Determine Boric Acid Addition Following Control Room Evacuation (JPM IP-049) Modified KIA G2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10/43.5/45.12) RO 3.9 SRO 4.2 The Control Room has been evacuated and the MCB transfer to the ACP has been completed. Plant management has directed a plant cooldown to mode 5 utilizing AOP-004.

Given an OST-1036 cold shutdown boron requirement, the candidate must use curves to calculate gallons of Boric Acid and change in Boric Acid Tank level to complete section 3.2 step 25.

((

This JPM has been modified by providing new initial conditions which leads to a new answer that is substantially different than the original.

20098 NRC SRO A2 (Common) - Perform OP-111, Att. 3 Low Head SI Standby Lineup Checklist. (JPM ADM-024) Modified G2.2. 15 - Knowledge of the process for controlling equipment configuration or status.

KIA G2.2.

(CFR: 41.10/43.3/45.13) RO 3.9 SRO 4.3 NOTE: The Simulator will be utilized for this Admin JPM.

The initial setup will have the plant in Mode 4 with the RHR System aligned for ECCS operation. Several of the components will be intentionally mispositioned for this lineup. The applicant will be assigned to perform an independent verification of OP-111, Residual Heat Removal System, Attachment 3 "Low Head Safety Injection Standby Lineup Checklist".

They will be expected to identify all errors with the lineup and list the mis-positioned controls and valves in the remarks section of the procedure.

This JPM has been modified by selecting different mis-aligned components and not using any of the components that were originally selected.

(

20098 NRC SRO Admin JPM Summary (continued)

( 2009B NRC SRO A3 (Common) - Using Survey Maps, Simplified Drawings, Plant Maps and valve lists, determine stay times while performing a clearance activity. (JPM ADM-1ADM-100)00)

KIA G2.3.4 - Know/edge Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12/43.4 /45.10) RO 3.2 SRO 3.7 The applicant will be supplied a survey map of a location in the RAB and a clearance mission to complete in this radioactive area. The location also contains one or more hot spots. They must determine the individual stay times for themselves and another Auxiliary Operator (AO) without exceeding the annual administrative dose limits. They will be provided Survey Maps, Simplified plant drawings to locate valves, Plant Maps of the area and a plant valve list to determine the location of the valves they will be hanging a clearance on. The given information will supply the accumulated annual whole body doses for themselves and the other AO who has recently worked for another utility. They must perform their calculations based on Progress Energy Administrative Dose Limits.

2009B NRC SRO A4 - Given a set of conditions, Classify an Event (JPM ADM-099) New KIA G2.4.41 Knowledge of the emergency action level thresholds and classifications (CFR: 41.10/43.5/45.11) RO 2.9 SRO 4.6 Given a set of initial conditions and the EAL Flow Path, the candidate must classify the

( appropriate Emergency Action Level for the event in progress.

(

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2

(

Facility: Shearon Harris Date of Examination: 11/30/2009 Exam Level: RO SRO-I SRO(U) Operating Test No.: 05000400/2009302 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

/ JPM Title System 1 Type Code* Safety Function

a. Continuous Withdrawal of a Control Bank (at power) A, D,S 1 (AOP-001, Attachment 1) (JPM CR-048)

KIA APE 003 AA1.05

b. Malfunction of RMU Control A,N,S 2 (AOP-003) (JPM CR-237)

KIA 004A4.13 Loss Of Service Water (Line Break)

c. Loss N,S 45 4S (AOP-022) (JPM CR-238)

KIA APE 062 AK3.03

d. Decreasing CCW Surge Tank Level A, D,S 8

( (AOP-014) (JPM CR-044)

KIA APE 026 AA1.05

e. Respond to High RCS Pressure at Power A, D,S 3 (AOP-019) (JPM CR-051)

KIA APE 027 AA2. 16 RO ONLY

f. Loss of All AC A,D, 6 (EOP-EPP-001) (JPM CR-059) (EN), S (EN),S KIA APE 056 AK3.02
g. Startup the RHR System D, 0, L, S 4P (OP-111) (JPM CR-023)

KIA 005 A4.01

h. Place an Excore NI Channel Out Of Service at Power D,S 7 (OWP-RP-26) (JPM CR-019) 015 A4.03 KIA 015A4.03

{

\.

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Shearon Harris Date of Examination: 11/30/2009 Exam Level: RO SRO-I SRO(U) Operating Test No.: 05000400/2009302 In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Restore Power To An Emergency Bus E,EN,N 6 (OP-155) (JPM IP-238)

KIA 068AA1.10 068 AA 1.10

j. Transfer Control To the ACP D, E, L 7 (AOP-004) (JPM IP-050)

KIA 068 AA 1.21

k. Torque Shut The veT VeT Outlet Valves A,D,E,R 2 (AOP-017) (JPM IP-212)
1. 06 KIA 004 A 1.06

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

(

  • Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-6 I 4-6 12-3 4-6/4-6 (6/5/3)

(C)ontrol room (D)irect from bank  ::;9/::;8/::;4

~9/~8/~4 (8/712)

(8/7/2)

(E)mergency or abnormal in-plant  ;::1/;::1/;::1

<:1/<:1/<:1 (3/312)

(EN)gineered safety feature - I - I ;::1

<:1 (2/2/2)

(L)ow-Power I Shutdown  ;::1/;::1/;::1

<:1/<:1/<:1 (2/212)

(N)ew or (M)odified from bank including 1(A)  ;::2/;::2/;::1

<:2/<:2/<:1 (3/313)

(P)revious 2 exams  ::;3/::;3/::;2

~3/~3/~2 (01010)

(R)CA  ;::1/;::1/;::1

<:1/<:1/<:1 (1/111)

(S)imulator

(

2009a NRC Control Room/In-Plant JPM Summary JPM a - Continuous Withdrawal of a Control Bank (JPM CR-048) Alternate Path KIA APE 003 AA1.05 AA 1.05 - Ability to operate and 1 I or monitor the following as they apply to the Continuous Rod Withdrawal: Reactor trip switches (CFR 41.7145.5145.6)

(CFR41.7 145.5145.6) RO 4.3 SRO 4.2 The unit will be operating at -50% power steady state conditions with the applicant maintaining current plant conditions. A failure of Tref will cause the rod control system to step out at maximum speed (72 steps per minute). The response to this condition will be to perform the immediate actions of AOP-001, Malfunction of Rod Control and Indication System. This will consist of a verification of less than 2 rods dropped (YES), then positioning the rod bank selector switch to manual. The next immediate action is to check that control bank motion has stopped. A second failure in the rod control system will cause the rods to continue to step out after they are placed in manual. The RNO will be to TRIP the Reactor and GO TO EOP Path-1.

The applicant then performs the immediate actions of Path-1, verifies the Reactor and Turbine is tripped, SI is not required and power to the Emergency busses.

JPM b - Malfunction of Rx Makeup Control (JPM CR-237) Alternate Path - New KIA 004 A4.13 - Ability to manually operate andlor monitor in the control room: VCT level control and pressure control (CFR: 41.7145.5 to 45.8) RO 3.31 SRO 2.9

(

With the unit operating at 100% power steady state conditions, a VCT makeup was required when level reached the low level auto makeup setpoint of 20%. The makeup system malfunctioned and a makeup did not occur. When the operators attempted a manual makeup the Reactor Makeup Mode Selector switch stayed in the STOP position. AOP-003, Malfunction of Reactor Makeup Control was entered and the crew has performed steps 1-14 of section 3.2. The applicant will be directed to continue from this point. This will require the applicant to select from the procedure table what attachment to perform from the given conditions. After making the selection (Attachment 2) the applicant will have to calculate the amount of flow for a local manual makeup to the VCT based on current RCS boron concentration from the status board. They will then need to perform a lineup on the MCB and start a Boric Acid pump. Next they will have to coordinate the actions of a local operator to throttle open boration and dilution valves to the correct positions based on MCR indications until VCT level has reached 40% (normal full auto makeup setpoint).

2009a NRC Control Roomlln-Plant JPM Summary (continued)

(

JPM c - Loss of Service Water Line (line break) (JPM CR-238) New KIA APE 062 AK3. 03 Knowledge of the operational implications of the following concepts as they apply to Loss of Instrument Air: Guidance actions contained in EOP for Loss of nuclear service water (CFR 41.8/41.10/45.3) RO 4.0 SRO 4.2 With the unit operating at 100% power steady state conditions a large service water leak occurs. The applicant will identify that the condition by MCB annunciators and notification from the RAB NLO. The applicant will be expected to verbalize the immediate actions of AOP-022, Loss of Service Water. They will be challenged with determining the appropriate section the procedure to transition into and then continue with the identification of the leak. They will be required to secure the running ESW pump on the ruptured header and maintain the pump in STOP until the discharge pressure is low enough to lock out the auto start feature.

JPM d - Decreasing CCW Surge Tank Level (JPM CR-044) Alternate Path KIA APE 026 AA 1.05 - Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: The CCWS surge tank, including level control and level alarms, and radiation alarm (CFR 41.7/45.5/45.6) RO 3.1/ 3.1 / SRO 3.1

( With the plant operating at 100% steady state the applicant will respond to a computer alarm for the Component Cooling Water systems. When the alarm is checked the applicant will identify that the CCW surge tank level is lower than normal. Soon after the first indication additional alarms will actuate indicating that the tank level is continuing on a lowering trend. The applicant is expected to recognize entry conditions for AOP-014, Loss of Component Cooling Water are met.

The applicant will work through the procedure in attempts to maintain surge tank level and the running CCW pump in service. The leak size is large enough that level will not be maintained. lAW AOP-014 they must isolate charging flow and willnot subsequently manually trip the Reactor.

JPM e - Respond to High RCS Pressure (JPM CR-051) RO ONLY KIA APE 027 AA2. 16 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: Actions to be taken if PZR pressure instrument fails low (CFR: 43.5/45.13) RO 3.6 SRO 3.9 With the plant operating at 100% steady state the applicant will respond to increasing RCS pressure which will be identified by MCB annunciators and Pressurizer pressure indications. The applicant is expected to recognize entry conditions are met for AOP-019, Malfunction of RCS Pressure Control and perform the immediate actions. This will require RNO actions to manually control the Pressurizer spray

( valves and control RCS pressure prior to reaching the auto Reactor protection actuation setpoint of 2385 psig.

2009a NRC Control Room/In-Plant Roomlln-Plant JPM Summary (continued)

JPM f - Loss of All AC Power- (JPM-CR-059) Alternate Path KIA APE 056 AK 3.02 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Actions contained in EOP for loss of offsite power (CFR 41.5,41.10/45.6/45.13) RO 4.4 SRO 4.7 With the plant at 100% power the applicant will be directed to maintain current plant conditions. Subsequently the Reactor will trip on a loss of all AC power. The applicant will be expected to perform the immediate actions associated with EOP-EPP-001, Loss of AC Power to 1A-SA and 1B-SB Buses. During the performance of this emergency procedure they will have to manipulate components that have failed to go to their expected position. They will also have to increase AFW flow to meet the minimum required flow rate and adjust the TDAFW pump speed control to obtain flow. All SG levels will be lower than required and flow will have to continue until adequate levels are being established.

JPM 9 - Startup the RHR System (JPM-CR-023)

KIA 010 A 1.07 Ability to manually operate and/or monitor in the control room: Controls and indication for RHR pumps (CFR: 41.7/45.5 to 45.8) RO 3.6 SRO 3.4

( The CRS has directed the applicant to initiate RCS cooling via the Train 'A' RHR pump per OP-111, Residual Heat Removal System Section 5.1. All prerequisites and initial conditions are met. The plant is being cooled down using Steam Dumps lAW GP-007, Normal Plant Cooldown Mode 3 to Mode 5. RCS temperature is 330°F and pressure is 325#. The applicant will start the 'A' RHR pump and place it in service to provide RCS cooling.

JPM h - Place an Excore NI Channel Out Of Service at Power (JPM CR-019)

KIA 015 A4. 03 Ability to manually operate and/or monitor in the control room: Trip bypasses (CFR: 41.7/45.5 to 45.8) RO 3.8 SRO 3.9 Prior to taking watch, with the plant operating at 100% power steady state conditions, Nuclear Instrument 44 has failed low. The CRS has directed the applicant to remove NI-44 from service lAW OWP-RP-26, Reactor Protection. This will require placing rod control to manual. The applicant will then remove the detector from service at the detector current comparator drawer, the miscellaneous control and indication panel, and the comparator and rate drawer. Then contact I&C to lift leads from the circuit. They will then check the bistable status panels for proper responses. They will also have to log onto the ERFIS computer and remove the channel from scan.

After removing the channel from scan they CRS will direct them to either place rod control in auto or remain manual control.

(

2009a NRC Control Room/In-Plant JPM Summary (continued)

(

JPM i-Restore Power to an Emergency Bus (JPM IP-238) New KIA 068 AA 1.10 Ability to operate and / or monitor the following as they apply to the Control Room Evacuation: Power distribution: ac and dc de (CFR 41.7/45.5/45.6) RO 3.7/ SRO 3.9 Following a Main Control Room Evacuation due to a fire, Emergency Bus 'A' is not powered. The applicant will be directed to start and load 'A' Emergency Diesel locally lAW OP-155, Diesel Generator Emergency Power System section 8.13 and 8.14.

JPM i-Transfer Control to the ACP (JPM-IP-050)

KIA 068 AA 1.21 Ability to operate and / or monitor the following as they apply to the Control Room Evacuation: Transfer of controls from control room to shutdown panel or local control 41.7 / 45.5/ 45.6) RO 3.9/ SRO 4.1 (CFR 41.7/45.5/45.6)

The applicant informed that the plant was in Hot Standby at 557°F when a fire stared in the Main Control Room. The Shift Manager has directed that the Control Room be evacuated. The Reactor has been tripped and the CRS has relocated to the ACP.

The applicant will be directed by the CRS to perform a transfer of control to the ACP.

( NOTE: This is a time critical JPM based on AOP-004 caution prior to the step for transferring control to the ACP - "Transfer to the ACP in the next two steps must be done as soon as possible to minimize spurious actuations caused by Control Room area fire. The transfer should be complete within 10 minutes."

JPM k - Torque Shut the VeT Outlet Valves (JPM IP-212) Alternate Path KIA 004 A 1.06 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: VCT level (CFR: 41.5/45.5) RO 3.0 SRO 3.2 With the plant operating at 100% steady state conditions an instrument air leak causes IA pressure to decrease. The crew has entered AOP-017, Loss of Instrument Air and air pressure continues to decrease. A Reactor Trip was required based on the continued decreasing IA pressure. The operators performed the actions of Path-1 and have transitioned to EPP-004, Reactor Trip Response while continuing the implementation of AOP-017. The RO has just identified that VCT level cannot be maintained> 5%. The RO will direct the applicant to perform the local operator actions of AOP-017 Section 3.2 step 2 RNO 2.c to realign RCP Seal return to the Charging Pump suctions and step 3 RNO to locally torque shut the VCT outlet valves. NOTE: This JPM will be performed in the RCA.

R fT FT ES-401.1 Rev. 9 ES-401 PWR Examination Outline Form ES-401*2 I Facility: HARRIS Date of Exam: 2010 I

RO KIA Category Points SRO-Only Points Tier Group K

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

2 1 1 0 1 1 1 1 1 1 1 1 10 1 2 3 Plant Systems Tier Totals 3 4 2 2 3 4 4 4 4 4 4 38 3 5 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 2 2 3 3 2 2 1 2
1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of ofthe the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" ofthe 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. topics from as many systems and evolutions as possible; sample every system or evolution Select topiCS 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 of the KIA Catalog, but the 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; ififfuel fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note
  1. 1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of ofthe the 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, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A 1 A2 A3 A4 G TOPIC:

RO SRO 007EK3.01 Reactor Trip - Stabilization - Recovery 4 4.6 D D ~ D D D D D D D D DD~DDDDDDDD Actions contained in EOP for reactor trip 1/ 1 008AK1.01 Pressurizer Vapor Space Accident 13 /3 3.2 3.7 ~ 0 D D 0 0D D0 D 0 D 0 D 0 D0 D0 D0 Thermodynamics and flow characteristics of open or leak- ing valves 009EK1.02 Small Break LOCA 1 /3 3.5 4.2 ~ 0 D D 0 D0 D0 D 0 0D D 0 D0 D0 D0 Use of steam tables 011 EG2.4.4 Large Break LOCA 1 /33 4.5 4.7 0 D D 0 D0 D0 D0 D 0 D 0 D 0 D0 D0 ~ Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

015AK1.02 RCPMaifunctions/4 RCP Malfunctions / 4 3.7 4.1 ~ 0 D D 0 D0 D0 D 0 D 0 D 0 DDODD D Consequences of an RCPS failure 022AA1.03 022AA 1.03 Loss of Rx Coolant Makeup 1 /2 3.2 3.2 D D D D D D ~ D D D D DDDDDD~DDDD PZR level trend 025AK2.03 Loss of RHR System 14 /4 2.7 2.7 D ~ D D D D D D D D D D~DDDDDDDDD Service water or closed cooling water pumps 027AK3.02 Pressurizer Pressure Control System 2.9 3 DD~DDDDDDDD Verification of alternate transmitter and/or plant computer Malfunction 1/3 prior to shifting flow chart transmitters 029EA2.02 ATWS/1 ATWS /1 4.2 4.4 DDDDDDD~DDD Reactor trip alarm 038EG2.4.4 Steam Gen. Tube Rupture 1 /3 4.5 4.7 DDDDDDDDDD~ Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

040AA2.04 Steam Line Rupture - Excessive Heat 4.5 4.7 DDDDDDD~DDD Conditions requiring ESFAS initiation Transfer 1

/4 Page 1 of 2 5/6/2009 10:05 AM

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 KS K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO OS6AA1.01 056AA1.01 Loss of Off-site Power //6 6 4 3.8 D D D D D D ~ D D D D DDDDDD~DDDD Power relief controllers to maintain no-load T -ave T-ave OS7 AG2.1.23 057AG2.1.23 Loss of Vital AC Inst. Bus //6 6 4.3 4.4 D D D D D D D D D D ~ Ability to perform specific system and integrated plant procedures during all modes of plant operation.

058AA 1.01 OS8AA1.01 Loss of DC Power //66 3.4 3.S 3.5 D D D D D D ~ D D D D altemate supply Cross-tie of the affected dc bus with the alternate 062AA2.03 Loss of Nuclear Svc Water //4 4 2.6 2.9 D D D D D D D ~ D D D The valve lineups necessary to restart the SWS while bypassing the portion of the system causing the abnormal condition 06SAK3.03 065AK3.03 Loss of Instrument Air //8 8 2.9 3.4 D D ~ D D D D D D D D Knowing effects on plant operation of isolating certain equipment from instrument air WE04EK2.1 LOCA Outside Containment / 3 3.S 3.5 3.9 D ~ D D D D D D D D D Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

WEOSEK2.2 WE05EK2.2 Inadequate Heat Transfer - Loss of 3.9 4.2 D ~ D D D D D D D D D Facility's heat removal systems, including primary Secondary Heat Sink / 4 coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

Page 2 of 2 S/6/2009 5/6/2009 10:0S 10:05 AM

ES-401, REV 9 T1 G2 PWR EXAMINATION OUTLINE T1G2 FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 001AA2.04 Continuous Rod Withdrawal Withdrawal/1/1 4.2 4.3 D D D D D D D ~ D D D Reactor power and its trend 003AK3.05 Dropped Control Rod 11

/1 3.4 4.1 D D ~ D D D D D D D D Tech-Spec limits for reduction of load to 50% power if flux cannot be brought back within specified target band 005A~ Inoperable/Stuck Control Rod 11

/1 3.8 3.9 D D D D D D D D D D D ~ (multi-unit license) Knowledge of the design, procedural G

G2.'."U)

2. ,. 'Zi> and operational differences between units.

S ...

S" .. ,k

,le qAi.,..

qit',t"'

024AK1.04 Emergency Boration 11

/1 2.8 3.6 ~ D D D D D D D D D D Low temperature limits for boron concentration 076AG2.1.25 High Reactor Coolant Activity 1

/9 3.9 4.2 D D D D D D D D D D D ~ Ability to interpret reference materials such as graphs, monographs and tables which contain performance data.

WE03EK1.1 LOCA Cooldown - Depress. 14 /4 3.4 4.0 ~ D D D D D D D D D D D Components, capacity, and function of emergency systems.

WE06EK3.3 Degraded Core Cooling 1/44 4.0 3.9 D D ~ D D D D D D D D Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

WE15EK2.1 Containment Flooding 1

/55 2.8 2.9 D ~ D D D D D D D D D Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

WE16EA 1.1 High Containment Radiation 1/9 3.1 3.2 D D D D D D ~ D D D D Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

Page 1 of 1 5/6/2009 10:05 AM

ES-401, REV 9 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 003K2.02 Reactor Coolant Pump 2.5 2.6 D~DDDDDDDDD CCW pumps 004A1.11 Chemical and Volume Control 3.0 3.0 DDDDDD~DDDD Letdown and charging flows 004K2.05 Chemical and Volume Control 2.7 2.9 D~DDDDDDDDD MOVs 005A1.02 Residual Heat Removal 3.3 3.4 DDDDDD~DDDD RHR flow rate 005K5.03 Residual Heat Removal 2.9 3.1 DDDD~DDDDDD Reactivity effects of RHR fill water 006K6.13 Emergency Core Cooling 2.8 3.1 DDDDD~DDDDD Pumps 007A1.03 Pressurizer Relief/Quench Tank 2.6 2.7 DDDDDD~DDDD Monitoring quench tank temperature 008K~ Component Cooling Water 2.6 2.7 DDD~DDDDDDD Operation of the CCW swing-bus power supply and its associated breakers and controls J( 4.01-010A3.01 Pressurizer Pressure Control 3.0 3.2 DDDDDDDD~DD PRT temperature and pressure during PORV testing 010K3.03 Pressurizer Pressure Control 4.0 4.2 DD~DDDDDDDD ESFAS 012A2.04 Reactor Protection 3.1 3.2 DDDDDDD~DDD Erratic power supply operation Page 1 of 3 5/6/2009 10:05 AM

ES-401, REV 9 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 0D ~ 0D 0D 0D 0D 0D 0D 0D 0D 0D ESFAS/safeguards equipment control 013K6.01 Engineered Safety Features Actuation 2.7 3.1 0D 0D 0D 0D 0D ~ 0D 0D 0D 0D 0D Sensors and detectors 022A3.01 Containment Cooling 4.1 4.3 0D 0D 0D 0D 0D 0D 0D 0D ~ 0D 0D Initia tion of safeguards mode of operation 026K3.01 Containment Spray 3.9 4.1 0D 0D ~ 0D 0D 0D 0D 0D 0D 0D 0D CCS 039A4.04 Main and Reheat Steam 3.8 3.9 0D 0D 0D 0D 0D 0D 0D 0D 0D ~ 0D feed water pump turbines Emergency feedwater 059A3.02 Main Feedwater 2.9 3.1 0D 0D 0D 0D 0D 0D 0D 0D ~ 0D 0D Programmed levels of the S/G 059A~ Main Feedwater 059A+.+et""" 0D 0D 0D 0D 0D 0D 0D 0D 0D ~ 0D ICS 3.9 3.8 (Ub (DO xes) 4*(2.-

4*(Z, 061A2.03 Auxiliary/Emergency Feedwater 3.1 3.4 0D 0D 0D 0D 0D 0D 0D ~ 0D 0D 0D Loss of dc power 062A2.03 AC Electrical Distribution 2.9 3.4 0D 0D 0D 0D 0D 0D 0D ~ 0D 0D 0D Consequences of improper sequencing when transferring to or from an inverter 062G2.2.36 AC Electrical Distribution 3.1 4.2 0D 0D 0D 0D 0D 0D 0D 0D 0D 0D ~ Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions of operations 063G~ DC Electrical Distribution 3.7 4.1 0D 0D 0D 0D 0D 0D 0D 0D 0D 0D ~ Knowledge of surveillance procedures.

C, f.tU Page 2 of 3 5/6/2009 10:05 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 063K1.03 DC Electrical Distribution 2.9 2.9 3.5 3.5 Ii'] D D D D D D D D D D

~DDDDDDDDDD Battery charger and battery 064K6.08 Emergency Diesel Generator 3.2 3.2 3.3 3.3 D D D D D Ii'] D D D D D DDDDD~DDDDD Fuel oil storage tanks 073K5.01 Process Radiation Monitoring 2.5 2.5 3.0 3.0 D D D D Ii'] D D D D D D DDDD~DDDDDD Radiation theory, including sources, types, units and effects 076K1.01 Service Water 3.4 3.4 3.3 3.3 Ii'] D D D D D D D D D D

~DDDDDDDDDD CCWsystem CCW system 078G~ Instrument Air 3.7 3.7 3.9 3.9 D D D D D D D D D D Ii']

DDDDDDDDDD~ Ability to identify post-accident instrumentation.

2.4.41 103A4.01 Containment 3.2 3.2 3.3 3.3 D D D D D D D D D Ii'] D DDDDDDDDD~D Flow control, pressure control and temperature control valves, including pneumatic valve controller Page 3 of 3 5/6/2009 10:05 AM

ES-401, REV 9 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 001K2.05 Control Rod Drive 3.1 3.5 D~DDDDDDDDD MIG sets 011K5.09 Pressurizer Level Control 2.6 2.7 DDDD~DDDDDD Reason for manually controlling PZR level 014G2.4.21 Rod Position Indication 4.0 4.6 DDDDDDDDDD~ Knowledge of the parameters and logic used to assess the status of safety functions 015A1.03 Nuclear Instrumentation 3.7 3.7 DDDDDD~DDDD NIS power indication 016A~ Non-nuclear Instrumentation 2.7 2.6 DDDDDDDDD~D Recorders 4.4>1 4.1>'

017A3.02 In-core Temperature Monitor 3.4 3.1 DDDDDDDD~DD Measurement of in-core thermocouple temperatures at panel outside control room 028A2.02 Hydrogen Recombiner and Purge 3.5 3.9 DDDDDDD~DDD LOCA condition and related concern over hydrogen Control 035K6.01 Steam Generator 3.2 3.6 DDDDD~DDDDD MSIVs 041K4.09 Steam DumpfTurbine Dump/Turbine Bypass Control 3.0 3.3 DDD~DDDDDDD Relationship of low/low lowllow T-ave. setpoint in SDS to primary cooldown cool down 056K1.03 Condensate 2.6 2.6 ~DDDDDDDDDD MFW Page 1 of 1 5/6/2009 10:06 AM

ES-401, REV 9 T3 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 G2.1.20 Conduct of operations 4.6 4.6 D D D D D D D D D D ~ Ability to execute procedure steps.

G2.1.38 Conduct of operations 3.7 3.8 D D D D D DOD D D D D D ~ Knowledge of the stations requirements for verbal communication when implamenting procedures G2.2.13 Equipment Control 4.1 4.3 D D D D D D D D D D ~ Knowledge of tagging and clearance procedures.

G2.2.5 Equipment Control 2.2 3.2 D D D D D D D D D D ~ Knowledge of the process for making design or operating changes to the facility G2.3.11 G'2.3.11 Radiation Control 3.8 4.3 D D D D D D D D D D ~ Ability to control radiation releases.

G2.3.5 Radiation Control 2.9 2.9 D D D D D D D D D D ~ Ability to use radiation monitoring systems G2.3.6 Radiation Control 2.0 3.8 D D D D D D D D D D ~ Ability to aprove release permits G2.4.2 Em ergency Procedures/Plans Emergency 4.5 4.6 D D D D D D D D D D ~ Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

G2.4.38 Em ergency Procedures/Plans Emergency 2.4 4.4 D D D D D D D D D D ~ Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator.

G2.4.4 Emergency Procedures/Plans 4.5 4.7 D D D D D D D D D D ~ Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

Page 1 of 1 5/6/2009 10:06 AM

ES-401, REV 9 SRO T1G1 T1 G1 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 007EG2.4.45 Reactor Trip - Stabilization - Recovery 4.1 4.3 D DDDDDDDDD~

DDDDDDDDDD~ Ability to prioritize and interpret the significance of each

/ 1 annunciator or alarm.

009EA2.11 Small Break LOCA / 3 3.8 3.8 4.1 4.1 D DDDDDD~ DDD DDDDDDD~DDD Containment temperature, pressure, and humidity 040AA2.01 Steam Line Rupture - Excessive Heat 4.2 4.7 DDDDDDD~DDD Occurrence and location of a steam line rupture from Transfer /4 pressure and flow indications 054AG2.2.40 Loss of Main Feedwater / 4 3.4 4.7 DDDDDDDDDD~ Ability to apply technical specifications for a system.

055EA2.02 Station Blackout / 6 4.4 4.6 DDDDDDD~DDD RCS core cooling through natural circulation cooling to S/G cooling 057AG2.2.4 Loss of Vital AC Inst. Bus / 6 3.6 3.6 DDDDDDDDDD~ (multi-unit)

(m ulti-unit) Ability to explain the variati ro board layouts, s . entation and Rfocedural procedural 2.2.38 2.Z.38 ~~ actio een units at a facility. ~~

S P~

5 k/..e UII-..r-UIl-..,(-

Page 1 of 1 5/6/2009 10:06 AM

ES-401, REV 9 SRO T1 T1G2 G2 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 005AG~.~.

005AG .. Inoperable/Stuck Control Rod /1

/1 3.6 3.6 DDDDDDDDDD~ (multi-unit) Ability to e In the variations in control

~

board layouts, s ms, instrumentation a.nd pr p~cedural cedural/

/

(Jl~~t:z..

UI<@5 ~.t, 2. ( actions be en units at a facility. c- dIf" l11 Lh t/f

(/~

e(

033AG2.4.30 Loss of Intermediate Range NI/7 NI / 7 2.7 4.1 DDDDDDDDDD~ Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies.

061AA2.02 ARM System Alarms /7 /7 2.9 3.2 DDDDDDD~DDD Normal radiation intensity for each ARM system channel WE10EA2.2 Natural Circ. With Seam Void/

Void/44 3.4 3.9 DDDDDDD~DDD Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Page 1 of 1 5/6/2009 10:06 AM

ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 003G2.4.41 Reactor Coolant Pump 2.9 4.6 D D D D D D D D D D ~ Knowledge of the emergency action level thresholds and classifications.

005G2.4.34 00SG2.4.34 Residual Heat Removal 4.2 4.1 D D D D D D D D D D ~ Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects 008A2.05 008A2.0S Component Cooling Water 3.3 3.5 3.S D D D D D D D ~ D D D Effect of loss of instrument and control air on the position of the CCW valves that are air operated 059A2.04 OS9A2.04 Main Feedwater 2.9 3.4 D D D D D D D ~ D D D Feeding a dry S/G 064G2.2.37 Emergency Diesel Generator 3.6 4.6 D D D D D D D D D D ~ Ability to determine operability and/or availability of safety related equipment Page 1 of 1 S/6/2009 10:06 AM 5/6/2009

ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 033A2.01 Spent Fuel Pool Cooling 3.0 "3.5

  • 3.5 0D 0D 0D 0D 0D 0D 0D ~ 0D 0D 0D Inadequate SDM 068G2.1.27 068G2.1 .27 Liquid Radwaste 3.9 4 0D 0D 0D 0D 0D 0D 0D 0D 0D 0D ~ Knowledge of system purpose and or function.

{/t:dIf it C; 2 I 1:+

071 G2.1.31 Waste Gas Disposal 4.6 4.3 0D 0D 0D 0D 0D 0D 0D 0D 0D 0D ~ Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup.

Page 1 of 1 5/6/2009 10:06 AM

ES-401, REV 9 SRO T3 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 G2.1.28 Conduct of operations 4.1 4.1 D D D D D D D D D D ~ Knowledge of the purpose and function of major system components and controls.

G2.1.44 Conduct of operations 3.9 3.8 D D D D D D D D D D ~ Knowledge of RO duties in the control room during fuel handling. .

G2.2.20 Equipment Control 2.6 3.8 D D D D D D D D D D ~ Knowledge of the process for managing troubleshooting activities.

G2.2.35 Equipment Control 3.6 4.5 D D D D D D D D D D ~ Ability to determine Technical Specification Mode of Operation G2.3.4 Radiation Control 3.2 3.7 D D D D D D D D D D ~ Knowledge of radiation exposure limits under normal and emergency conditions G2.4.16 Emergency Procedures/Plans 3.5 4.4 D D D D D D D D D D ~ Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines guidelines..

. G~ Emergency Procedures/Plans 3.9 3.8 D D D D D D D D D D ~ Knnw'eQ~e

~Knowlooge ef tlthe of RO s responsibilities itif Ii efflergeflcy Ie PIC elllergeFioy p+an plan ifflplei illl pleillIema'ttOn.

Jematton.

~ ~ z. tf*z..u G,2.ct*u)

P Page 1 of 1 5/6/2009 10:06 AM

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