Text

WT-2011-03 Final Operating Test
	ML110970454
Person / Time
Site:	Waterford
Issue date:	03/21/2011
From:	; NRC Region 4
To:	; Entergy Operations
References
	50-382/11-301
	Download: ML110970454 (391)
	v • d • e

ES-301 Administrative Topics Outline Form ES-301-1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Examination Level: RO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

2.1.23, Ability to perform specific system and A1 R, D integrated plant procedures during all modes of plant operation.

Conduct of Operations Perform a Shutdown Margin with an immoveable CEA in accordance with OP-903-090, Shutdown Margin, K/A Importance: section 7.3, Shutdown Margin Verification -

4.3 Untrippable CEA.

A2 S, M 2.1.18, Ability to make accurate, clear, and concise Conduct of Operations logs, records, status boards, and reports.

Perform OP-903-001, Technical Specification K/A Importance: Surveillance Logs, Attachment 11.18, Adjustment of 3.6 CPC and Excore Nuclear Instrumentation Data.

A3 S, N 2.2.12, Knowledge of surveillance procedures Equipment Control Complete surveillance OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring K/A Importance: Instrumentation Channel Checks.

3.7 A4 R, M 2.3.4, Knowledge of radiation exposure limits under Radiation Control normal and emergency conditions.

Calculate stay time to perform a tagout verification.

K/A Importance: Room dose rate & operators yearly dose provided.

3.2 Emergency Plan Not selected NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

Type Codes & Criteria: (C)ontrol room, (S)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)

Revision 1 RO

ES-301 Administrative Topics Outline Form ES-301-1 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Examination Level: SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*

A5 2.1.25, Ability to interpret reference materials, such as R, D graphs, curves, tables, etc.

Conduct of Operations Review and approve a completed Shutdown Margin with K/A Importance:

an immoveable CEA in accordance with OP-903-090, 3.9 Shutdown Margin, section 7.3, Shutdown Margin Verification - Un-trippable CEA.

A6 2.1.18, Ability to make accurate, clear, and concise logs, S, M records, status boards, and reports.

Conduct of Operations Review and approve OP-903-001, Technical K/A Importance:

Specification Surveillance Logs, Attachment 11.18, 3.8 Adjustment of CPC and Excore Nuclear Instrumentation Data.

A7 2.2.37, Ability to determine operability and/or availability S, M of safety related equipment.

Equipment Control Review and approve a completed Equipment Out of K/A Importance:

Service document in accordance with OP-100-010, 4.6 Equipment Out of Service.

A8 2.3.4, Knowledge of radiation exposure limits under R, M normal and emergency conditions.

Radiation Control Calculate dose and assign non-licensed operators to K/A Importance:

perform work in radiological restricted areas. Given 3.7 dose rate with and without shielding installed, time to install shielding, and job completion time using 1 operator or using 2 operators, determine proper job assignment.

A9 2.4.38, Ability to take actions called for in the facility S, M emergency plan, including supporting or acting as Emergency Plan emergency coordinator if required.

K/A Importance:

Determine appropriate classification and actions based 4.4 on a toxic gas release in accordance with EP-004-010, Toxic Chemical Contingency Procedure.

NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

Type Codes & Criteria: (C)ontrol room, (S)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)

Revision 1 SRO

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A1 Calculate Shutdown Margin with a Stuck CEA Applicant:

Examiner:

JPM A1 JOB PERFORMANCE MEASURE DATA PAGE Task: Calculate Shutdown Margin with a stuck CEA.

Task Standard: Applicant calculates Shutdown Margin in accordance with OP-903-090, Shutdown Margin. The results must conform to the answer key and conclude that reactor power is greater than the allowed power level.

References:

OP-903-090, Shutdown Margin Plant Data Book COLR Time Critical: No Validation Time: 30 mins.

K/A 2.1.23 Ability to perform specific system and Importance Rating 4.3 integrated plant procedures during all modes RO of plant operation Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 7 2011 NRC Exam

JPM A1 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-090, Shutdown Margin Plant Data Book COLR

==

Description:==

The applicant will be required to calculate Shutdown Margin with 1 mechanically bound CEA. The results will indicate that Shutdown Margin is not met and Emergency Boration is required.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 7 2011 NRC Exam

JPM A1 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Do not use Simulator data for this JPM INITIAL CONDITIONS:

Core burnup is 365 EFPD.

TCOLD is 543.5 °F.

Power is 35%.

Power is being held due to a Chemistry hold.

Regulating Group P CEAs are being withdrawn for ASI control as xenon builds in.

CEA 20 failed to withdraw or insert on the last move.

I&C has completed troubleshooting and there are no problems associated with CEDMCS.

The Shift Manager has concluded that CEA 20 is mechanically bound.

INITIATING CUES:

The CRS directs you to perform OP-903-090, Shutdown Margin, section 7.3.

Revision 1 Page 4 of 7 2011 NRC Exam

JPM A1 7.3.1 If an Untrippable CEA Condition exists and the other CEAs are not inserted, then determine Shutdown Margin and record on Attachment 10.3 as follows:

TASK ELEMENT 1 STANDARD NOTE (1) Use 541°F when using PDB Figure 1.5.7.

(2) When using graphs and tables in the Plant Data Book (PDB), to obtain Note reviewed.

the necessary data, it may be necessary and is acceptable to interpolate (approximate between data points or curves). However, extrapolation (approximation outside of the bounds of the data or curves) should not be used.

Comment:

Use Figure 1.5.7.4 due to EOC.

SAT / UNSAT TASK ELEMENT 2 STANDARD 7.3.1.1 Using current Cycle Burnup and 541 °F temperature, determine 5.65 - 5.8 Net Worth Worst Pair Stuck out (WPSO) from Figure 1.5.7.

Comment: Critical Refer to A1 Key.

SAT / UNSAT Use Figure 1.5.7.4 due to EOC.

TASK ELEMENT 3 STANDARD 7.3.1.2 Determine Shutdown Margin required by COLR. 5.15 Comment: Critical Refer to A1 Key.

SAT / UNSAT TASK ELEMENT 4 STANDARD 7.3.1.3 Subtract Step 7.3.1.2 from Step 7.3.1.1 to determine Shutdown 0.5 - 0.65 Margin Allowed Power Defect % K/K.

Comment: Critical Refer to A1 Key.

SAT / UNSAT Revision 1 Page 5 of 7 2011 NRC Exam

JPM A1 TASK ELEMENT 5 STANDARD 7.3.1.4 Record current Reactor Power on Attachment 10.3. Data recorded.

Comment:

SAT / UNSAT TASK ELEMENT 6 STANDARD 7.3.1.5 Using result from step 7.3.1.3 and Power Defect vs. Power Level, 22% - 29.5%

Figure 1.2.1, Determine Shutdown Margin Allowed Power Level.

Comment: Critical Refer to A1 Key.

SAT / UNSAT TASK ELEMENT 7 STANDARD 7.3.1.6 Verify Shutdown Margin greater than or equal to that required by Shutdown Margin is not the COLR by verifying that current power level is less than or equal to the met.

Shutdown Margin Allowed Power Level.

Comment: Critical Refer to A1 Key.

SAT / UNSAT TASK ELEMENT 8 STANDARD 7.3.2 If Shutdown Margin does not meet the requirements of Technical Communicate step to Specifications, then Commence Emergency Boration and go to OP-901-examiner.

103, Emergency Boration.

Comment:

SAT / UNSAT END OF TASK Revision 1 Page 6 of 7 2011 NRC Exam

JPM A1 SIMULATOR OPERATOR INSTRUCTIONS There is no Simulator setup for this JPM Revision 1 Page 7 of 7 2011 NRC Exam

10.3 SHUTDOWN MARGIN VERIFICATION W ORK SHEET FOR UNTRIPPABLE CEA A1 Key 7.3.1.1 Net Worth WPSO 5.65 - 5.8 K/K 7.3.1.2 Shutdown Margin required by COLR 5.15 K/K 7.3.1.3 Shutdown Margin Allowed Power Defect K/K step 7.3.1.1 ( 5.65 - 5.8 ) - step 7.3.1.2 ( 5.15 ) 0.5 - 0.65 K/K 7.3.1.4 Current Reactor Power 35 %Power 7.3.1.5 Shutdown Margin Allowed Power Level 22% - 29.5% Power 7.3.1.6 Current Power Level Shutdown Margin Allowed Power Level (Circle one) YES NO REMARKS: Emergency Boration required due to not meeting Shutdown Margin. (not required)

Performed by:

(Signature) (Date)

IV of Calculations by:

(Signature) (Date)

SM/CRS Review: /

(Signature) (Date/Time)

OP-903-090 Revision 301 Attachment 10.3 (1 of 1) 34

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 3.0 PRECAUTIONS AND LIMITATIONS 3.1 PRECAUTIONS 3.1.1 Shutdown Margin shall be greater than or equal to that specified in the Core Operating Limits Report (COLR), as required by Technical Specification 3.1.1.1 or 3.1.1.2.

3.1.2 Shutdown Margin less conservative than specified by Technical Specification 3.1.1.1 or 3.1.1.2 is a Reportable Occurrence.

3.1.3 RHOBAL Program shall be used to determine initial Xenon Reactivity Worth if a Reactor Trip occurs during non-equilibrium Xenon conditions. Post-trip transient Xenon Worth may be obtained by running a RHOBAL poison transient in accordance with section 7.6, RHOBAL Poison Transient of this procedure or by contacting Reactor Engineering.

3.1.4 For worksheets which perform projections in the RHOBAL program, a poison transient must be performed to update the Xenon and Net Samarium worths. If manual Xenon and/or Net Samarium data is input, the projection will not be performed.

3.2 LIMITATIONS 3.2.1 Information from Plant Data Book (PDB) and Reactor Engineering Book is necessary to perform this procedure unless using RHOBAL. Figure numbers contained in this procedure refer to appropriate section of the PDB. When using graphs and tables in the Plant Data Book (PDB), to obtain the necessary data, it may be necessary and is acceptable to interpolate (approximate between data points or curves). However, extrapolation (approximation outside of the bounds of the data or curves) should not be used.

3.2.2 When using graphs and tables in the PDB, the Beginning of Cycle (BOC), Peak Boron, Middle of Cycle (MOC), and End of Cycle (EOC) periods are defined as follows:

BOC = <30 EFPD Peak Boron = 30 EFPD up to 170 EFPD MOC = 170 EFPD up to 340 EFPD EOC = 340 EFPD 4

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 3.2.3 When using Xenon Worth graphs and tables in the Plant Data Book (PDB), use the figure or table associated with the Effective Full Power Days (EFPD) range listed below:

Figure 1.6.3.1: 0 up to 170 EFPD (BOC and Peak Boron)

Figure 1.6.3.2: 170 up to 340 EFPD (MOC)

Figure 1.6.3.3: 340 EFPD (EOC) 3.2.4 Column F of Attachment 11.1 of OP-004-019, Estimated Critical Configuration, can be used to satisfy Shutdown Margin in Mode 3 per either Technical Specification 4.1.1.1 (5.15% k/k when Shutdown Bank CEAs are not fully inserted) or Technical Specification 4.1.1.2 (4.6% k/k when all CEAs are fully inserted) as directed by OP-010-003, Plant Startup [CRs 98-0970, 01-0209]. This is done by verifying that the actual RCS boron concentration is no more than 20 ppm below Critical Boron Concentration of Column F, Att.11.1 and the question of Allowable CEA Range is verified to be above Transient Insertion Limit for critical operations (Group 5 60 inches) is answered yes.

3.2.5 In the RHOBAL program when calculating Critical Boron Concentration, the adjusted boron concentration column of step 4.2 on Worksheet 1 for Critical Boron Concentration of OP-004-019, Estimated Critical Configuration, can be used to satisfy Shutdown Margin in Mode 3 per either Technical Specification 4.1.1.1 (5.15% k/k when Shutdown Bank CEAs are not fully inserted) or Technical Specification 4.1.1.2 (4.6% k/k when all CEAs are fully inserted) as directed by OP-010-003, Plant Startup [CRs 98-0970, 01-0209]. This is done by verifying that the Actual RCS Boron Concentration is no more than 20 ppm below the Estimated Critical Boron in step 4.2 of Worksheet 1 for Critical Boron Concentration, and the

-0.5% k/k rod position in step 5.3 of Worksheet 1 for Critical Rod Position is above the Transient Insertion Limit for critical operations (Group 5 60 inches).

3.2.6 In the RHOBAL program, if an input parameter is beyond the range of the cycle specific input database, a warning message is printed on the screen and/or the error log. This is intended to prevent performing a calculation outside the analyzed window.

3.2.7 In the RHOBAL program, screen minimization is not allowed. When the calculations are completed on a particular screen, it must be closed for control to return to a previous screen.

3.2.8 Changes to this procedure shall be reviewed by the Reactor Engineering (RE)

Department prior to approval. [P-21855]

5

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 6.0 ACCEPTANCE CRITERIA 6.1 Shutdown Margin is that specified in the COLR by either:

[T.S. 3.1.1.2, T.S. 4.1.1.1.1.a, T.S. 4.1.1.1.1.e]

6.1.1 Current Boron Concentration is Shutdown Margin Boron Concentration.

or 6.1.2 If Reactor is critical with no Untrippable CEAs, and all CEAs are above Transient Insertion Limit. (Operation outside the Transient Insertion Limit is allowed up to two hours per Technical Specification 3.1.3.6).

or 6.1.3 For Dropped or Untrippable CEA, Current Power Level is Shutdown Margin Allowed Power Level.

6.1.4 Current Shutdown Margin is required Shutdown Margin.

8

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 7.3 SHUTDOWN MARGIN VERIFICATION - UNTRIPPABLE CEA 7.3.1 If an Untrippable CEA Condition exists and the other CEAs are not inserted, then determine Shutdown Margin and record on Attachment 10.3 as follows:

NOTE (1) Use 541 F when using PDB Figure 1.5.7.

(2) When using graphs and tables in the Plant Data Book (PDB), to obtain the necessary data, it may be necessary and is acceptable to interpolate (approximate between data points or curves). However, extrapolation (approximation outside of the bounds of the data or curves) should not be used.

7.3.1.1 Using current Cycle Burnup and 541oF temperature, determine Net Worth Worst Pair Stuck out (WPSO) from Figure 1.5.7.

7.3.1.2 Determine Shutdown Margin required by COLR.

7.3.1.3 Subtract Step 7.3.1.2 from Step 7.3.1.1 to determine Shutdown Margin Allowed Power Defect % K/K.

7.3.1.4 Record current Reactor Power on Attachment 10.3.

7.3.1.5 Using result from step 7.3.1.3 and Power Defect vs. Power Level, Figure 1.2.1, Determine Shutdown Margin Allowed Power Level.

7.3.1.6 Verify Shutdown Margin greater than or equal to that required by the COLR by verifying that current power level is less than or equal to the Shutdown Margin Allowed Power Level.

7.3.2 If Shutdown Margin does not meet the requirements of Technical Specifications, then Commence Emergency Boration and go to OP-901-103, Emergency Boration.

NOTE Subsection 7.4, Shutdown Margin Verification - Untrippable CEA, Other CEAs Inserted, is applicable when all other CEAs are inserted.

7.3.3 If the Reactor has been shutdown less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then determine Required Shutdown Margin Boron Concentration required to meet Shutdown Margin for the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by performing Subsection 7.4, Shutdown Margin Verification -

Untrippable CEA, Other CEAs Inserted.

15

10.3 SHUTDOWN MARGIN VERIFICATION W ORK SHEET FOR UNTRIPPABLE CEA 7.3.1.1 Net Worth WPSO K/K 7.3.1.2 Shutdown Margin required by COLR K/K 7.3.1.3 Shutdown Margin Allowed Power Defect K/K step 7.3.1.1 ( ) - step 7.3.1.2 ( ) K/K 7.3.1.4 Current Reactor Power %Power 7.3.1.5 Shutdown Margin Allowed Power Level % Power 7.3.1.6 Current Power Level Shutdown Margin Allowed Power Level (Circle one) YES NO REMARKS:

Performed by:

(Signature) (Date)

IV of Calculations by:

(Signature) (Date)

SM/CRS Review: /

(Signature) (Date/Time)

OP-903-090 Revision 301 Attachment 10.3 (1 of 1) 34

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A2 Perform OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Applicant:

Examiner:

JPM A2 JOB PERFORMANCE MEASURE DATA PAGE Task: Perform OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Task Standard: Applicant correctly calculates new values for Core Protection Calculator constants KCAL, TCREF, and TPC.

References:

OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Time Critical: No Validation Time: 20 mins.

K/A 2.1.18 Ability to make accurate, clear, and Importance Rating 3.6 concise logs, records, status boards, and RO reports.

Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 6 2011 NRC Exam

JPM A2 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Simulator

==

Description:==

Applicant will use the simulator PMC and Core Protection Calculator B to gather data and calculate CPC constants KCAL, TCREF, and TPC. The procedure has provisions to collect 5 sets of data and average them over a 5 minute period, but it is optional at the supervisors request. The applicants will not have to collect 5 sets of data. The applicants will not have to change the constants, just perform the calculations.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 6 2011 NRC Exam

JPM A2 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Plant conditions are as you see on the Simulator. Use of the Simulator, RM-11, and Plant Monitoring Computer is acceptable. You are not allowed to use any network computers.

Do not go to a location if another applicant is there.

INITIAL CONDITIONS:

The UFM is in service.

The plant is in steady state operation as displayed on the control panels.

INITIATING CUES:

The CRS directs you to complete the following calculations for Channel B on 1.18:

step 11.10.6 for KCAL and TCREF step 11.10.7 for TPC Linear potentiometer data collection is not required for this task.

Data collection at CP-10 is not required for this task.

The CRS directs you to gather required data once for column 0, averaging columns 0 through 4 is not required.

This task is complete when you reach step 11.10.8.

Revision 1 Page 4 of 6 2011 NRC Exam

JPM A2 TASK ELEMENT 1 STANDARD Student must record values and make Complete OP-903-001, Attachment 11.18 according to key.

calculations for 3 points.

Comment: Critical Data collection and calculations are displayed on the key.

SAT / UNSAT END OF TASK Revision 1 Page 5 of 6 2011 NRC Exam

JPM A2 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-190 After all simulator generated alarms are clear, place the simulator in Run.

Verify the parameters on the PMC match the values on the key.

Revision 1 Page 6 of 6 2011 NRC Exam

A2 Key 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA DATE _______________ CHANNEL UNDER ADJUSTMENT: A B C D 11.10.3.1 Calculate and record the averages of each parameter in the space provided.

0 1 2 3 4 Average Adjusted BSCAL 99.87 PMC PID-C24230 HI LINEAR POWER N/A BISTABLE 1 VOLTS HIGH LINEAR POWER %

N/A VOLT X 20 PHICAL.

(Calibrated Neutron Flux Power) 99.96 CPC PID 171 BDT (Static Thermal Power) 99.82 CPC PID 177 Calculations Performed by: Verified by:

Signature Signature Refer to attachment 11.1 Note 9.1 to determine appropriate power indication if linear power is not 35% steady state. Document indication used in Remarks.

If COLSS is Inoperable, then use NE-005-201, Heat Balance Calculations, to determine Secondary Calorimetric Power substitute when PMC or CORE POWER is specified.

Adjusted is the average value plus 8.5% (8% to10%) if adjustments are being made to PHICAL and/or BDT as listed in Notes above steps 11.10.6 or 11.10.7 (refer to Attachment 11.1 Note 9.8). Otherwise N/A this block. Use the Average value, not the Adjusted value, for DVM calculation.

11.10.4 Record the following for channel under adjustment:

TPC (Thermal Power Calibration Constant) 0.84399 CPC PID 064 ...................................................

KCAL (Neutron Flux Power Cal. Constant) 1.0070 CPC PID 065 ...................................................

PCALIB (Secondary Calorimetric Power Used in 100 Latest CPC Power Calibration) CPC PID 104 TC 1 (Loop 1 Cold Leg Temperature) 543.59 CPC PID 160 ...................................................

TC 2 (Loop 2 Cold Leg Temperature) 543.46 CPC PID 161 ...................................................

TCORF (Temp Shadowing Correction Factor) .99902 CPC PID 180 ...................................................

EXCORE LINEAR POWER CALIBRATE N/A POTENTIOMETER POSITION ROM ..............

OP-903-001 Revision 042 Attachment 11.18 (1 of 3) 142

A2 Key 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.5 Calculate the new DVM reading or new potentiometer position as follows:

Avg. Core Power (Step 11.10.3.1)* N/A DVM (new) = =

20 20

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

DVM (new) = N/A or Potentiometer position (new) =

Avg. Core Power % (Step 11.10.3.1)* X Old Potentiometer Setting (Step 11.10.4)

Avg. Linear Power % (Step 11.10.4)

N/A X N/A N/A

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

Potentiometer position (new) = ______N/A_______

Performed by: N/A Verified by: N/A (Initials) (Initials) 11.10.6.1 Calculate KCAL (new):

(Step 11.10.3.1)* (step 11.10.4) (step 11.10.4)

KCAL Avg. Core Power (%) x KCAL x TCORF

=

(new) Avg. PHICAL (step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.6.

KCAL 99.87 x 1.0070 x .99902

=

(new) 99.96 KCAL 1.0051

=

(new)

KCAL (new) = 1.0051 (CPC PID 065) 11.10.6.3 New TCREF (CPC PID 098) = Minimum TC from step 11.10.4:

TC 1 (CPC PID 160) or TC 2 (CPC PID 161).

TCREF (new) = 543.46 (CPC PID 098)

Performed by: Verified by:

(Initials) (Initials)

OP-903-001 Revision 042 Attachment 11.18 (2 of 3) 143

A2 Key 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.7 Calculate TPC (CPC PID 064):

Avg. Core Power % (Step 11.10.3.1)* X TPC (Step 11.10.4)

TPC (new)=

Avg. BDT (Step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.7.

99.87 X 0.84399 TPC (new)=

99.82 TPC (new) = 0.84441 (CPC PID 064).

Performed by: Verified by:

(Initials) (Initials) 11.10.11.1 Record the following:

Applicable CORE POWER PMC....................................___________

PCALIB CPC PID 104 ..............................................___________

HI LINEAR POWER BISTABLE 1 VOLTS .....................___________

HI LINEAR POWER % VOLTS x 20 ..............................___________

PHICAL CPC PID 171 ..............................................___________

BDT CPC PID 177 ..............................................___________

11.10.11.2.1 Record answers:

HI LINEAR POWER %, VOLTS X 20.................... YES/NO ________

PHICAL, CPC PID 171 ..................................... YES/NO ________

BDT, CPC PID 177 ..................................... YES/NO ________

11.10.12 Performed by: Verified by:

(Initials) (Initials) 11.10.13 Reviewed by: _________________________ _________________

SM/CRS Date/Time OP-903-001 Revision 042 Attachment 11.18 (3 of 3) 144

11.10 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION INSTRUCTION DATE ______________ CHANNEL UNDER ADJUSTMENT: A B C D NOTE During physics testing, Note 2 of TS Table 4.3-1 applies and allows the suspension of this adjustment until the next power plateau is reached. The assigned Shift Test Director should be notified during testing prior to making any adjustments per this attachment.

11.10.1 If a non-conservative adjustment (indicated power is lowered) is required to be made on CPC and Excore Nuclear Instrumentation, then verify that both of the following conditions are met:

BDELT (C24104) is within 2% of BTFSP (C24101)

The average of BDELT (C24104) and BTFSP (C24101) is within 2% of Calorimetric Power 11.10.1.1 If either of these conditions cannot be met, then notify Reactor Engineering (RE) for further evaluation prior to making any adjustment.

11.10.2 If either of the following conditions are not met, then use the applicable power for CORE POWER PMC as specified in Note 9.1 of Attachment 11.1.

UFM is in service Plant is in steady state operation NOTE Averages are recommended for accuracy, but are not required.

11.10.3 Record the following data at approximately one minute intervals on Attachment 11.18:

BSCAL- PMC PID C24230 HI LINEAR POWERBISTABLE 1 VOLTS HIGH LINEAR POWER % VOLT X 20 PHICAL (Calibrated Neutron Flux Power) CPC PID 171 BDT (Static Thermal Power) CPC PID 177 11.10.3.1 Calculate and record averages of each parameter on Attachment 11.18.

OP-903-001 Revision 042 Attachment 11.10 (1 of 9) 110

11.10.4 Record the following data for channel under adjustment on Attachment 11.18:

TPC (Thermal Power Calibration Constant) CPC PID 064 KCAL (Neutron Flux Power Cal. Constant) CPC PID 065 PCALIB (Secondary Calorimetric Power Used in Latest CPC Power Calibration) CPC PID 104 TC 1 (Loop 1 Cold Leg Temperature) CPC PID 160 TC 2 (Loop 2 Cold Leg Temperature) CPC PID 161 TCORF (Temp Shadowing Correction Factor) CPC PID 180 EXCORE LINEAR POWER CALIBRATE POTENTIOMETER POSITION ROM (may be N/Ad if not using in step 11.10.5) 11.10.5 If Hi Linear Power requires adjustment, then calculate the new DVM reading or new potentiometer position on Attachment 11.18.

OP-903-001 Revision 042 Attachment 11.10 (2 of 9) 111

NOTE Under the following conditions, PHICAL (CPC PID 171) must be adjusted to between +8.0%

and +10.0% above the calorimetric power indication (refer to Note 9.8 on Attachment 11.1).

This is performed by adding 8.5% (8% to10%) to the average core power value on 1.18 to obtain an Adjusted value that will be used as the Avg. Core Power value in the KCAL (new) calculation (this requirement does not apply during initial power ascension to <80% RTP following refueling): [CR-WF3-2006-03726]

Calorimetric power is between 15% RTP and 80% RTP.

and PHICAL is greater than 10.0% above Calorimetric power.

11.10.6 If KCAL (PHICAL) requires adjustment, then perform the following.

11.10.6.1 Calculate KCAL (new) on Attachment 11.18.

CAUTION CPC PID 065 (KCAL) limit is 0.7 to 2.0. [CR-WF3-1998-00919]

11.10.6.2 If KCAL (new) is 0.75, then notify RE and PMI that Linear Channel Sub-Gain needs to be adjusted. [CR-WF3-1998-00919]

11.10.6.3 Document new TCREF(CPC PID 098) on Attachment 11.18.

NOTE Under the following conditions, BDT (CPC PID 177) must be adjusted to between +8.0% and

+10.0% above the calorimetric power indication (refer to Note 9.8 on Attachment 11.1). This is performed by adding 8.5% (8% to10%) to the average core power value on Attachment 11.18 to obtain an Adjusted value that will be used as the Avg. Core Power value in the TPC calculation (this requirement does not apply during initial power ascension to <80% RTP following refueling): [CR-WF3-2006-03726]

Calorimetric power is between 15% RTP and 80% RTP.

and BDT is greater than 10.0% above Calorimetric power.

11.10.7 If BDT requires adjustment, then Calculate TPC (CPC PID 064) on Attachment 11.18.

OP-903-001 Revision 042 Attachment 11.10 (3 of 9) 112

NOTE (1) Only one PPS Channel may be placed in Bypass at a time. If High Linear Power/

DNBR/LPD trips are already bypassed in one PPS Channel or if the reactor is shutdown, then PPS bypass installation and removal are to be performed at SM/CRS discretion while maintaining Tech Spec requirements.

(2) PCALIB provides Core Protection Calculators the last calibrated power value at which PHICAL and/or BDT was adjusted to meet T.S. 4.3.1.1 requirements. PCALIB should be verified and adjusted as necessary anytime BDT and/or PHICAL is adjusted or upon Reactor Engineering request. If current CPC PCALIB value meets requirements, then no adjustment is necessary. The required value of CPC PID 104 (PCALIB) is determined as follows:

If 82% Calorimetric power, then verify PCALIB is set to 100.

If <82% Calorimetric power, then verify PCALIB is set to the calorimetric power value.

11.10.8 If KCAL (PHICAL), TPC (BDT), or PCALIB requires adjustment, then perform the following and enter values on Attachment 11.19:

11.10.8.1 Refer to Tech Spec 3.3.1, for any applicable LCOs.

11.10.8.2 Enter reason for change to Addressable Constant.

11.10.8.3 Check applicable Channel to be adjusted (Channel A, B, C, or D).

11.10.8.4 Record PID number to be changed.

CAUTION CPC PID 065 (KCAL) limit is 0.7 to 2.0. [CR-WF3-1998-00919]

11.10.8.5 Record new value to be entered.

11.10.8.5.1 If CPC PID 065 (KCAL) is being adjusted and the value is 0.75, then notify RE and PMI that Linear Channel Sub-Gain needs to be adjusted. [CR-WF3-1998-00919]

11.10.8.6 Obtain SM/CRS authorization.

11.10.8.7 Repeat steps 11.10.8.2 through step 11.10.8.6 for all addressable constants that will be changed for this adjustment.

OP-903-001 Revision 042 Attachment 11.10 (4 of 9) 113

11.10.9 If Excore Nuclear Instrumentation requires adjustment, then perform the following:

11.10.9.1 Refer to Tech Spec 3.3.1, for any applicable LCOs.

CAUTION ONLY ONE PPS CHANNEL MAY BE PLACED IN BYPASS AT A TIME. IF LINEAR POWER OR DNBR/LPD TRIPS ARE ALREADY BYPASSED IN ONE PPS CHANNEL OR IF THE REACTOR IS SHUTDOWN, THEN PPS BYPASS INSTALLATION AND REMOVAL ARE TO BE PERFORMED AT SM/CRS DISCRETION WHILE MAINTAINING TECH SPEC REQUIREMENTS AND THE APPLICABLE BLOCKS MAY BE N/A.

11.10.9.2 Place the following bistables in BYPASS for the channel to be adjusted:

LINEAR POWER DNBR LPD 11.10.9.3 Establish communications between an operator at CP-10 and an operator at CP-7.

11.10.9.4 For the channel in BYPASS, position the BISTABLE SELECT SWITCH, on CP-10 to bistable 1 position.

11.10.9.5 For the channel in BYPASS, verify the following switch positions:

Excore Detector METER SELECT switch is in the CAL SUM position METER INPUT SELECT switch is in the INPUT position 11.10.9.6 The operator at CP-7 shall coordinate with the operator at CP-10 and adjust the CORE LINEAR POWER CALIBRATE potentiometer (located on the PPS ROM) in accordance with one of the following:

The DVM reads the DVM (new) value calculated in step 11.10.5.

EXCORE LINEAR POWER CALIBRATE potentiometer to the Potentiometer position (new) value calculated in step 11.10.5.

11.10.9.7 Verify the DVM reading on CP-10 for the channel in BYPASS is consistent with the value obtained in step 11.10.5 and current Reactor Power.

OP-903-001 Revision 042 Attachment 11.10 (5 of 9) 114

11.10.10 If KCAL (PHICAL), TPC (BDT), or PCALIB adjustments are required, then perform the following:

11.10.10.1 Refer to Tech Spec 3.3.1, for any applicable LCOs.

CAUTION ONLY ONE PPS CHANNEL MAY BE PLACED IN BYPASS AT A TIME. IF DNBR/LPD TRIPS ARE ALREADY BYPASSED IN ONE PPS CHANNEL OR IF THE REACTOR IS SHUTDOWN, THEN PPS BYPASS INSTALLATION AND REMOVAL ARE TO BE PERFORMED AT SM/CRS DISCRETION WHILE MAINTAINING TECH SPEC REQUIREMENTS AND THE APPLICABLE BLOCKS MAY BE N/A.

11.10.10.2 Verify the following bistables in BYPASS for the channel to be adjusted:

DNBR LPD 11.10.10.3 Place Calculator Select switch for desired channel to CPC.

11.10.10.4 Have another Licensed Operator or STA observe the performance of steps 11.10.10.6 through 11.10.20 to verify the correct data is entered.

11.10.10.5 Place Function keyswitch to ON.

11.10.10.6 Display current value of PID listed in step 11.10.8.4.

11.10.10.6.1 Record value on Attachment 11.19.

11.10.10.7 Display current value of X1, Flow Adjusted DNBR (PID 406).

11.10.10.7.1 Record value on Attachment 11.19.

11.10.10.8 Display current value of LPDDC, Compensated Local Power Density (PID 179).

11.10.10.8.1 Record value on Attachment 11.19.

11.10.10.9 Depress Change Value pushbutton.

11.10.10.10 Enter three-digit number for Addressable Constant to be changed from step 11.10.8.4.

11.10.10.11 Depress Enter pushbutton.

OP-903-001 Revision 042 Attachment 11.10 (6 of 9) 115

NOTE Addressable Constant value must be five digits. Values must be within limits specified in OP-004-006, Core Protection Calculator System, Attachment 11.3, CPC Addressable Constants.

11.10.10.12 Enter new value for Addressable Constant.

11.10.10.13 Depress Execute pushbutton.

11.10.10.14 Verify value entered into Addressable Constant is correct by displaying PID listed in step 11.10.8.4.

11.10.10.14.1 Record value on Attachment 11.19.

11.10.10.15 Record current value of X1, Flow Adjusted DNBR (PID 406) on Attachment 11.19.

11.10.10.16 Record current value of LPDDC Compensated Local Power Density (PID 179) on Attachment 11.19.

11.10.10.17 Place Function keyswitch to OFF.

11.10.10.18 Verify new values for DNBR and LPD are consistent with new constant(s).

11.10.10.19 Document change performance on Attachment 11.19.

11.10.10.20 Obtain independent verification of As-Left value for changed PID on Attachment 11.19.

CAUTION DNBR/ LPD PRETRIP(S) AND TRIP(S) SHALL BE RESET PRIOR TO REMOVING ASSOCIATED PPS CHANNEL BISTABLES FROM BYPASS.

11.10.10.21 Repeat steps 11.10.10.3 through step 11.10.10.20 until all desired addressable constant changes have been made.

11.10.10.22 Request SM/CRS to document review of Addressable Constant Change on Attachment 11.19.

11.10.10.23 When all intended Addressable Constant changes are complete, then perform a channel check for LPD and DNBR in accordance with OP-903-001, Technical Specification Surveillance Logs.

OP-903-001 Revision 042 Attachment 11.10 (7 of 9) 116

11.10.10.24 When all intended Addressable Constant changes are complete and in Mode 1 or 2, then perform a validity check for LPD and DNBR margins in accordance with OI-004-000, Operations Narrative and Shift Logs.

11.10.10.25 Request SM/CRS to complete OP-004-006 Attachment 11.8, Addressable Constant Change Log and document verification of correct data transfer.

11.10.10.26 Place current change request in CPC Addressable Constant Book.

11.10.10.26.1 Transmit old change request.

11.10.11 Verification 11.10.11.1 Record the following on Attachment 11.18:

Applicable CORE POWER PMC PCALIB CPC PID 104 HI LINEAR POWER BISTABLE 1 VOLTS HI LINEAR POWER % VOLTS x 20 PHICAL CPC PID 171 BDT CPC PID 177 OP-903-001 Revision 042 Attachment 11.10 (8 of 9) 117

NOTE Under the following conditions, the values of BDT (CPC PID 177) and/or PHICAL (CPC PID 171) after adjustment must be between +8.0% and +10.0% above the calorimetric power indication (refer to Note 9.8 on Attachment 11.1). This requirement does not apply during initial power ascension to <80% RTP following refueling: [CR-WF3-2006-03726]

Calorimetric power is between 15% RTP and 80% RTP.

and PHICAL is greater than 10.0% above Calorimetric power.

11.10.11.2 Verify the following parameters are within the limits of Tech Spec 4.3.1.1 Table 4.3-1(2) for CORE POWER:

HI LINEAR POWER %, VOLTS X 20 PHICAL, CPC PID 171 BDT, CPC PID 177 11.10.11.2.1 Document answers on Attachment 11.18.

11.10.11.2.2 If all of the above are YES, then this channel's adjustments are correct.

11.10.11.2.3 If any of the above are NO, then re-perform the necessary parts of this attachment.

11.10.12 If selected channel is to be returned to service, then remove/verify removed the following bistables from BYPASS:

Linear Power LPD Bistable DNBR Bistable.

11.10.13 Submit Attachment 11.18 to SM/CRS for review.

OP-903-001 Revision 042 Attachment 11.10 (9 of 9) 118

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA DATE _______________ CHANNEL UNDER ADJUSTMENT: A B C D 11.10.3.1 Calculate and record the averages of each parameter in the space provided.

0 1 2 3 4 Average Adjusted BSCAL N/A N/A N/A N/A PMC PID-C24230 HI LINEAR POWER N/A N/A N/A N/A N/A N/A BISTABLE 1 VOLTS HIGH LINEAR POWER %

N/A N/A N/A N/A N/A N/A VOLT X 20 PHICAL.

(Calibrated Neutron Flux Power) N/A N/A N/A N/A CPC PID 171 BDT (Static Thermal Power) N/A N/A N/A N/A CPC PID 177 Calculations Performed by: Verified by:

Signature Signature Refer to attachment 11.1 Note 9.1 to determine appropriate power indication if linear power is not 35% steady state. Document indication used in Remarks.

If COLSS is Inoperable, then use NE-005-201, Heat Balance Calculations, to determine Secondary Calorimetric Power substitute when PMC or CORE POWER is specified.

Adjusted is the average value plus 8.5% (8% to10%) if adjustments are being made to PHICAL and/or BDT as listed in Notes above steps 11.10.6 or 11.10.7 (refer to Attachment 11.1 Note 9.8). Otherwise N/A this block. Use the Average value, not the Adjusted value, for DVM calculation.

11.10.4 Record the following for channel under adjustment:

TPC (Thermal Power Calibration Constant)

CPC PID 064 ...................................................

KCAL (Neutron Flux Power Cal. Constant)

CPC PID 065 ...................................................

PCALIB (Secondary Calorimetric Power Used in Latest CPC Power Calibration) CPC PID 104 TC 1 (Loop 1 Cold Leg Temperature)

CPC PID 160 ...................................................

TC 2 (Loop 2 Cold Leg Temperature)

CPC PID 161 ...................................................

TCORF (Temp Shadowing Correction Factor)

CPC PID 180 ...................................................

EXCORE LINEAR POWER CALIBRATE N/A POTENTIOMETER POSITION ROM ..............

OP-903-001 Revision 042 Attachment 11.18 (1 of 3) 142

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.5 Calculate the new DVM reading or new potentiometer position as follows:

Avg. Core Power (Step 11.10.3.1)* N/A DVM (new) = =

20 20

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

DVM (new) = N/A or Potentiometer position (new) =

Avg. Core Power % (Step 11.10.3.1)* X Old Potentiometer Setting (Step 11.10.4)

Avg. Linear Power % (Step 11.10.4)

N/A X N/A N/A

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

Potentiometer position (new) = _______ N/A _________

Performed by: N/A Verified by: N/A (Initials) (Initials) 11.10.6.1 Calculate KCAL (new):

(Step 11.10.3.1)* (step 11.10.4) (step 11.10.4)

KCAL Avg. Core Power (%) x KCAL x TCORF

=

(new) Avg. PHICAL (step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.6.

KCAL x x

=

(new)

KCAL

=

(new)

KCAL (new) = (CPC PID 065) 11.10.6.3 New TCREF (CPC PID 098) = Minimum TC from step 11.10.4:

TC 1 (CPC PID 160) or TC 2 (CPC PID 161).

TCREF (new) = (CPC PID 098)

Performed by: Verified by:

(Initials) (Initials)

OP-903-001 Revision 042 Attachment 11.18 (2 of 3) 143

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.7 Calculate TPC (CPC PID 064):

Avg. Core Power % (Step 11.10.3.1)* X TPC (Step 11.10.4)

TPC (new)=

Avg. BDT (Step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.7.

X TPC (new)=

TPC (new) = (CPC PID 064).

Performed by: Verified by:

(Initials) (Initials) 11.10.11.1 Record the following:

Applicable CORE POWER PMC....................................___________

PCALIB CPC PID 104 ..............................................___________

HI LINEAR POWER BISTABLE 1 VOLTS .....................___________

HI LINEAR POWER % VOLTS x 20 ..............................___________

PHICAL CPC PID 171 ..............................................___________

BDT CPC PID 177 ..............................................___________

11.10.11.2.1 Record answers:

HI LINEAR POWER %, VOLTS X 20.................... YES/NO ________

PHICAL, CPC PID 171 ..................................... YES/NO ________

BDT, CPC PID 177 ..................................... YES/NO ________

11.10.12 Performed by: Verified by:

(Initials) (Initials) 11.10.13 Reviewed by: _________________________ _________________

SM/CRS Date/Time OP-903-001 Revision 042 Attachment 11.18 (3 of 3) 144

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A3 Complete OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks Applicant:

Examiner:

JPM A3 JOB PERFORMANCE MEASURE DATA PAGE Task: Complete surveillance OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks.

Task Standard: Applicant correctly completes OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks in accordance with key and determines channel operability as listed under Task Element 1 and the key.

References:

OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks Time Critical: No Validation Time: 15 mins.

K/A 2.2.12, Knowledge of surveillance Importance Rating 3.7 procedures RO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 6 2011 NRC Exam

JPM A3 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks

==

Description:==

Applicant will be directed to perform OP-903-013, Monthly Channel Checks, 0.3 for Accident Monitoring Instrumentation Channel Checks, specific to the section for CP-8. There are reading failed for a variety of parameters as indicated in the key. These reactor operator applicants will not be required to make the associated Tech Spec calls.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 6 2011 NRC Exam

JPM A3 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Plant conditions are as you see on the Simulator. Use of the Simulator, RM-11, and Plant Monitoring Computer is acceptable. You are not allowed to use any network computers.

INITIAL CONDITIONS:

QSPDS -1 is out of service.

Plant conditions are as displayed.

INITIATING CUES:

The CRS directs you to perform OP-903-013, Monthly Channel Checks, Attachment 10.3 for Accident Monitoring Instrumentation Channel Checks.

You are only required to perform the section related to the instruments on CP-8 and their related QSPDS points.

Another operator will perform the surveillance on the other panels.

Perform this review and mark any comments on this sheet.

Revision 1 Page 4 of 6 2011 NRC Exam

JPM A3 TASK ELEMENT 1 STANDARD OP-903-013, Monthly Channel Checks, Attachment 10.3 for Applicant must record values for the CP-Accident Monitoring Instrumentation Channel Checks, for 8 instrumentation as listed on the A3 Key.

the instruments on CP-8 ONLY.

Comment: Critical Issues identified on the A3 Key:

SAT / UNSAT Containment Wide Range pressure from ESF-IPR-6750 B reads 17.2 PSIA, which fails the channel check limit.

QSPDS is not allowed as a substitute for this parameter.

Minimum channels are not operable.

Hot Leg temperature from RC-ITI-0122HA reads 575 °F, which exceeds the channel check limit. QSPDS - 2 is allowed as a substitution instrument. QSPDS - 1 is out of service. Minimum channels are operable.

Containment Wide Wide Range pressure from ESF-IPR-6755 A reads 15 PSIG, which exceeds the channel check limit. QSPDS - 2 is allowed as a substitute. Minimum channels are operable.

END OF TASK Revision 1 Page 5 of 6 2011 NRC Exam

JPM A3 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-190 After all simulator generated alarms are clear, verify the simulator in RUN.

Verify the parameters on the PMC match the values on the key.

Revision 1 Page 6 of 6 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A4 Calculate Stay Times Based on Dose Rates Applicant:

Examiner:

JPM A4 JOB PERFORMANCE MEASURE DATA PAGE Task: Calculate Stay Times Based on Dose Rates Task Standard: Applicant correctly calculated the allowed stay time to complete the described tagout without exceeding his yearly Waterford 3 administrative radiation dose limits.

References:

None Time Critical: No Validation Time: 10 mins.

K/A 2.3.4, Knowledge of radiation exposure limits Importance Rating 3.2 under normal and emergency conditions. RO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 7 2011 NRC Exam

JPM A4 Revision 0 Page 3 of 7 2011 NRC Exam

JPM A4 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

None READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Instructions on applicants cue sheet You have been assigned to verify a 5 valve tagout on the -11 Elevation in Containment.

Your yearly dose to date is 1875 mrem TEDE for the year.

The dose rate in that area is 850 mrem/hour.

Based on Waterford 3 yearly administrative limits, what is your stay time in the room?

Do all of your calculations on this sheet.

Revision 0 Page 4 of 7 2011 NRC Exam

JPM A4 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

You have been assigned to verify a 5 valve tagout on the -11 Elevation in Containment.

Your yearly dose to date is 1875 mrem TEDE for the year.

The dose rate in that area is 850 mrem/hour.

Based on Waterford 3 yearly administrative limits, what is your stay time in the room?

Do all of your calculations on this sheet.

Revision 0 Page 5 of 7 2011 NRC Exam

JPM A4 TASK ELEMENT STANDARD Calculate stay time based on dose rate and Waterford 3 yearly Applicant calculated the stay time as TEDE limits. 8.8 minutes.

Comment: Critical SAT / UNSAT Waterford 3 administrative TEDE limit: 2000 mrem Dose for the year: 1875 mrem Remaining dose for the year: 125 mrem Time allowed in room: 125 mrem / 850 mrem/hour 0.147 hour0.0017 days 0.0408 hours 2.430556e-4 weeks 5.59335e-5 months or 8.8 minutes END OF TASK Revision 0 Page 6 of 7 2011 NRC Exam

JPM A4 SIMULATOR OPERATOR INSTRUCTIONS There are no Simulator setup requirements for this JPM.

Revision 0 Page 7 of 7 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A5 Review and Approve a Shutdown Margin with a Stuck CEA Applicant:

Examiner:

JPM A5 JOB PERFORMANCE MEASURE DATA PAGE Task: Review and Approve a Shutdown Margin with a stuck CEA.

Task Standard: Applicant reviews the completed Shutdown Margin in accordance with OP-903-090, Shutdown Margin. The applicant must identify that the incorrect Shutdown Margin was used and correct that error.

The applicant must also identify that the plant is operating above the allowed power level and Emergency Boration is required.

References:

OP-903-090, Shutdown Margin Plant Data Book COLR Time Critical: No Validation Time: 30 mins.

K/A 2.1.23 Ability to perform specific system and Importance Rating 4.4 integrated plant procedures during all modes SRO of plant operation Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 7 2011 NRC Exam

JPM A5 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-090, Shutdown Margin Plant Data Book COLR

==

Description:==

The applicant will be required to review a completed Shutdown Margin with 1 mechanically bound CEA. The review will indicate that Shutdown Margin is not met and Emergency Boration is required.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 7 2011 NRC Exam

JPM A5 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Do not use Simulator data for this JPM INITIAL CONDITIONS:

Core burnup is 365 EFPD.

TCOLD is 543.5 °F.

Power is 35%.

Power is being held due to a Chemistry hold.

Regulating Group P CEAs are being withdrawn for ASI control as xenon builds in.

CEA 20 failed to withdraw or insert on the last move.

I&C has completed troubleshooting and there are no problems associated with CEDMCS.

The Shift Manager has concluded that CEA 20 is mechanically bound.

INITIATING CUES:

The STA has provided you with a completed OP-903-090, Shutdown Margin, for review.

Review and approve the completed surveillance.

Revision 1 Page 4 of 7 2011 NRC Exam

JPM A5 7.3.1 If an Untrippable CEA Condition exists and the other CEAs are not inserted, then determine Shutdown Margin and record on Attachment 10.3 as follows:

TASK ELEMENT 1 STANDARD NOTE (1) Use 541°F when using PDB Figure 1.5.7.

(2) When using graphs and tables in the Plant Data Book (PDB), to obtain Note reviewed.

the necessary data, it may be necessary and is acceptable to interpolate (approximate between data points or curves). However, extrapolation (approximation outside of the bounds of the data or curves) should not be used.

Comment:

Use Figure 1.5.7.4 due to EOC.

SAT / UNSAT TASK ELEMENT 2 STANDARD 7.3.1.1 Using current Cycle Burnup and 541 °F temperature, determine Recorded correctly on Net Worth Worst Pair Stuck out (WPSO) from Figure 1.5.7. surveillance.

Comment:

Use Figure 1.5.7.4 due to EOC.

SAT / UNSAT TASK ELEMENT 3 STANDARD 7.3.1.2 Determine Shutdown Margin required by COLR. 5.15 Comment: Critical Surveillance lists 4.6, the COLR Shutdown Margin for > 500 F with CEAs inserted. SAT / UNSAT TASK ELEMENT 4 STANDARD 7.3.1.3 Subtract Step 7.3.1.2 from Step 7.3.1.1 to determine Shutdown 0.5 - 0.65 Margin Allowed Power Defect % K/K.

Comment: Critical Surveillance lists 1.147 because of previous error.

SAT / UNSAT Revision 1 Page 5 of 7 2011 NRC Exam

JPM A5 TASK ELEMENT 5 STANDARD 7.3.1.5 Using result from step 7.3.1.3 and Power Defect vs. Power Level, 22% - 29.5%

Figure 1.2.1, Determine Shutdown Margin Allowed Power Level.

Comment: Critical Surveillance lists 54%.

SAT / UNSAT TASK ELEMENT 7 STANDARD 7.3.1.6 Verify Shutdown Margin greater than or equal to that required by Shutdown Margin is not the COLR by verifying that current power level is less than or equal to the met.

Shutdown Margin Allowed Power Level.

Comment: Critical Surveillance lists 54% allowed, which is greater than the current power level. Applicant must recognize that the actual limit is 22% - 29.5% which SAT / UNSAT is below the current power level.

TASK ELEMENT 8 STANDARD 7.3.2 If Shutdown Margin does not meet the requirements of Technical Specifications, then Commence Emergency Boration and go to OP-901- Direct Emergency Boration.

103, Emergency Boration.

Comment: Critical SAT / UNSAT END OF TASK Revision 1 Page 6 of 7 2011 NRC Exam

JPM A5 SIMULATOR OPERATOR INSTRUCTIONS There is no Simulator setup for this JPM Revision 1 Page 7 of 7 2011 NRC Exam

10.3 SHUTDOWN MARGIN VERIFICATION W ORK SHEET FOR UNTRIPPABLE CEA A5 Key 7.3.1.1 Net Worth WPSO 5.747 K/K 7.3.1.2 Shutdown Margin required by COLR 5.15 K/K 7.3.1.3 Shutdown Margin Allowed Power Defect K/K step 7.3.1.1 ( 5.747 ) - step 7.3.1.2 ( 5.15 ) 0.597 K/K 7.3.1.4 Current Reactor Power 35 %Power 7.3.1.5 Shutdown Margin Allowed Power Level 27% Power 7.3.1.6 Current Power Level Shutdown Margin Allowed Power Level (Circle one) YES NO REMARKS: Emergency Boration required due to not meeting Shutdown Margin. (not required)

Performed by: Joe Operator 3/21/2011 (Signature) (Date)

IV of Calculations by: Bob E Verifier 3/21/2011 (Signature) (Date)

SM/CRS Review: /

(Signature) (Date/Time)

OP-903-090 Revision 301 Attachment 10.3 (1 of 1) 34

10.3 SHUTDOWN MARGIN VERIFICATION W ORK SHEET FOR UNTRIPPABLE CEA A5 Student 7.3.1.1 Net Worth WPSO 5.747 K/K 7.3.1.2 Shutdown Margin required by COLR 4.6 K/K 7.3.1.3 Shutdown Margin Allowed Power Defect K/K step 7.3.1.1 ( 5.747 ) - step 7.3.1.2 ( 4.6 ) 1.147 K/K 7.3.1.4 Current Reactor Power 35 %Power 7.3.1.5 Shutdown Margin Allowed Power Level 54% Power 7.3.1.6 Current Power Level Shutdown Margin Allowed Power Level (Circle one) YES NO REMARKS: None Performed by: Joe Operator 3/21/2011 (Signature) (Date)

IV of Calculations by: Bob E Verifier 3/21/2011 (Signature) (Date)

SM/CRS Review: /

(Signature) (Date/Time)

OP-903-090 Revision 301 Attachment 10.3 (1 of 1) 34

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 3.0 PRECAUTIONS AND LIMITATIONS 3.1 PRECAUTIONS 3.1.1 Shutdown Margin shall be greater than or equal to that specified in the Core Operating Limits Report (COLR), as required by Technical Specification 3.1.1.1 or 3.1.1.2.

3.1.2 Shutdown Margin less conservative than specified by Technical Specification 3.1.1.1 or 3.1.1.2 is a Reportable Occurrence.

3.1.3 RHOBAL Program shall be used to determine initial Xenon Reactivity Worth if a Reactor Trip occurs during non-equilibrium Xenon conditions. Post-trip transient Xenon Worth may be obtained by running a RHOBAL poison transient in accordance with section 7.6, RHOBAL Poison Transient of this procedure or by contacting Reactor Engineering.

3.1.4 For worksheets which perform projections in the RHOBAL program, a poison transient must be performed to update the Xenon and Net Samarium worths. If manual Xenon and/or Net Samarium data is input, the projection will not be performed.

3.2 LIMITATIONS 3.2.1 Information from Plant Data Book (PDB) and Reactor Engineering Book is necessary to perform this procedure unless using RHOBAL. Figure numbers contained in this procedure refer to appropriate section of the PDB. When using graphs and tables in the Plant Data Book (PDB), to obtain the necessary data, it may be necessary and is acceptable to interpolate (approximate between data points or curves). However, extrapolation (approximation outside of the bounds of the data or curves) should not be used.

3.2.2 When using graphs and tables in the PDB, the Beginning of Cycle (BOC), Peak Boron, Middle of Cycle (MOC), and End of Cycle (EOC) periods are defined as follows:

BOC = <30 EFPD Peak Boron = 30 EFPD up to 170 EFPD MOC = 170 EFPD up to 340 EFPD EOC = 340 EFPD 4

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 3.2.3 When using Xenon Worth graphs and tables in the Plant Data Book (PDB), use the figure or table associated with the Effective Full Power Days (EFPD) range listed below:

Figure 1.6.3.1: 0 up to 170 EFPD (BOC and Peak Boron)

Figure 1.6.3.2: 170 up to 340 EFPD (MOC)

Figure 1.6.3.3: 340 EFPD (EOC) 3.2.4 Column F of Attachment 11.1 of OP-004-019, Estimated Critical Configuration, can be used to satisfy Shutdown Margin in Mode 3 per either Technical Specification 4.1.1.1 (5.15% k/k when Shutdown Bank CEAs are not fully inserted) or Technical Specification 4.1.1.2 (4.6% k/k when all CEAs are fully inserted) as directed by OP-010-003, Plant Startup [CRs 98-0970, 01-0209]. This is done by verifying that the actual RCS boron concentration is no more than 20 ppm below Critical Boron Concentration of Column F, Att.11.1 and the question of Allowable CEA Range is verified to be above Transient Insertion Limit for critical operations (Group 5 60 inches) is answered yes.

3.2.5 In the RHOBAL program when calculating Critical Boron Concentration, the adjusted boron concentration column of step 4.2 on Worksheet 1 for Critical Boron Concentration of OP-004-019, Estimated Critical Configuration, can be used to satisfy Shutdown Margin in Mode 3 per either Technical Specification 4.1.1.1 (5.15% k/k when Shutdown Bank CEAs are not fully inserted) or Technical Specification 4.1.1.2 (4.6% k/k when all CEAs are fully inserted) as directed by OP-010-003, Plant Startup [CRs 98-0970, 01-0209]. This is done by verifying that the Actual RCS Boron Concentration is no more than 20 ppm below the Estimated Critical Boron in step 4.2 of Worksheet 1 for Critical Boron Concentration, and the

-0.5% k/k rod position in step 5.3 of Worksheet 1 for Critical Rod Position is above the Transient Insertion Limit for critical operations (Group 5 60 inches).

3.2.6 In the RHOBAL program, if an input parameter is beyond the range of the cycle specific input database, a warning message is printed on the screen and/or the error log. This is intended to prevent performing a calculation outside the analyzed window.

3.2.7 In the RHOBAL program, screen minimization is not allowed. When the calculations are completed on a particular screen, it must be closed for control to return to a previous screen.

3.2.8 Changes to this procedure shall be reviewed by the Reactor Engineering (RE)

Department prior to approval. [P-21855]

5

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 6.0 ACCEPTANCE CRITERIA 6.1 Shutdown Margin is that specified in the COLR by either:

[T.S. 3.1.1.2, T.S. 4.1.1.1.1.a, T.S. 4.1.1.1.1.e]

6.1.1 Current Boron Concentration is Shutdown Margin Boron Concentration.

or 6.1.2 If Reactor is critical with no Untrippable CEAs, and all CEAs are above Transient Insertion Limit. (Operation outside the Transient Insertion Limit is allowed up to two hours per Technical Specification 3.1.3.6).

or 6.1.3 For Dropped or Untrippable CEA, Current Power Level is Shutdown Margin Allowed Power Level.

6.1.4 Current Shutdown Margin is required Shutdown Margin.

8

Surveillance Procedure OP-903-090 Shutdown Margin Revision 301 7.3 SHUTDOWN MARGIN VERIFICATION - UNTRIPPABLE CEA 7.3.1 If an Untrippable CEA Condition exists and the other CEAs are not inserted, then determine Shutdown Margin and record on Attachment 10.3 as follows:

NOTE (1) Use 541 F when using PDB Figure 1.5.7.

(2) When using graphs and tables in the Plant Data Book (PDB), to obtain the necessary data, it may be necessary and is acceptable to interpolate (approximate between data points or curves). However, extrapolation (approximation outside of the bounds of the data or curves) should not be used.

7.3.1.1 Using current Cycle Burnup and 541oF temperature, determine Net Worth Worst Pair Stuck out (WPSO) from Figure 1.5.7.

7.3.1.2 Determine Shutdown Margin required by COLR.

7.3.1.3 Subtract Step 7.3.1.2 from Step 7.3.1.1 to determine Shutdown Margin Allowed Power Defect % K/K.

7.3.1.4 Record current Reactor Power on Attachment 10.3.

7.3.1.5 Using result from step 7.3.1.3 and Power Defect vs. Power Level, Figure 1.2.1, Determine Shutdown Margin Allowed Power Level.

7.3.1.6 Verify Shutdown Margin greater than or equal to that required by the COLR by verifying that current power level is less than or equal to the Shutdown Margin Allowed Power Level.

7.3.2 If Shutdown Margin does not meet the requirements of Technical Specifications, then Commence Emergency Boration and go to OP-901-103, Emergency Boration.

NOTE Subsection 7.4, Shutdown Margin Verification - Untrippable CEA, Other CEAs Inserted, is applicable when all other CEAs are inserted.

7.3.3 If the Reactor has been shutdown less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then determine Required Shutdown Margin Boron Concentration required to meet Shutdown Margin for the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by performing Subsection 7.4, Shutdown Margin Verification -

Untrippable CEA, Other CEAs Inserted.

15

10.3 SHUTDOWN MARGIN VERIFICATION W ORK SHEET FOR UNTRIPPABLE CEA 7.3.1.1 Net Worth WPSO K/K 7.3.1.2 Shutdown Margin required by COLR K/K 7.3.1.3 Shutdown Margin Allowed Power Defect K/K step 7.3.1.1 ( ) - step 7.3.1.2 ( ) K/K 7.3.1.4 Current Reactor Power %Power 7.3.1.5 Shutdown Margin Allowed Power Level % Power 7.3.1.6 Current Power Level Shutdown Margin Allowed Power Level (Circle one) YES NO REMARKS:

Performed by:

(Signature) (Date)

IV of Calculations by:

(Signature) (Date)

SM/CRS Review: /

(Signature) (Date/Time)

OP-903-090 Revision 301 Attachment 10.3 (1 of 1) 34

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A6 Review and Approve OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data.

Applicant:

Examiner:

JPM A6 JOB PERFORMANCE MEASURE DATA PAGE Task: Review and approve OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data.

Task Standard: Applicant discovers 3 errors on the surveillance affecting power calibrations in accordance with the JPM key.

References:

OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Time Critical: No Validation Time: 20 mins.

2.1.18, Ability to make accurate, clear, and K/A Importance Rating 3.8 concise logs, records, status boards, and SRO reports Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 6 2011 NRC Exam

JPM A6 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data

==

Description:==

Applicant will use the simulator PMC and Core Protection Calculator D to review a completed set of calculations for CPC constants KCAL, TCREF, and TPC. The procedure has provisions to collect 5 sets of data and average them over a 5 minute period, but it is optional at the supervisors request. Only 1 set of data was recorded; this is not an error. The reactor operator applicants have a similar JPM that they will perform related to CPC B; the SRO applicant will work on CPC D.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 6 2011 NRC Exam

JPM A6 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Plant conditions are as you see on the Simulator. Use of the Simulator, RM-11, and Plant Monitoring Computer is acceptable. You are not allowed to use any network computers.

Do not go to a location if another applicant is there.

INITIAL CONDITIONS:

The UFM is in service.

The plant is in steady state operation as displayed on the control panels.

INITIATING CUES:

OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data, has been completed for Channel D.

Data collection from CP-10 and for the Linear Power Potentiometer is not required for this task. N/A has been placed on the surveillance for those items.

Review and approve the provided surveillance The previous CRS directed the reactor operator to gather required data once for column 0, averages were not required.

Revision 1 Page 4 of 6 2011 NRC Exam

JPM A6 The Simulator is required to be in RUN to collect data from the CPCs. There will be some drift on the CPC data during the administration of these JPMs. This drift will be minor and will not be confused with the planned JPM errors.

TASK ELEMENT 1 STANDARD Applicant must identify 3 error that were Review and approve OP-903-001, Attachment 11.18. committed performing the surveillance in accordance with the A6 Key.

Comment: Critical Error 1: PHICAL power was recorded and used as 99.90 vice 99.57. SAT / UNSAT Error 2: The new TCREF was completed using the incorrect TCOLD data. The higher of the 2 T COLD values were used vice the lower of the 2.

Error 3: In the calculation for TPC, the PHICAL value was used in the calculation vice Average Core Power from BSCAL as required.

END OF TASK Revision 1 Page 5 of 6 2011 NRC Exam

JPM A6 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-190 After all simulator generated alarms are clear, verify the simulator in RUN.

Verify the parameters on the PMC match the values on the key.

Revision 1 Page 6 of 6 2011 NRC Exam

A6 KEY 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA DATE _______________ CHANNEL UNDER ADJUSTMENT: A B C D 11.10.3.1 Calculate and record the averages of each parameter in the space provided.

0 1 2 3 4 Average Adjusted BSCAL 99.87 PMC PID-C24230 HI LINEAR POWER BISTABLE 1 VOLTS HIGH LINEAR POWER %

VOLT X 20 PHICAL.

(Calibrated Neutron Flux Power) 99.90 99.57 CPC PID 171 BDT (Static Thermal Power) 99.91 CPC PID 177 Calculations Performed by: Verified by:

Signature Signature Refer to attachment 11.1 Note 9.1 to determine appropriate power indication if linear power is not 35% steady state. Document indication used in Remarks.

If COLSS is Inoperable, then use NE-005-201, Heat Balance Calculations, to determine Secondary Calorimetric Power substitute when PMC or CORE POWER is specified.

Adjusted is the average value plus 8.5% (8% to10%) if adjustments are being made to PHICAL and/or BDT as listed in Notes above steps 11.10.6 or 11.10.7 (refer to Attachment 11.1 Note 9.8). Otherwise N/A this block. Use the Average value, not the Adjusted value, for DVM calculation.

11.10.4 Record the following for channel under adjustment:

TPC (Thermal Power Calibration Constant) 0.84500 CPC PID 064 ...................................................

KCAL (Neutron Flux Power Cal. Constant) 1.0062 CPC PID 065 ...................................................

PCALIB (Secondary Calorimetric Power Used in 100 Latest CPC Power Calibration) CPC PID 104 TC 1 (Loop 1 Cold Leg Temperature) 543.59 CPC PID 160 ...................................................

TC 2 (Loop 2 Cold Leg Temperature) 543.46 CPC PID 161 ...................................................

TCORF (Temp Shadowing Correction Factor) .99874 CPC PID 180 ...................................................

EXCORE LINEAR POWER CALIBRATE 1.55 POTENTIOMETER POSITION ROM ..............

OP-903-001 Revision 042 Attachment 11.18 (1 of 3) 142

A6 KEY 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.5 Calculate the new DVM reading or new potentiometer position as follows:

N/A Avg. Core Power (Step 11.10.3.1)*

DVM (new) = =

20 20

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

N/A DVM (new) =

or Potentiometer position (new) =

Avg. Core Power % (Step 11.10.3.1)* X Old Potentiometer Setting (Step 11.10.4)

Avg. Linear Power % (Step 11.10.4)

N/A X N/A N/A

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

Potentiometer position (new) = ________________

Performed by: N/A Verified by: N/A (Initials) (Initials) 11.10.6.1 Calculate KCAL (new):

(Step 11.10.3.1)* (step 11.10.4) (step 11.10.4)

KCAL Avg. Core Power (%) x KCAL x TCORF

=

(new) Avg. PHICAL (step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.6.

KCAL 99.87 x 1.0062 x .99874

=

(new) 99.90 99.57 KCAL 1.0046 1.0080

=

(new)

KCAL (new) = 1.0080 (CPC PID 065) 11.10.6.3 New TCREF (CPC PID 098) = Minimum TC from step 11.10.4:

TC 1 (CPC PID 160) or TC 2 (CPC PID 161).

TCREF (new) = 543.59 543.46 (CPC PID 098)

Performed by: Verified by:

(Initials) (Initials)

OP-903-001 Revision 042 Attachment 11.18 (2 of 3) 143

A6 KEY 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.7 Calculate TPC (CPC PID 064):

Avg. Core Power % (Step 11.10.3.1)* X TPC (Step 11.10.4)

TPC (new)=

Avg. BDT (Step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.7.

99.90 99.87 X 0.84500 TPC (new)=

99.91 TPC (new) = 0.84492 0.84466 (CPC PID 064).

Performed by: Verified by:

(Initials) (Initials) 11.10.11.1 Record the following:

Applicable CORE POWER PMC....................................___________

PCALIB CPC PID 104 ..............................................___________

HI LINEAR POWER BISTABLE 1 VOLTS .....................___________

HI LINEAR POWER % VOLTS x 20 ..............................___________

PHICAL CPC PID 171 ..............................................___________

BDT CPC PID 177 ..............................................___________

11.10.11.2.1 Record answers:

HI LINEAR POWER %, VOLTS X 20.................... YES/NO ________

PHICAL, CPC PID 171 ..................................... YES/NO ________

BDT, CPC PID 177 ..................................... YES/NO ________

11.10.12 Performed by: Verified by:

(Initials) (Initials) 11.10.13 Reviewed by: _________________________ _________________

SM/CRS Date/Time OP-903-001 Revision 042 Attachment 11.18 (3 of 3) 144

A6 Student Copy 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA DATE _____3/22/2011___ CHANNEL UNDER ADJUSTMENT: A B C D 11.10.3.1 Calculate and record the averages of each parameter in the space provided.

0 1 2 3 4 Average Adjusted BSCAL 99.87 N/A N/A N/A N/A 99.87 PMC PID-C24230 N/A HI LINEAR POWER N/A N/A N/A N/A N/A N/A BISTABLE 1 VOLTS HIGH LINEAR POWER %

N/A N/A N/A N/A N/A N/A VOLT X 20 PHICAL.

(Calibrated Neutron Flux Power) 99.90 N/A N/A N/A N/A 99.90 CPC PID 171 BDT (Static Thermal Power) 99.91 N/A N/A N/A N/A 99.91 CPC PID 177 Calculations Performed by: N/A Verified by: N/A Signature Signature Refer to attachment 11.1 Note 9.1 to determine appropriate power indication if linear power is not 35% steady state. Document indication used in Remarks.

If COLSS is Inoperable, then use NE-005-201, Heat Balance Calculations, to determine Secondary Calorimetric Power substitute when PMC or CORE POWER is specified.

Adjusted is the average value plus 8.5% (8% to10%) if adjustments are being made to PHICAL and/or BDT as listed in Notes above steps 11.10.6 or 11.10.7 (refer to Attachment 11.1 Note 9.8). Otherwise N/A this block. Use the Average value, not the Adjusted value, for DVM calculation.

11.10.4 Record the following for channel under adjustment:

TPC (Thermal Power Calibration Constant) 0.84500 CPC PID 064 ...................................................

KCAL (Neutron Flux Power Cal. Constant) 1.0062 CPC PID 065 ...................................................

PCALIB (Secondary Calorimetric Power Used in 100 Latest CPC Power Calibration) CPC PID 104 TC 1 (Loop 1 Cold Leg Temperature) 543.59 CPC PID 160 ...................................................

TC 2 (Loop 2 Cold Leg Temperature) 543.46 CPC PID 161 ...................................................

TCORF (Temp Shadowing Correction Factor) .99874 CPC PID 180 ...................................................

EXCORE LINEAR POWER CALIBRATE N/A POTENTIOMETER POSITION ROM ..............

OP-903-001 Revision 042 Attachment 11.18 (1 of 3) 142

A6 Student Copy 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.5 Calculate the new DVM reading or new potentiometer position as follows:

Avg. Core Power (Step 11.10.3.1)* N/A DVM (new) = =

20 20

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

DVM (new) = N/A or Potentiometer position (new) =

Avg. Core Power % (Step 11.10.3.1)* X Old Potentiometer Setting (Step 11.10.4)

Avg. Linear Power % (Step 11.10.4)

N/A X N/A N/A

Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

Potentiometer position (new) = ______N/A__________

Performed by: N/A Verified by: N/A (Initials) (Initials) 11.10.6.1 Calculate KCAL (new):

(Step 11.10.3.1)* (step 11.10.4) (step 11.10.4)

KCAL Avg. Core Power (%) x KCAL x TCORF

=

(new) Avg. PHICAL (step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.6.

KCAL 99.87 x 1.0062 x .99874

=

(new) 99.90 KCAL 1.0046

=

(new)

KCAL (new) = 1.0046 (CPC PID 065) 11.10.6.3 New TCREF (CPC PID 098) = Minimum TC from step 11.10.4:

TC 1 (CPC PID 160) or TC 2 (CPC PID 161).

TCREF (new) = 543.59 (CPC PID 098)

Performed by: JVS Verified by: LDK (Initials) (Initials)

OP-903-001 Revision 042 Attachment 11.18 (2 of 3) 143

A6 Student Copy 11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.7 Calculate TPC (CPC PID 064):

Avg. Core Power % (Step 11.10.3.1)* X TPC (Step 11.10.4)

TPC (new)=

Avg. BDT (Step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.7.

99.90 X 0.84500 TPC (new)=

99.91 TPC (new) = 0.84492 (CPC PID 064).

Performed by: JVS Verified by: LDK (Initials) (Initials) 11.10.11.1 Record the following:

Applicable CORE POWER PMC....................................___________

PCALIB CPC PID 104 ..............................................___________

HI LINEAR POWER BISTABLE 1 VOLTS .....................___________

HI LINEAR POWER % VOLTS x 20 ..............................___________

PHICAL CPC PID 171 ..............................................___________

BDT CPC PID 177 ..............................................___________

11.10.11.2.1 Record answers:

HI LINEAR POWER %, VOLTS X 20.................... YES/NO ________

PHICAL, CPC PID 171 ..................................... YES/NO ________

BDT, CPC PID 177 ..................................... YES/NO ________

11.10.12 Performed by: Verified by:

(Initials) (Initials) 11.10.13 Reviewed by: _________________________ _________________

SM/CRS Date/Time OP-903-001 Revision 042 Attachment 11.18 (3 of 3) 144

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A7 Review and Approve an Equipment Out of Service document in accordance with OP-100-010, Equipment Out of Service Applicant:

Examiner:

JPM A7 JOB PERFORMANCE MEASURE DATA PAGE Task: Review and approve a completed Equipment Out of Service document for BD-103 A, S/G 1 Blowdown Outside Containment Isolation.

Task Standard: Applicant identifies applicable Tech Specs and retests associated with BD-103 A, S/G 1 Blowdown Outside Containment Isolation, in accordance with task standards.

References:

Tech Spec 3.6.3 and 3.3.3.6 OP-100-010, Equipment Out of Service Time Critical: No Validation Time: 20 mins.

K/A 2.2.37, Ability to determine operability and/or Importance Rating 4.6 availability of safety related equipment SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 7 2011 NRC Exam

JPM A7 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

Tech Spec 3.6.3 and 3.3.3.6 OP-100-010, Equipment Out of Service

==

Description:==

Applicant will be directed to complete an Equipment Out of Service document (an EOS) for BD-103 A, S/G 1 Blowdown Outside Containment Isolation. The applicant will determine the correct Tech Specs and retest requirements and identify that the current Tech Spec action requirements are not being complied with at that time. This determination will be determined using conditions as displayed on the simulator.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 7 2011 NRC Exam

JPM A7 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Plant conditions are as you see on the Simulator. Use of the Simulator, RM-11, and Plant Monitoring Computer is acceptable. You are not allowed to use any network computers.

INITIAL CONDITIONS:

The Inservice Test Engineer called the Control Room and reported that he received a report of defective parts associated with the packing gland for BD-103 A, S/G 1 Blowdown Outside Containment Isolation.

He has completed his review and the defective parts are installed on BD-103 A.

The valve limit switches are not affected by the defective part.

The ESFAS solenoid for BD-103 A is not affected by the defective part.

Work order 52475640 is being planned to replace the defective parts on BD-103 A.

The Shift Manager has declared BD-103 A INOPERABLE 10 minutes ago.

INITIATING CUES:

The Shift Manager directs you to prepare an EOS for BD-103 A.

Include the necessary retest in your EOS.

Indicate whether there are any Tech Spec actions applicable.

Perform this review and mark any comments on this sheet.

Revision 1 Page 4 of 7 2011 NRC Exam

JPM A7 Evaluator Note The applicants will record their conclusions on the paperwork provided. This paperwork is typically a product of a database that the operator would enter the data into. The A7 Key displays the forms filled out with critical and non-critical data. These task elements describe the critical elements of the applicants response.

TASK ELEMENT 1 STANDARD BD-103 A is listed as the Identify component Equipment Number, Description and System.

inoperable component.

Comment: Critical SAT / UNSAT TASK ELEMENT 2 STANDARD If mode changes are not allowed (T.S. 3.0.4 or TRM 3.0.4 is applicable),

Mode change is allowed.

then indicate by checking NO. Otherwise, check YES.

Comment:

SAT / UNSAT TASK ELEMENT 3 STANDARD RO/SRO/STA identifies TS/TRMs required to be entered for component Tech Spec 3.6.3 entered.

being removed from service for the current Plant Mode.

Comment: Critical SAT / UNSAT Revision 1 Page 5 of 7 2011 NRC Exam

JPM A7 TASK ELEMENT 4 STANDARD Enter a brief description of the Limiting condition for operation Description listed.

Comment:

SAT / UNSAT TASK ELEMENT 5 STANDARD Applicant lists that the component must be Enter a brief description of the Required Action(s) to be taken. declared operable or isolated within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in accordance with key.

Comment: Critical Blowdown is listed as a closed system in tech Spec 3.6.3 bases. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months action does not apply to BD-103 A. SAT / UNSAT TASK ELEMENT 6 STANDARD Tech Spec 3.3.3.6 RO/SRO/STA identifies applicable TS/TRMs not required to be entered, applicable but entry not with justification and any notes needed to conduct a brief.

required.

Comment:

3.3.3.6 entry is not required with the provisions of Tech Spec 3.6.3 being complied with. SAT / UNSAT TASK ELEMENT 7 STANDARD If an operability retest is required, then enter a brief description of the OP-903-119 listed as the operability retest, otherwise enter NONE. applicable retest.

Comment: Critical SAT / UNSAT END OF TASK Revision 1 Page 6 of 7 2011 NRC Exam

JPM A7 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-190 After all simulator generated alarms are clear, place the simulator in Run.

Revision 1 Page 7 of 7 2011 NRC Exam

A7 Key 7.1 TS/TRM ENTRY GUIDELINE

1) Inoperable Equipment/System: BD-103 A, S/G 1 Blowdown Outside Containment Isolation

2) Reason: Defective Parts

3) Licensed Operator, Senior Reactor Operator or Shift Technical Advisor review Technical Specifications and the Technical Requirements Manual to determine if an entry is required.

Consider the following questions during the review:

Does this work affect a Containment Isolation Valve listed in TRM Table 3.6-2?

Does this work affect a component that requires entry into the following TS/TRM?

TS 3.3.1, Reactor Protection TS 3.3.2, Engineering Safeguards TS 3.3.3.1 or TRM 3.3.3.1, 3.3.3.10/11, Radiation Monitoring TS 3.3.3.5, Remote Shutdown TS 3.3.3.6, Accident Monitoring including Valve Position Indication TS 3.8.4.1, Overcurrent Protective Device Has OP-100-014 been reviewed? (i.e. Room Coolers, cascading TSs, OP-903-066, etc.)

Does this work affect a Dual Function Valve? (See Attachment 7.4)

Is the Redundant train/component operable and has a board walkdown been performed on the redundant train? (Reference Section 4.0 Responsibilities)

Have restoration requirements been reviewed?

Is the Component part of the IST Program and has the IST retest been identified?

Does this work affect an instrument used to take Technical Specification Logs? (OP-903-001)

Will invoking TS 3.0.5 be necessary to demonstrate operability?

If an emergent equipment unavailability, then has a risk assessment been performed?

BRIEF NOTES

4) TS/TRM ENTERED: 3.6.3

5) NOTES Restore within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or isolate the penetration within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by use of at least one deactivated automatic valve secured in the isolation position or by use by use of at least one closed manual valve.

(Signatures not required if documented on Attachment 7.2, EOS Checklist) 5a) Reviewed by: ________________________

6) SM/CRS or authorizing SRO reviews TS/TRM entry to ensure they are appropriate.

6a) Approved by: ________________________

7) Assemble the correct personnel and conduct a shift brief.

8) Log all Technical Specification and Technical Requirements Manual entries in Station Log.

9) Update EOOS program if required.

10) If the work requires an Operability retest other than OP-903-001 channel check, or the work will be carried into the next shift, then verify Attachment 7.2, EOS Checklist, is completed.

OP-100-010 Revision 304 Attachment 7.1 (1 of 1) 19

A7 Key 7.2 EOS CHECKLIST

1) EOS Checklist No: ___________________

2) Equipment Tag Number. BD MVAAA103 A (BD-103 A) 2) System Blowdown (BD)

2) Description of the component and reason for being Inoperable:

Defective parts installed.

3) % Reactor 5) Mode Change 100% 4) Mode 1 Yes X No Power Allowed

6) TS Addendum Sheet Yes X No Required

7) Department Action Notice Yes No X 7) DAN No. N/A

8) Equipment Outage Planned Yes No X If the Planned Block is checked No, notify Shift Manager to perform a risk assessment in accordance with OI-037-000, Operations Risk Assessment Guideline, for emergent equipment unavailability within the scope of W2.502, Configuration Risk Management Program Implementation

9) Compensatory Actions / Comments: None EOS OPENING

22) REVIEWED BY:

RO/SRO/STA SIGNATURE

23) APPROVED BY:

SM/CRS (or authorizing SRO) SIGNATURE DATE/TIME Print SM/CRS Contacted (may be N/A)

EOS CLOSURE

24) REVIEWED BY:

RO/SRO/STA SIGNATURE

25) APPROVED BY:

SM/CRS (or authorizing SRO) SIGNATURE DATE/TIME Print SM/CRS Contacted (may be N/A)

OP-100-010 Revision 304 Attachment 7.2 (1 of 3) 20

A7 Key EOS CHECKLIST (CONTD)

11) EQUIPMENT TAG 13) OPERABILITY RETEST 14) DOC COMPLETED SAT

10) DOC # 10) TYPE NUMBER (SIGNATURE / DATE /

12) WORK DESCRIPTION TIME) 52475640 WO BD-103 A OP-903-119, Secondary Auxiliaries Quarterly IST Valve Tests OP-100-010 Revision 304 Attachment 7.2 (2 of 3) 21

A7 Key EOS CHECKLIST (CONTD)

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No 3.6.3 Each Containment Isolation shall be Restore the inoperable valve to Yes X No operable. OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or

19) Component Isolate each affected penetration 20) Justification / Comments within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by use of at least one BD-103 A deactivated automatic valve secured None in the isolation position and verify the affected penetration flow path is isolated once per 31 days, or Isolate each affected penetration within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by use of at least one closed manual valve or blind flange and verify the affected penetration flow path is isolated once per 31 days.

21) Mode Info Tech Spec 3.0.4 does not apply

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No 3.3.3.6 Position indication for every Restore within 30 days. Yes No X containment isolation valve shall be

19) Component operable. 20) Justification / Comments BD-103 A 3.3.3.6 not required with provisions of Tech Spec 3.6.3 being complied with.

21) Mode Info Tech Spec 3.0.4 does not apply.

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No Yes No

19) Component 20) Justification / Comments

21) Mode Info

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No Yes No

19) Component 20) Justification / Comments

21) Mode Info OP-100-010 Revision 304 Attachment 7.2 (3 of 3) 22

(Typical) 7.1 TS/TRM ENTRY GUIDELINE

1) Inoperable Equipment/System:

2) Reason:

3) Licensed Operator, Senior Reactor Operator or Shift Technical Advisor review Technical Specifications and the Technical Requirements Manual to determine if an entry is required.

Consider the following questions during the review:

Does this work affect a Containment Isolation Valve listed in TRM Table 3.6-2?

Does this work affect a component that requires entry into the following TS/TRM?

TS 3.3.1, Reactor Protection TS 3.3.2, Engineering Safeguards TS 3.3.3.1 or TRM 3.3.3.1, 3.3.3.10/11, Radiation Monitoring TS 3.3.3.5, Remote Shutdown TS 3.3.3.6, Accident Monitoring including Valve Position Indication TS 3.8.4.1, Overcurrent Protective Device Has OP-100-014 been reviewed? (i.e. Room Coolers, cascading TSs, OP-903-066, etc.)

Does this work affect a Dual Function Valve? (See Attachment 7.4)

Is the Redundant train/component operable and has a board walkdown been performed on the redundant train? (Reference Section 4.0 Responsibilities)

Have restoration requirements been reviewed?

Is the Component part of the IST Program and has the IST retest been identified?

Does this work affect an instrument used to take Technical Specification Logs? (OP-903-001)

Will invoking TS 3.0.5 be necessary to demonstrate operability?

If an emergent equipment unavailability, then has a risk assessment been performed?

BRIEF NOTES

4) TS/TRM ENTERED:

5) NOTES (Signatures not required if documented on Attachment 7.2, EOS Checklist) 5a) Reviewed by: ________________________

6) SM/CRS or authorizing SRO reviews TS/TRM entry to ensure they are appropriate.

6a) Approved by: ________________________

7) Assemble the correct personnel and conduct a shift brief.

8) Log all Technical Specification and Technical Requirements Manual entries in Station Log.

9) Update EOOS program if required.

10) If the work requires an Operability retest other than OP-903-001 channel check, or the work will be carried into the next shift, then verify Attachment 7.2, EOS Checklist, is completed.

OP-100-010 Revision 304 Attachment 7.1 (1 of 1) 19

(Typical) 7.2 EOS CHECKLIST

1) EOS Checklist No: ___________________

2) Equipment Tag Number. 2) System

2) Description of the component and reason for being Inoperable:

3) % Reactor 5) Mode Change

4) Mode Yes No Power Allowed

6) TS Addendum Sheet Yes No Required

7) Department Action Notice Yes No 7) DAN No.

8) Equipment Outage Planned Yes No If the Planned Block is checked No, notify Shift Manager to perform a risk assessment in accordance with OI-037-000, Operations Risk Assessment Guideline, for emergent equipment unavailability within the scope of W2.502, Configuration Risk Management Program Implementation

9) Compensatory Actions / Comments:

EOS OPENING

22) REVIEWED BY:

RO/SRO/STA SIGNATURE

23) APPROVED BY:

SM/CRS (or authorizing SRO) SIGNATURE DATE/TIME Print SM/CRS Contacted (may be N/A)

EOS CLOSURE

24) REVIEWED BY:

RO/SRO/STA SIGNATURE

25) APPROVED BY:

SM/CRS (or authorizing SRO) SIGNATURE DATE/TIME Print SM/CRS Contacted (may be N/A)

OP-100-010 Revision 304 Attachment 7.2 (1 of 3) 20

(Typical)

EOS CHECKLIST (CONTD)

11) EQUIPMENT TAG 13) OPERABILITY RETEST 14) DOC COMPLETED SAT

10) DOC # 10) TYPE NUMBER (SIGNATURE / DATE /

12) WORK DESCRIPTION TIME)

OP-100-010 Revision 304 Attachment 7.2 (2 of 3) 21

(Typical)

EOS CHECKLIST (CONTD)

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No Yes No

19) Component 20) Justification / Comments

21) Mode Info

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No Yes No

19) Component 20) Justification / Comments

21) Mode Info

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No Yes No

19) Component 20) Justification / Comments

21) Mode Info

15) TS/TRM 16) Condition 17) Required Action 18) TS/TRM Entry? Yes/No Yes No

19) Component 20) Justification / Comments

21) Mode Info OP-100-010 Revision 304 Attachment 7.2 (3 of 3) 4

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A8 Plan Work and Assign Workers Based on Dose Rates and Shielding Applicant:

Examiner:

JPM A8 JOB PERFORMANCE MEASURE DATA PAGE Task: Plan Work and Assign Workers Based on Dose Rates and Shielding Task Standard: Applicant calculates dose with and without shielding, and with 1 or 2 workers, and directs job to achieve the lowest dose.

References:

None Time Critical: No Validation Time: 30 mins.

K/A 2.3.4, Knowledge of radiation exposure limits Importance Rating 3.7 under normal and emergency conditions. SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 7 2011 NRC Exam

JPM A8 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

None READ TO APPLICANT DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

Refuel 17 is in progress and Safety Injection Train A is being aligned from its Shutdown Cooling alignment to its Safety Injection alignment. You are the Work Management Center SRO and have been assigned to coordinate the valve line up verification for Safety Injection Train A.

The dose rates in Safeguards Room A are 365 mrem/hour unshielded.

Installing shielding will reduce the dose rate to 130 mrem/hour.

It will take 2 workers 10 minutes to install the shielding (10 minutes each worker).

It will take 1 person 30 minutes to complete the valve verification if he works alone.

It will take 2 people 20 minutes to complete the valve verification (20 minutes each worker).

How will you direct the execution of the Safety Injection System venting to allow the least amount of total worker dose?

Revision 0 Page 3 of 7 2011 NRC Exam

JPM A8 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Refuel 17 is in progress and Safety Injection Train A is being aligned from its Shutdown Cooling alignment to its Safety Injection alignment. You are the Work Management Center SRO and have been assigned to coordinate the valve line up verification for Safety Injection Train A.

The dose rates in Safeguards Room A are 365 mrem/hour unshielded.

Installing shielding will reduce the dose rate to 130 mrem/hour.

It will take 2 workers 10 minutes to install the shielding (10 minutes each worker).

It will take 1 person 30 minutes to complete the valve verification if he works alone.

It will take 2 people 20 minutes to complete the valve verification (20 minutes each worker).

How will you direct the execution of the Safety Injection System valve verification to allow the least amount of total worker dose?

Show all calculations to support your answer.

Revision 0 Page 4 of 7 2011 NRC Exam

JPM A8 TASK ELEMENT STANDARD Applicant calculated 182.5 mrem total Calculate dose for 1 worker with no shielding installed.

dose.

Comment: Critical SAT / UNSAT TASK ELEMENT STANDARD Applicant calculated 243.4 mrem total Calculate dose for 2 workers with no shielding installed.

dose.

Comment: Critical SAT / UNSAT TASK ELEMENT STANDARD Applicant calculated 186.6 mrem total Calculate dose for 1 worker with shielding installed.

dose.

Comment: Critical Installing the shielding results in 121.6 mrem total dose.

SAT / UNSAT Performing the work with 1 worker will result in 65 mrem +

121.6 mrem = total dose of 186.6 mrem TASK ELEMENT STANDARD Applicant calculated 208.3 mrem total Calculate dose for 2 workers with shielding installed.

dose.

Comment: Critical Installing the shielding results in 121.6 mrem total dose.

SAT / UNSAT Performing the job with 2 workers will result in 86.6 mrem +

121.6 mrem = total dose of 208.3 mrem Revision 0 Page 5 of 7 2011 NRC Exam

JPM A8 TASK ELEMENT STANDARD Job assigned to 1 workers without Applicant assigns job.

installing shielding.

Comment: Critical SAT / UNSAT Performing the job with 1 worker and no shielding will result in 182.5 mrem total dose.

END OF TASK Revision 0 Page 6 of 7 2011 NRC Exam

JPM A8 SIMULATOR OPERATOR INSTRUCTIONS There is no simulator setup for this JPM.

Revision 0 Page 7 of 7 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE A9 Determine Actions for a Toxic Chemical Release Applicant:

Examiner:

JPM A9 JOB PERFORMANCE MEASURE DATA PAGE Task: Determine Actions for a Toxic Chemical Release Task Standard: Applicant determines Procedure Tab B, Site Evacuation is to be implemented based on the given conditions. Applicants response to flow chart is in accordance with answer key.

References:

EP-004-010, Toxic Chemical Contingency Procedure Time Critical: No Validation Time: 15 mins.

K/A 2.4.38 Ability to take actions called for in the Importance Rating 4.4 facility emergency plan, including supporting SRO or acting as emergency coordinator if.

required Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 6 2011 NRC Exam

JPM A9 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

EP-004-010, Toxic Chemical Contingency Procedure Simulator setup to preset Initial Condition (IC).

==

Description:==

This JPM requires the candidate to determine the correct Procedure Tab to implement as a result of an Offsite Toxic Chemical Release. The JPM will be performed in the simulator using data obtained from the Plant Monitoring Computer and Initial Conditions. Ensure that the examinee turns in all paperwork prior to releasing.

READ TO APPLICANT DIRECTION TO APPLICANT:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 6 2011 NRC Exam

JPM A9 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

Plant conditions are as you see on the Simulator. Use of the Simulator, RM-11, and Plant Monitoring Computer is acceptable. You are not allowed to use any network computers.

INITIAL CONDITIONS:

You are the dayshift Shift Manager.

A release of Hydrochloric Acid from Hexion Chemical has been confirmed via the Industrial Hotline.

The release is still in progress, duration is unknown.

The release started 5 minutes ago.

The plant is at normal dayshift staffing.

Weather conditions are good.

EP-004-010, Toxic Chemical Contingency Procedure, Section 5.2, Initial Assessment of has been completed through Step 5.2.1.2.

INITIATING CUES:

Determine Procedure Tab to be implemented in accordance with EP-004-010, Toxic Chemical Contingency Procedure.

Document results on Applicant Cue Sheet and return the Cue Sheet and all paperwork used to support your decision to the examiner.

Revision 1 Page 4 of 6 2011 NRC Exam

JPM A9 Evaluator Note This JPM will be performed in the simulator using data obtained from the Plant Monitoring Computer and Initial Conditions. Ensure that the examinee turns in all paperwork prior to releasing.

TASK ELEMENT 1 STANDARD Using EP-004-010, the candidate determines that Evacuation Tab B Tab B, Site Evacuation, should be implemented. selected Comment: Critical Refer to A9 Key for details of each step.

SAT / UNSAT .2 Flow Chart Toxic Chemicals within the EAB? NO Waterford < 5 miles from release YES Large or unknown hazard? YES Released secured? NO Conduct Protective Response evaluation, Attachment 7.4.

Waterford down wind from the release? YES Response time > 45 minutes? YES Consider Evacuation, Tab B .4 Response Evaluation Wind direction should be obtained from PMC point C48530 as 120 degrees. Worst case wind direction for Hexion is 80 degrees. The +/- 45 degree area is 165 to 75 degrees. Waterford is downwind. Using other wind direction points will lead to the wrong conclusion.

Plume travel time is the distance from Waterford, 2.75 miles, and the wind speed from PMC point C48526, 1.21 m/s. Using the provided conversion factor, this results in a 60.9 minute response time. Given that the release started 5 minutes ago, the plant response time will be 55.9 minutes.

Using the wrong wind speed PMC point will give a response time of < 45 minutes.

END OF TASK Revision 1 Page 5 of 6 2011 NRC Exam

JPM A9 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-190 After all simulator generated alarms are clear, place the simulator in Run.

Verify the parameters on the PMC match the values on the key.

Revision 1 Page 6 of 6 2011 NRC Exam

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level Reactor Operator Operating Test No.: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Safety Code* Function S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in A, D, S 1 accordance with OP-903-005, Control Element Assembly Operability Check.

Fault: CEA 20 will insert after initially moved, CEA will subsequently drop, the combination requiring a reactor trip.

A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 S2 004 Chemical and Volume Control System; Makeup to the Volume A, M, S 2 Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control.

Fault: The Boric Acid counter will fail to secure the Boric Acid addition, requiring the applicant to manually secure Boric Acid flow. The applicant will then need to add the Primary Makeup Water for the initial calculation, plus the additional based on the extra boric acid added.

A4.07 Boration/dilution RO - 3.9, SRO - 3.7 S3 005 Shutdown Cooling System; Secure Shutdown Cooling Train B and D, L, S 4-P place it in standby in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 S4 039 Main and Reheat Steam System; BOP operator immediate operator A, M, S 4-S actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 028 Hydrogen Recombiner and Purge Control System D, L, P, S 5 Start Hydrogen Recombiner A in accordance with OP-008-006.

A4.01 HRPS controls RO - 4.0, SRO - 4.0 S6 064 Emergency Diesel Generator (ED/G) System; Parallel Emergency A, D, S 6 Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator.

Fault: After EDG A load is raised, EDG A load will rise without manipulation requiring a trip of EDG A.

A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 S7. 029 Containment Purge System; Perform surveillance OP-903-052, N, S 8 Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A.

K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 S8. 012 Reactor Protection System; Place Reactor Power Cutback in service D, S 7 and remove reactor trip on turbine trip in accordance with OP-004-015, Reactor Power Cutback.

A4.03 Channel blocks and bypasses RO - 3.6, SRO - 3.6 1 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Emergency Feedwater System; Reset overspeed device on D, E, L, 4-S Emergency Feedwater Pump AB in accordance with OP-902-005, P, R Station Blackout Recovery.

A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency A, D, R 6 Diesel Generator B locally.

Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B.

K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 P3 068 Control Room Evacuation E, L, N 2 Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0

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

Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 7 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) -

(L)ow-Power / Shutdown 1/ 1/ 1 4 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 4 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 2 (R)CA 1/ 1/ 1 2 (S)imulator 8 2 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level SRO - Instant Operating Test No.: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Safety Code* Function S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in A, D, S 1 accordance with OP-903-005, Control Element Assembly Operability Check.

Fault: CEA 20 will insert after initially moved, CEA will subsequently drop, the combination requiring a reactor trip.

A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 S2 004 Chemical and Volume Control System; Makeup to the Volume A, M, S 2 Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control.

Fault: The Boric Acid counter will fail to secure the Boric Acid addition, requiring the applicant to manually secure Boric Acid flow. The applicant will then need to add the Primary Makeup Water for the initial calculation, plus the additional based on the extra boric acid added.

A4.07 Boration/dilution RO - 3.9, SRO - 3.7 S3 005 Shutdown Cooling System; Secure Shutdown Cooling Train B and D, L, S 4-P place it in standby in accordance with OP-009-005, Shutdown Cooling.

A4.01 Controls and indication for RHR pumps RO - 3.6, SRO - 3.4 S4 039 Main and Reheat Steam System; BOP operator immediate operator A, M, S 4-S actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 S6 064 Emergency Diesel Generator (ED/G) System; Parallel Emergency A, D, S 6 Diesel Generator A for EDG testing in accordance with OP-009-002, Emergency Diesel Generator.

Fault: After EDG A load is raised, EDG A load will rise without manipulation requiring a trip of EDG A.

A4.06 Manual start, loading, and stopping of the ED/G RO - 3.9, SRO - 3.9 S7. 029 Containment Purge System; Perform surveillance OP-903-052, N, S 8 Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A.

K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 S8. 012 Reactor Protection System; Place Reactor Power Cutback in service D, S 7 and remove reactor trip on turbine trip in accordance with OP-004-015, Reactor Power Cutback.

A4.03 Channel blocks and bypasses RO - 3.6, SRO - 3.6 3 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 061 Emergency Feedwater System; Reset overspeed device on D, E, L, 4-S Emergency Feedwater Pump AB in accordance with OP-902-005, P, R Station Blackout Recovery.

A2.04 Pump failure or improper operation RO - 3.4, SRO - 3.8 P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency A, D, R 6 Diesel Generator B locally.

Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B.

K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 P3 068 Control Room Evacuation E, L, N 2 Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0

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

Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 5 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 6 (E)mergency or abnormal in-plant 1/ 1/ 1 2 (EN)gineered safety feature - / - / 1 (control room system) -

(L)ow-Power / Shutdown 1/ 1/ 1 3 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 4 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 1 (R)CA 1/ 1/ 1 2 (S)imulator 7 4 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: WATERFORD 3 Date of Examination: March 21, 2011 Exam Level SRO - Upgrade Operating Test No.: NRC Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Safety Code* Function S1 001 Control Rod Drive, Perform CEA testing for Regulating Group 6 in A, D, S 1 accordance with OP-903-005, Control Element Assembly Operability Check.

Fault: CEA 20 will insert after initially moved, CEA will subsequently drop, the combination requiring a reactor trip.

A4.01 Controls for CCWS RO - 3.1, SRO - 2.9 S2 S3 S4 039 Main and Reheat Steam System; BOP operator immediate operator A, M, S 4-S actions on evacuation of the Control Room in accordance with OP-901-502, Control Room Evacuation Fault: Atmospheric Dump Valve B will spuriously open, requiring the applicant to take contingency actions to control Steam Generator pressure.

A4.01 Main steam supply. Valves RO - 2.9, SRO - 2.8 S5 S6 S7. 029 Containment Purge System; Perform surveillance OP-903-052, N, EN, S 8 Controlled Ventilation Area System Operability Check, and secure RAB Normal Ventilation and start CVAS Train A.

K1.03 Engineering safeguards RO - 3.6, SRO - 3.8 S8.

5 NRC 2011 Revision 1

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U)

P1 P2 064 Emergency Diesel Generator (ED/G) System; Trip Emergency A, D, R 6 Diesel Generator B locally.

Fault: The first method the applicant performs to trip the EDG B will fail, requiring contingency actions to secure EDG B.

K4.02 Trips for ED/G while operating (normal or emergency)

RO - 3.9, SRO - 4.2 P3 068 Control Room Evacuation E, L, N 2 Close Train B Safety Injection Tank outlet valves during a Control Room Evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

AA1.28 PZR level control and pressure control RO - 3.8, SRO - 4.0

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

Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 / 2-3 3 (C)ontrol room 0 (D)irect from bank 9/ 8/ 4 2 (E)mergency or abnormal in-plant 1/ 1/ 1 1 (EN)gineered safety feature - / - / 1 (control room system) 1 (L)ow-Power / Shutdown 1/ 1/ 1 1 (N)ew or (M)odified from bank including 1(A) 2/ 2/ 1 3 (P)revious 2 exams 3/ 3/ 2 (randomly selected) 0 (R)CA 1/ 1/ 1 1 (S)imulator 3 6 NRC 2011 Revision 1

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S1 Perform CEA Operability Checks Applicant:

Examiner:

JPM S1 JOB PERFORMANCE MEASURE DATA PAGE Task: Perform CEA testing for Regulating Group 6 in accordance with OP-903-005, Control Element Assembly Operability Check.

Task Standard: Applicant cycled Regulating Group CEA 20 in accordance with OP-903-005 and identified continuous motion. Applicant recognized CEA 8 dropped and took immediate operator action to trip reactor.

References:

OP-903-005, Control Element Assembly Operability Check OP-901-102, CEA or CEDMCS Malfunction Alternate Path: Yes Time Critical: No Validation Time: 15 mins.

K/A 001 A4.01 Controls for CCWS Importance Rating 3.1 / 2.9 RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 10 2011 NRC Exam

JPM S1 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-005, Control Element Assembly Operability Check

==

Description:==

The applicant will be directed to complete OP-903-005. Regulating Group 6 and P remain to be cycled to complete the surveillance. When CEA 20 is cycled, it will continuously insert. When CEA 20 reaches 140 inches, CEA 8 will drop. The applicant must determine that 2 CEAs are misaligned by > 19 inches and trip the reactor.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 10 2011 NRC Exam

JPM S1 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

Plant is at 100% power OP-903-005, Control Element Assembly Operability Check, is in progress All CEAs have been cycled except for Regulating Group 6 and P CEAs are being parked at the Upper Electrical Limit.

INITIATING CUES:

The CRS directs you to resume performance of OP-903-005 and cycle Regulating Group 6 followed by cycling Group P CEAs.

Revision 1 Page 4 of 10 2011 NRC Exam

JPM S1 Evaluator Note Cue the Simulator Operator to place the Simulator in RUN.

TASK ELEMENT 1 STANDARD 7.1.3 For each CEA inserted 5 inches for Axial Shape Index control, perform the following on Attachment 10.1:

7.1.3.1 Record CEA position prior to ASI control as Initial Position.

Step reviewed 7.1.3.2 Record present CEA Position as Test Position.

7.1.3.3 Document CEA movement 5 inches.

7.1.3.4 Record NA for Final Position.

Comment:

No CEAs will be inserted > 5 inches; this step only needs to be reviewed.

SAT / UNSAT TASK ELEMENT 2 STANDARD 7.1.4 On Attachment 10.1, check the CEAC that will be used for N/A indication.

Comment:

This was performed for the previous CEAs moved in the setup.

SAT / UNSAT TASK ELEMENT 3 STANDARD 7.1.5 Perform Attachment 10.2 for each CEA inserted <5 inches for Axial Acknowledge step Shape Index control.

Comment:

SAT / UNSAT Revision 1 Page 5 of 10 2011 NRC Exam

JPM S1 TASK ELEMENT 4 STANDARD 10.2.1 Position Individual CEA Selection switches to select desired CEA: CEA 20 is selected Comment: Critical SAT / UNSAT TASK ELEMENT 5 STANDARD 10.2.2 Position Group Select switch to the group containing the CEA to be Group 6 selected.

tested.

Comment: Critical SAT / UNSAT TASK ELEMENT 6 STANDARD 10.2.3 Place Mode Select switch to MI. MI selected Comment: Critical SAT / UNSAT TASK ELEMENT 7 STANDARD 10.2.4 Verify MI light Illuminated. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 8 STANDARD 10.2.5 Verify white lights illuminated on Group Selection Matrix for the Verification complete.

group that contains the CEA to be moved.

Comment:

SAT / UNSAT Revision 1 Page 6 of 10 2011 NRC Exam

JPM S1 TASK ELEMENT 9 STANDARD 10.2.6 Verify white selection light Illuminated for the individual CEA to be Verification complete.

moved.

Comment:

SAT / UNSAT TASK ELEMENT 10 STANDARD NOTE Note reviewed.

Step 10.2.7 verifies CEA can be withdrawn prior to insertion.

Comment:

SAT / UNSAT TASK ELEMENT 11 STANDARD 10.2.7 If the selected CEA is inserted 2 steps, then withdraw the CEA to Not applicable.

the Upper Electrical Limit.

Comment:

SAT / UNSAT TASK ELEMENT 12 STANDARD 10.2.8 If the selected CEA is inserted < 2 steps, then perform the following:

CEA is inserted and 10.2.8.1 Insert selected CEA 2 steps from the Upper Electrical Limit. withdrawn.

10.2.8.2 Withdraw selected CEA to Upper Electrical Limit.

Comment: Critical SAT / UNSAT Revision 1 Page 7 of 10 2011 NRC Exam

JPM S1 TASK ELEMENT 13 STANDARD 10.2.9 Record Initial Position of selected CEA on Attachment 10.1. Position recorded.

Comment:

SAT / UNSAT TASK ELEMENT 14 STANDARD 10.2.10 Insert selected CEA at least 5 inches. CEA inserted > 5 inches.

Comment: Critical SAT / UNSAT Evaluator Note Malfunction will be inserted when CEA 20 is < 146 inches.

TASK ELEMENT 15 STANDARD Shim switch moved to Attempt to withdraw selected CEA to its park position.

withdraw.

Comment:

This will have no affect on CEA 20. It will continue to insert.

SAT / UNSAT TASK ELEMENT 16 STANDARD 10.2.14 Place Mode Select Switch to OFF. Switch is in OFF.

Comment: Critical This will not stop CEA 20 motion.

SAT / UNSAT Revision 1 Page 8 of 10 2011 NRC Exam

JPM S1 Evaluator Note When CEA 20 reaches 140 inches, the malfunction for dropping CEA 2 will be inserted. The applicant will have 2 CEAs misaligned by greater than 19 inches when CEA 20 reaches 131 inches. In accordance with OP-903-102, CEA or CEDMCS Malfunction, the applicant should trip the reactor.

TASK ELEMENT 17 STANDARD Reactor pushbuttons are pressed. Buttons manipulated.

Comment: Critical SAT / UNSAT END OF TASK Revision 1 Page 9 of 10 2011 NRC Exam

JPM S1 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-195 Verify the following Overrides assigned to Trigger 1:

Di-02a05a1s04-1 Di-02a05a1s07-1 Verify the following Event Triggers are assigned:

RDPCC30119 < 146 set on trigger 1 RDPCC30119 <140, then IMF RD02A08 assigned on trigger 3 Revision 1 Page 10 of 10 2011 NRC Exam

Surveillance Procedure OP-903-005 Control Element Assembly Operability Check Revision 11 3.0 PRECAUTIONS AND LIMITATIONS 3.1 PRECAUTIONS 3.1.1 Entry into Technical Specification 3.1.3.5 is required in Mode 1 for any shutdown CEA inserted below 145 inches withdrawn.

3.2 LIMITATIONS 3.2.1 Regulating Group 6 CEAs should not be inserted below 120 inches.

3.2.2 The time that any CEA is misaligned > 7 inches from any other CEA in its group shall be minimized. If a CEA cannot be restored to within 7 inches of its group, refer to Tech Spec 3.1.3.1.

5

Surveillance Procedure OP-903-005 Control Element Assembly Operability Check Revision 11 4.0 INITIAL CONDITIONS 4.1 Reactor is in Mode 1 or 2.

4.2 Departure from Nucleate Boiling Ratio (DNBR) within limits of Technical Specifications 3.2.4 and documented on Attachment 10.1, CEA Exercise Data Sheet.

4.3 Pre-shift briefing held, to include as a minimum:

Review of Technical Specifications 3.1.1.1, 3.1.3.1, 3.1.3.5, and 3.1.3.6 OP-901-102, CEA or CEDMCS Malfunction OP-903-090, Shutdown Margin 6

Surveillance Procedure OP-903-005 Control Element Assembly Operability Check Revision 11 5.0 MATERIAL AND TEST EQUIPMENT NONE 7

Surveillance Procedure OP-903-005 Control Element Assembly Operability Check Revision 11 6.0 ACCEPTANCE CRITERIA 6.1 Each CEA not fully inserted shall be determined to be Operable by movement of at least 5 inches in any direction.

8

Surveillance Procedure OP-903-005 Control Element Assembly Operability Check Revision 11 7.0 PROCEDURE 7.1 CEA OPERABILITY CHECK NOTE The following alarms will be initiated if a CEA deviates more than 5.5 inches inward or outward from any other CEA in its subgroup:

CEA Channel B Deviation (H-12, Cabinet K)

CEA Channel C Deviation (H-13, Cabinet K)

Prepower Dependent Insertion Limit (H-9, Cabinet H)

CAUTION (1) IF ANY CEA IS NOTED TO BE IMMOVABLE, MISALIGNED BY >19 INCHES, OR DROPS AT ANY TIME DURING PERFORMANCE OF THIS TEST, THEN GO TO OP-901-102, CEA OR CEDMCS MALFUNCTION.

(2) ENTRY INTO TECHNICAL SPECIFICATION 3.1.3.5 IS REQUIRED IN MODE 1 FOR ANY SHUTDOWN CEA INSERTED BELOW 145 INCHES WITHDRAWN.

(3) IF A REGULATING CEA OR A GROUP P CEA IS INSERTED OUT OF SEQUENCE

<140 INCHES, THEN A GROUP OUT OF SEQUENCE ANNUNCIATOR (A-7, CABINET L) AND A TARGETED CPC CHANNEL TRIP WILL OCCUR.

(4) THE TIME THAT ANY CEA IS MISALIGNED >7 INCHES FROM ANY OTHER CEA IN ITS GROUP SHALL BE MINIMIZED. IF A CEA CANNOT BE RESTORED TO WITHIN 7 INCHES OF ITS GROUP, THEN REFER TO TECH SPEC 3.1.3.1.

(5) WHILE CEAS MUST BE INSERTED TO <145 INCHES TO SATISFY THE REQUIRED 5 INCHES MOVEMENT, THE PERIOD OF TIME THIS DEVIATION EXISTS SHALL BE MINIMIZED. IF A CEA CANNOT BE RESTORED TO >145 INCHES, THEN REFER TO T.S. 3.1.3.1 AND T.S. 3.1.3.6.

7.1.1 Obtain Shift Manager/Control Room Supervisor permission to begin test and document on Attachment 10.1, CEA Exercise Data Sheet.

7.1.2 Verify Departure from Nucleate Boiling Ratio (DNBR) within limits of Technical Specification 3.2.4, and document on Attachment 10.1.

9

Surveillance Procedure OP-903-005 Control Element Assembly Operability Check Revision 11 7.1.3 For each CEA inserted 5 inches for Axial Shape Index control, perform the following on Attachment 10.1:

7.1.3.1 Record CEA position prior to ASI control as Initial Position.

7.1.3.2 Record present CEA Position as Test Position.

7.1.3.3 Document CEA movement 5 inches.

7.1.3.4 Record NA for Final Position.

7.1.4 On Attachment 10.1, check the CEAC that will be used for indication.

7.1.5 Perform Attachment 10.2 for each CEA inserted <5 inches for Axial Shape Index control:

10

10.1 CEA EXERCISE DATA SHEET Test Permission: CR Supervisor Today / 0700 (SM/CRS Signature) (Date/Time)

Step 7.1.2 DNBR within limits T.S. 3.2.4: JVS (Initial) 7.1.4 CEAC 1 or CEAC 2 (check one)

SHUTDOWN GROUP A SUB- MOVEMENT 5 CEA # INITIAL POSITION TEST POSITION FINAL POSITION GROUP INCHES (Y/N) 7 28 150 145 Y 150 30 150 144.5 Y 150 32 150 145 Y 150 34 150 144 Y 150 8 29 150 144.5 Y 150 31 150 145 Y 150 33 150 144 Y 150 35 150 144.5 Y 150 12 48 150 144 Y 150 50 150 145 Y 150 52 150 144 Y 150 54 150 144.5 Y 150 13 49 150 144.5 Y 150 51 150 145 Y 150 53 150 144 Y 150 55 150 144.5 Y 150 23 2 150 144 Y 150 3 150 145 Y 150 OP-903-005 Revision 11 Attachment 10.1 (1 of 5) 14

CEA EXERCISE DATA SHEET SHUTDOWN GROUP B SUB- MOVEMENT 5 CEA # INITIAL POSITION TEST POSITION FINAL POSITION GROUP INCHES (Y/N) 3 12 150 145 Y 150 14 150 144.5 Y 150 16 150 144 Y 150 18 150 145 Y 150 4 13 150 144 Y 150 15 150 144.5 Y 150 17 150 144 Y 150 19 150 145 Y 150 9 36 150 144 Y 150 38 150 144.5 Y 150 40 150 144.5 Y 150 42 150 145 Y 150 10 37 150 144 Y 150 39 150 144.5 Y 150 41 150 144 Y 150 43 150 145 Y 150 20 80 150 144.5 Y 150 82 150 144 Y 150 84 150 144.5 Y 150 86 150 145 Y 150 21 81 150 144 Y 150 83 150 145 Y 150 85 150 144.5 Y 150 87 150 145 Y 150 OP-903-005 Revision 11 Attachment 10.1 (2 of 5) 15

CEA EXERCISE DATA SHEET REGULATING GROUP 1 SUB- MOVEMENT 5 CEA # INITIAL POSITION TEST POSITION FINAL POSITION GROUP INCHES (Y/N) 1 4 150 145 Y 150 5 150 144.5 Y 150 6 150 145 Y 150 7 150 145 Y 150 18 69 150 144 Y 150 72 150 144 Y 150 75 150 145 Y 150 78 150 144.5 Y 150 REGULATING GROUP 2 SUB- MOVEMENT 5 FINAL POSITION CEA # INITIAL POSITION TEST POSITION GROUP INCHES (Y/N) 17 68 150 145 Y 150 71 150 144 Y 150 74 150 144.5 Y 150 77 150 145 Y 150 1 150 144 Y 150 19 70 150 144.5 Y 150 73 150 144.5 Y 150 76 150 145 Y 150 79 150 145 Y 150 REGULATING GROUP 3 SUB- MOVEMENT 5 FINAL POSITION CEA # INITIAL POSITION TEST POSITION GROUP INCHES (Y/N) 15 60 150 144 Y 150 62 150 145 Y 150 64 150 145 Y 150 66 150 144.5 Y 150 16 61 150 145 Y 150 63 150 144 Y 150 65 150 144 Y 150 67 150 145 Y 150 OP-903-005 Revision 11 Attachment 10.1 (3 of 5) 16

CEA EXERCISE DATA SHEET REGULATING GROUP 4 SUB- MOVEMENT 5 CEA # INITIAL POSITION TEST POSITION FINAL POSITION GROUP INCHES (Y/N) 2 8 150 145 Y 150 9 150 144 Y 150 10 150 144.5 Y 150 11 150 145 Y 150 14 56 150 144 Y 150 57 150 144 Y 150 58 150 145 Y 150 59 150 144 Y 150 REGULATING GROUP 5 SUB- MOVEMENT 5 FINAL POSITION CEA # INITIAL POSITION TEST POSITION GROUP INCHES (Y/N) 11 44 150 145 Y 150 45 150 145 Y 150 46 150 144.5 Y 150 47 150 145 Y 150 REGULATING GROUP 6 SUB- MOVEMENT 5 CEA # INITIAL POSITION TEST POSITION FINAL POSITION GROUP INCHES (Y/N) 5 20 21 22 23 GROUP P SUB- MOVEMENT 5 FINAL POSITION CEA # INITIAL POSITION TEST POSITION GROUP INCHES (Y/N) 6 24 25 26 27 OP-903-005 Revision 11 Attachment 10.1 (4 of 5) 17

CEA EXERCISE DATA SHEET ACCEPTANCE CRITERIA AND TEST ACCEPTANCE All CEAs not fully inserted have been demonstrated operable by movement of at least 5 inches in any direction.

REMARKS:

Performed by:

(Signature) (Date)

SM/CRS Review: /

(Signature) (Date/Time)

OP-903-005 Revision 11 Attachment 10.1 (5 of 5) 18

10.2 INDIVIDUAL CEA OPERABILITY CHECK 10.2.1 Position Individual CEA Selection switches to select desired CEA: ____________

(selected CEA) 10.2.2 Position Group Select switch to the group containing the CEA to be tested.

10.2.3 Place Mode Select switch to MI.

10.2.4 Verify MI light Illuminated.

10.2.5 Verify white lights illuminated on Group Selection Matrix for the group that contains the CEA to be moved.

10.2.6 Verify white selection light Illuminated for the individual CEA to be moved.

NOTE Step 10.2.7 verifies CEA can be withdrawn prior to insertion.

10.2.7 If the selected CEA is inserted 2 steps, then withdraw the CEA to the Upper Electrical Limit.

10.2.8 If the selected CEA is inserted < 2 steps, then perform the following:

10.2.8.1 Insert selected CEA 2 steps from the Upper Electrical Limit.

10.2.8.2 Withdraw selected CEA to Upper Electrical Limit.

10.2.9 Record Initial Position of selected CEA on Attachment 10.1.

10.2.10 Insert selected CEA at least 5 inches.

10.2.11 Record the Test Position for selected CEA on Attachment 10.1.

10.2.12 Document on Attachment 10.1 that selected CEA moved at least 5 inches.

10.2.13 Withdraw selected CEA to its park position in accordance with OP-010-004, Power Operations, Attachment 9.7, CEA Park Positions.

10.2.14 Place Mode Select Switch to OFF.

10.2.15 Record Final Position of selected CEA on Attachment 10.1.

10.2.16 If an additional CEA is to be tested, then go to step 10.2.1.

OP-903-005 Revision 11 [LAST PAGE] Attachment 10.2 (1 of 1) 19

10.2 INDIVIDUAL CEA OPERABILITY CHECK 10.2.1 Position Individual CEA Selection switches to select desired CEA: ____________

(selected CEA) 10.2.2 Position Group Select switch to the group containing the CEA to be tested.

10.2.3 Place Mode Select switch to MI.

10.2.4 Verify MI light Illuminated.

10.2.5 Verify white lights illuminated on Group Selection Matrix for the group that contains the CEA to be moved.

10.2.6 Verify white selection light Illuminated for the individual CEA to be moved.

NOTE Step 10.2.7 verifies CEA can be withdrawn prior to insertion.

10.2.7 If the selected CEA is inserted 2 steps, then withdraw the CEA to the Upper Electrical Limit.

10.2.8 If the selected CEA is inserted < 2 steps, then perform the following:

10.2.8.1 Insert selected CEA 2 steps from the Upper Electrical Limit.

10.2.8.2 Withdraw selected CEA to Upper Electrical Limit.

10.2.9 Record Initial Position of selected CEA on Attachment 10.1.

10.2.10 Insert selected CEA at least 5 inches.

10.2.11 Record the Test Position for selected CEA on Attachment 10.1.

10.2.12 Document on Attachment 10.1 that selected CEA moved at least 5 inches.

10.2.13 Withdraw selected CEA to its park position in accordance with OP-010-004, Power Operations, Attachment 9.7, CEA Park Positions.

10.2.14 Place Mode Select Switch to OFF.

10.2.15 Record Final Position of selected CEA on Attachment 10.1.

10.2.16 If an additional CEA is to be tested, then go to step 10.2.1.

OP-903-005 Revision 11 [LAST PAGE] Attachment 10.2 (1 of 1) 19

10.2 INDIVIDUAL CEA OPERABILITY CHECK 10.2.1 Position Individual CEA Selection switches to select desired CEA: ____________

(selected CEA) 10.2.2 Position Group Select switch to the group containing the CEA to be tested.

10.2.3 Place Mode Select switch to MI.

10.2.4 Verify MI light Illuminated.

10.2.5 Verify white lights illuminated on Group Selection Matrix for the group that contains the CEA to be moved.

10.2.6 Verify white selection light Illuminated for the individual CEA to be moved.

NOTE Step 10.2.7 verifies CEA can be withdrawn prior to insertion.

10.2.7 If the selected CEA is inserted 2 steps, then withdraw the CEA to the Upper Electrical Limit.

10.2.8 If the selected CEA is inserted < 2 steps, then perform the following:

10.2.8.1 Insert selected CEA 2 steps from the Upper Electrical Limit.

10.2.8.2 Withdraw selected CEA to Upper Electrical Limit.

10.2.9 Record Initial Position of selected CEA on Attachment 10.1.

10.2.10 Insert selected CEA at least 5 inches.

10.2.11 Record the Test Position for selected CEA on Attachment 10.1.

10.2.12 Document on Attachment 10.1 that selected CEA moved at least 5 inches.

10.2.13 Withdraw selected CEA to its park position in accordance with OP-010-004, Power Operations, Attachment 9.7, CEA Park Positions.

10.2.14 Place Mode Select Switch to OFF.

10.2.15 Record Final Position of selected CEA on Attachment 10.1.

10.2.16 If an additional CEA is to be tested, then go to step 10.2.1.

OP-903-005 Revision 11 [LAST PAGE] Attachment 10.2 (1 of 1) 19

10.2 INDIVIDUAL CEA OPERABILITY CHECK 10.2.1 Position Individual CEA Selection switches to select desired CEA: ____________

(selected CEA) 10.2.2 Position Group Select switch to the group containing the CEA to be tested.

10.2.3 Place Mode Select switch to MI.

10.2.4 Verify MI light Illuminated.

10.2.5 Verify white lights illuminated on Group Selection Matrix for the group that contains the CEA to be moved.

10.2.6 Verify white selection light Illuminated for the individual CEA to be moved.

NOTE Step 10.2.7 verifies CEA can be withdrawn prior to insertion.

10.2.7 If the selected CEA is inserted 2 steps, then withdraw the CEA to the Upper Electrical Limit.

10.2.8 If the selected CEA is inserted < 2 steps, then perform the following:

10.2.8.1 Insert selected CEA 2 steps from the Upper Electrical Limit.

10.2.8.2 Withdraw selected CEA to Upper Electrical Limit.

10.2.9 Record Initial Position of selected CEA on Attachment 10.1.

10.2.10 Insert selected CEA at least 5 inches.

10.2.11 Record the Test Position for selected CEA on Attachment 10.1.

10.2.12 Document on Attachment 10.1 that selected CEA moved at least 5 inches.

10.2.13 Withdraw selected CEA to its park position in accordance with OP-010-004, Power Operations, Attachment 9.7, CEA Park Positions.

10.2.14 Place Mode Select Switch to OFF.

10.2.15 Record Final Position of selected CEA on Attachment 10.1.

10.2.16 If an additional CEA is to be tested, then go to step 10.2.1.

OP-903-005 Revision 11 [LAST PAGE] Attachment 10.2 (1 of 1) 19

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S2 Makeup to the Volume Control Tank Applicant:

Examiner:

JPM S2 JOB PERFORMANCE MEASURE DATA PAGE Task: Makeup to the Volume Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control.

Task Standard: Applicant makes up to the Volume Control Tank in accordance with OP-002-005, Chemical and Volume Control. Applicant secures Boric Acid flow when the Boric Acid Batch Counter fails to secure flow. Applicant adds Primary Makeup to account for the additional Boric Acid.

References:

OP-002-005, Chemical and Volume Control Alternate Path: Yes Time Critical: No Validation Time: 30 mins.

K/A 004 A4.07 Boration / Dilution Importance Rating 3.9 / 3.7 RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 15 2011 NRC Exam

JPM S2 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-002-005, Chemical and Volume Control

==

Description:==

The applicant will be directed to makeup to the Volume Control Tank using Boric Acid and Primary Makeup Water batches. The applicant will be cued the quantity of Boric Acid and Primary Makeup to add. When the Boric Acid is added, the Boric Acid Batch Counter will fail to secure acid flow. The applicant will be required to manually secure Boric Acid flow. When the applicant reports the quantity of extra acid added to the Volume Control Tank, the CRS will direct him to add the required amount of Primary Makeup.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 15 2011 NRC Exam

JPM S2 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

Plant power is 100%

RCS boron concentration is 260 ppm Boric Acid Makeup Tank A boron concentration is 5850 ppm.

OP-002-005, Chemical and Volume Control, Attachment 11.8 has been completed and reviewed.

INITIATING CUES:

The CRS directs you to makeup to the Volume Control Tank using Boric Acid and Primary Makeup Water batches in accordance with OP-002-005, Chemical and Volume Control.

Add 9 gallons of Boric Acid and 200 gallons of Primary Makeup.

Revision 1 Page 4 of 15 2011 NRC Exam

JPM S2 Evaluator Note Cue the Simulator Operator to place the Simulator in RUN.

TASK ELEMENT 1 STANDARD Procedure Caution THIS SECTION AFFECTS REACTIVITY. THIS EVOLUTION SHOULD BE Caution reviewed.

CROSSCHECKED AND COMPLETED PRIOR TO LEAVING CP-4.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 6.11.1 Inform SM/CRS that this section is being performed. Notification made.

Comment:

SAT / UNSAT TASK ELEMENT 3 STANDARD Procedure Note (1) When performing a Plant down power where final RCS Boron Concentration needs to be determined, the following Plant Data Book figure(s) will assist the Operator in determining the required RCS Boron PPM change.

1.2.1.1 Power Defect Vs Power Level 1.4.3.1 Inverse Boron Worth Vs. Tmod at BOC (<30 EFPD)

Note reviewed.

1.4.4.1 Inverse Boron Worth Vs. Tmod at Peak Boron (30 EFPD up to 170 EFPD) 1.4.5.1 Inverse Boron Worth Vs. Tmod at MOC (170 EFPD up to 340 EFPD) 1.4.6.1 Inverse Boron Worth Vs. Tmod at EOC ( 340 EFPD)

(2) Batch makeup to the VCT should use 100 gallons of Primary Makeup Water to ensure all Boric Acid is added.

Comment:

SAT / UNSAT Revision 1 Page 5 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 4 STANDARD 6.11.2 At SM/CRS discretion, calculate the volume of Boric Acid to be added on Attachment 11.8, Calculation of Boric Acid Volume for VCT Provided in cue.

Acid/Water Batches.

Comment:

Add 9 gallons of Boric Acid and 200 gallons of Primary Makeup.

SAT / UNSAT TASK ELEMENT 5 STANDARD 6.11.3 Verify Direct Boration Valve, BAM-143, control switch in CLOSE. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 6 STANDARD 6.11.4 Set Boric Acid Makeup Batch Counter to volume of Boric Acid Counter set to 9 desired.

Comment: Critical SAT / UNSAT TASK ELEMENT 7 STANDARD 6.11.5 Verify Boric Acid Makeup Pumps selector switch aligned to desired BAM Tank A selected.

Boric Acid Makeup Pump A(B).

Comment:

SAT / UNSAT Revision 1 Page 6 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 8 STANDARD 6.11.6 Place Makeup Mode selector switch to BORATE. Switch is in Borate.

Comment: Critical SAT / UNSAT TASK ELEMENT 9 STANDARD 6.11.7 Verify selected Boric Acid Makeup Pump A(B) Starts. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 10 STANDARD 6.11.8 Open VCT Makeup Valve, CVC-510. CVC-510 is open.

Comment: Critical SAT / UNSAT TASK ELEMENT 11 STANDARD Procedure Note The Boric Acid Flow Totalizer will not register below 3 GPM. The Boric Note reviewed.

Acid Flow Totalizer is most accurate in the range of 10 - 25 GPM.

Comment:

SAT / UNSAT Revision 1 Page 7 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 12 STANDARD 6.11.9 If manual control of Boric Acid flow is desired, then perform the following:

6.11.9.1 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual. Flow is established > 3 gpm 6.11.9.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, output to >3 GPM flow rate.

Comment: Critical Manual mode should be used.

SAT / UNSAT Only manual control or auto control is Critical, not both.

TASK ELEMENT 13 STANDARD 6.11.10 If automatic control of Boric Acid flow is desired, then perform the following:

6.11.10.1 Place Boric Acid Flow controller, BAM-IFIC-0210Y, in Auto. Flow is established > 3 gpm 6.11.10.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, setpoint potentiometer to >3 GPM flow rate.

Comment: Critical It is not likely that the applicant will use this mode, but it is allowed.

SAT / UNSAT Only manual control or auto control is Critical, not both.

TASK ELEMENT 14 STANDARD 6.11.11 Verify Boric Acid Makeup Control Valve, BAM-141, Intermediate Verification complete.

or Open.

Comment:

SAT / UNSAT Revision 1 Page 8 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 15 STANDARD 6.11.12 Observe Boric Acid flow rate for proper indication. Verification complete.

Comment:

SAT / UNSAT Evaluator Note The Alternate Path becomes applicable when the Boric Acid Batch Counter reaches zero. It will count to zero and then start counting in negative numbers for this failure.

TASK ELEMENT 16 STANDARD 6.11.13 When Boric Acid Makeup Batch Counter has counted down to Applicant stops Boric Acid desired value, then verify Boric Acid Makeup Control Valve, BAM-141, flow.

Closed.

Comment: Critical The applicant could secure Boric Acid flow by going to minimum on the Boric Acid controller or by closing CVC-510. SAT / UNSAT Evaluator Note After the applicant secures Boric Acid flow, cue him to report how much extra Boric Acid was added.

Based on this report, inform him that the STA has performed the necessary calculations and direct the applicant to add Primary Makeup as follows:

12 total gallon of Boric Acid, then add 250 gallons of Primary Makeup.

15 total gallon of Boric Acid, then add 320 gallons of Primary Makeup.

20 total gallon of Boric Acid, then add 430 gallons of Primary Makeup.

TASK ELEMENT 17 STANDARD 6.11.14 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual. Verification complete.

Comment:

SAT / UNSAT Revision 1 Page 9 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 18 STANDARD 6.11.15 Verify both Boric Acid Flow controller, BAM-IFIC-0210Y, output Output set to zero and setpoint potentiometer set to zero.

Comment: Critical SAT / UNSAT TASK ELEMENT 19 STANDARD 6.11.16 Place Makeup Mode selector switch to MANUAL. Switch is in Manual.

Comment: Critical SAT / UNSAT TASK ELEMENT 20 STANDARD 6.11.17 Verify selected Boric Acid Makeup Pump A(B) Stops. Verification complete.

Comment:

SAT / UNSAT Evaluator Note After the applicant secures Boric Acid flow, cue him to report how much extra Boric Acid was added.

Based on this report, inform him that the STA has performed the necessary calculations and direct the applicant to add Primary Makeup as follows:

12 total gallon of Boric Acid, then add 250 gallons of Primary Makeup.

15 total gallon of Boric Acid, then add 320 gallons of Primary Makeup.

20 total gallon of Boric Acid, then add 430 gallons of Primary Makeup.

Revision 1 Page 10 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 21 STANDARD 6.11.18 Set Primary Makeup Water Batch Counter to volume of Primary Counter set to prompted Makeup water desired. amount.

Comment: Critical The PMU counter reads in values of 10. 250 is set as 25, 320 is set as 32, and 430 is set as 43. SAT / UNSAT TASK ELEMENT 22 STANDARD 6.11.19 Place Makeup Mode selector switch to DILUTE. Switch is in Dilute.

Comment: Critical SAT / UNSAT TASK ELEMENT 23 STANDARD Procedure Note The Dilution Flow Totalizer will not register below 5 GPM. The Dilution Note reviewed.

Flow Totalizer is most accurate at >10 GPM.

Comment:

SAT / UNSAT TASK ELEMENT 24 STANDARD Procedure Caution DILUTION SHALL IMMEDIATELY BE STOPPED IF PRE-POWER Caution reviewed.

DEPENDENT INSERTION LIMIT (H-9, CABINET H) ALARM IS INITIATED OR ANY UNEXPECTED REACTIVITY CHANGE OCCURS.

Comment:

SAT / UNSAT Revision 1 Page 11 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 25 STANDARD 6.11.20 If manual control of Primary Makeup Water flow is desired, then perform the following:

6.11.20.1 Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, PMU flow is established > 5 in Manual. gpm.

6.11.20.2 Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, output to >5 GPM flow rate.

Comment: Critical Only manual control or auto control is Critical, not both.

SAT / UNSAT TASK ELEMENT 26 STANDARD 6.11.21 If automatic control of Primary Makeup Water flow is desired, then perform the following:

6.11.21.1 Place Primary Makeup Water Flow controller, PMU-IFIC-0210X, PMU flow is established > 5 in Auto. gpm.

6.11.21.2 Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, setpoint potentiometer to >5 GPM flow rate.

Comment: Critical Only manual control or auto control is Critical, not both.

SAT / UNSAT TASK ELEMENT 27 STANDARD 6.11.22 Verify Primary Makeup Water Control Valve, PMU-144, Verification complete.

Intermediate or Open.

Comment:

SAT / UNSAT Revision 1 Page 12 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 28 STANDARD 6.11.23 Observe Primary Makeup water flow rate for proper indication. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 29 STANDARD 6.11.24 When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary Makeup Water Control Valve, PMU- Verification complete.

144, Closed.

Comment:

SAT / UNSAT TASK ELEMENT 30 STANDARD 6.11.25 Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, Verification complete.

in Manual.

Comment:

SAT / UNSAT TASK ELEMENT 31 STANDARD 6.11.26 Verify both Primary Makeup Water Flow controller, PMU-IFIC-Output set to 0%

0210X, output and setpoint potentiometer set to zero.

Comment: Critical SAT / UNSAT Revision 1 Page 13 of 15 2011 NRC Exam

JPM S2 TASK ELEMENT 32 STANDARD 6.11.27 Place Makeup Mode selector switch to MANUAL. Switch is in Manual.

Comment: Critical SAT / UNSAT TASK ELEMENT 33 STANDARD 6.11.28 Repeat Steps 6.11.4 through 6.11.27 to obtain desired VCT level. N/A Comment:

SAT / UNSAT TASK ELEMENT 34 STANDARD 6.11.29 When desired VCT level is achieved, then Close VCT Makeup CVC-510 is closed.

Valve, CVC-510.

Comment: Critical SAT / UNSAT END OF TASK Revision 1 Page 14 of 15 2011 NRC Exam

JPM S2 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-195 Verify the following Malfunctions:

o CV35b No Triggers are required for this JPM.

Revision 1 Page 15 of 15 2011 NRC Exam

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 3.0 PRECAUTIONS AND LIMITATIONS 3.1 PRECAUTIONS 3.1.1 Concentrated Boric Acid in unheated lines may precipitate if lines cool sufficiently.

3.1.2 RCS activity changes may occur when RCS chemical composition is altered (i.e.

Crud Burst).

3.1.3 The reactivity effects of boration or dilution should be observed in terms of:

Resulting CEA motion if CEDMCS is in Automatic Sequential mode Changes in source count rate Changes in Reactor Coolant Temperature Changes in Reactor Power The following backup and redundant nuclear instrumentation and other reactor and plant indications may be used by Operators when making reactivity changes:

PARAMETER INDICATORS RC-ITI-0102CA, RC-ITI-0102CB, TCold RC-ITI-0102CC, RC-ITI-0102CD ENI-IJR-0002A, ENI-IJR-0002B, Linear Power ENI-IJR-0002C, ENI-IJR-0002D Core T Power (BDELT) C24104 ENI-IJR-0001, ENI-IJI-0001A, ENI-IJI-0001B, Log power ENI-IJI-0001C, ENI-IJI-0001D Startup Countrate ENI-IJR-0005, ENI-IJI-0005, ENI-IJI-0006 Startup Rate Startup rate meter on CP-2 Steam Bypass MS-IHI-C0319A RC-IPI-0101A, RC-IPI-0101B, RC-IPI-0101C, Pressurizer Pressure (NR)

RC-IPI-0101D RC-IPI-0102A, RC-IPI-0102B, RC-IPI-0102C, Pressurizer Pressure (WR)

RC-IPI-0102D PHICAL CPC PID-171, Channels A, B, C, D BDT CPC PID-177, Channels A, B, C, D 3.1.4 Planned power changes are authorized by the Shift Manager. The necessary reactivity manipulations are then:

Preceded by a brief Ordered by the SM/CRS Directly supervised by an SRO.

3.1.5 Reactivity manipulations to maintain power level are:

6

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 Discussed by the ATC / BOP and SM/CRS (pertaining to expected power and temperature response)

Ordered by the SM/CRS.

3.1.6 If any of the following occur, then prompt corrective action should be taken:

Pre-Power Dependent Insertion Limit (H-9, Cabinet H) alarms CEAs or RCS temperature change in an unexpected manner High Linear Power Pre-Trip B/D (C-14, Cabinet K) alarms High Linear Power Pre-Trip A/C (B-14, Cabinet K) alarms 3.1.7 To minimize thermal transients in the system, Letdown and Charging flows should be started and stopped simultaneously.

3.1.8 When batching, adding chemicals, or filling an Ion Exchanger with resin, care should be taken to prevent foreign material from contaminating the Batching Tank, Chemical Addition Tank, Resin Addition Tank, Resin Slurry, or Granular Boric Acid.

3.1.9 An explosive mixture of hydrogen and air in the VCT shall be avoided at all times.

The oxygen concentration should be maintained below 2% by volume as verified by VCT gas space samples.

3.1.10 Letdown temperatures downstream of the Letdown Heat Exchanger should be maintained 120 F to prevent degradation of the Ion Exchanger resin.

3.1.11 RCS boron concentration, RCS temperature, Reactor Power, and RCS Chemistry should be closely monitored when placing an Ion Exchanger in service.

3.1.12 Charging Pump pulsation dampeners should be pressurized using expected RCS pressure in accordance with OP-904-007, Charging Pump Pulsation Dampener Pressure Check.

3.1.13 To minimize thermal transients, Letdown flow should not be throttled below 10 GPM.

3.1.14 Prior to adding water, borating, or blending to the RWSP, the RWSP vent line should be drained of accumulated water in order to prevent excessive pressure buildup in the RWSP during fill and the possibility of rupturing the RWSP plug seal.

3.1.15 To minimize thermal shock of charging nozzles, only one Charging Pump should be started at a time.

3.1.16 When filling Ion Exchangers, Chemistry should ensure the correct type resin is used for the desired operation (purifying, delithiating or deborating).

7

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 3.1.17 Reactivity should be monitored for changes when Letdown temperature changes.

Lowering letdown temperature may remove boron from Ion Exchanger effluent and raising letdown temperature may release boron into Ion Exchanger effluent.

3.1.18 During periods of maximum Letdown flow, Letdown header pressure may increase sufficiently to cause Letdown Header Relief Valve, CVC-126, to lift.

3.1.19 Any evolution at CP-4 affecting reactivity should be cross-checked and completed prior to leaving CP-4.

3.1.20 Delay in starting a Charging Pump following venting increases the potential for gas coming out of solution in the water used to fill the pump and to form air pockets inside the cylinder block. The venting should be performed as soon as practical prior to starting the pump. If the Charging Pump is not started within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following completion of venting, then the venting should be performed again.

[CR-WF3-2009-02782 CA-37, CR-WF3-2010-1829]

3.2 LIMITATIONS 3.2.1 If a change in Reactor Coolant Boron concentration of greater than or equal to 50 PPM occurs or is anticipated, then Boron Equalization should be performed in accordance with the applicable General Operating Procedure:

OP-010-003, Plant Startup (Attachment 9.20)

OP-010-004, Power Operations (Attachment 9.3)

OP-010-005, Plant Shutdown (Attachment 9.12) 3.2.2 Chemistry and Radiation Protection Departments should be notified prior to initiating CVC operations that could impact chemical or radiological controls.

These include but are not limited to:

VCT purging and venting Starting or stopping Charging Pumps Placing Purification Ion Exchangers in or out of service Purification Ion Exchangers resin replacement 3.2.3 During Plant Mode changes, plant configuration shall meet the requirements of Technical Specification 3.1.2.9 prior to racking in Charging Pump breakers.

3.2.4 To ensure proper chemical mixing, with the RCS filled, at least one Reactor Coolant Pump in each loop should be operating prior to performing any boration or dilution operation.

3.2.5 Proper flow through the Charging Pump Seal System will extend Charging Pump packing life. The Charging Pump Seal System is not required for Charging Pump Operability.

8

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 3.2.6 The CVC Purification Filter will degrade when subjected to temperatures above 180°F for several hours. If Letdown Heat Exchanger Outlet Temperature reaches 140°F and can not be corrected within two hours, then the CVC Purification Filter should be removed from service and Chemistry notified.

3.2.7 To minimize the effects of RCP seal perturbations, maintain Reactor Coolant Pump Control Bleedoff Pressure 40 PSIG to 65 PSIG during normal operations.

Reactor Coolant Pump Control Bleedoff Pressure operating band may be expanded to 30 PSIG to 120 PSIG during startup and shutdowns. In addition, the rate of backpressure change should not exceed 4 PSIG/minute. Refer to Attachment 11.19 for RCP Operating limits.

3.2.8 In Modes 1 - 4, Penetration #27 (CVC) is subject to thermally induced overpressurization during a Design Basis Accident. Overpressurization relief is administratively provided by maintaining CVC-220, Charging To RC Loop 1A Bypass Isolation, Open. [GL 96-06, CR-WF3-2004-00480, ER-W3-2004-0143-000]

3.2.9 Do not exceed 95% indicated level in either BAM Tank A or B. [EC-I92-001]

3.2.10 The Charging Pump seal packing life will be extended by flushing the Charging Pump packing cooling water system periodically (monthly or more frequently) to remove debris and fines inside the system. The packing seal water system should also be flushed just before every manual start of the Charging Pump for non-emergency use, as is directed in Step 6.1.1 or 6.2.1. The instructions in the watchstation logs may also be used for flushing the system. [CR-WF3-2009-02782 CA-10]

9

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 6.11 VCT MAKEUP USING ACID/W ATER BATCHES (C)

[P-22047]

CAUTION THIS SECTION AFFECTS REACTIVITY. THIS EVOLUTION SHOULD BE CROSS-CHECKED AND COMPLETED PRIOR TO LEAVING CP-4.

6.11.1 Inform SM/CRS that this section is being performed.

NOTE (1) When performing a Plant down power where final RCS Boron Concentration needs to be determined, the following Plant Data Book figure(s) will assist the Operator in determining the required RCS Boron PPM change.

1.2.1.1 Power Defect Vs Power Level 1.4.3.1 Inverse Boron Worth Vs. Tmod at BOC (<30 EFPD) 1.4.4.1 Inverse Boron Worth Vs. Tmod at Peak Boron (30 EFPD up to 170 EFPD) 1.4.5.1 Inverse Boron Worth Vs. Tmod at MOC (170 EFPD up to 340 EFPD) 1.4.6.1 Inverse Boron Worth Vs. Tmod at EOC ( 340 EFPD)

(2) Batch makeup to the VCT should use 100 gallons of Primary Makeup Water to ensure all Boric Acid is added.

6.11.2 At SM/CRS discretion, calculate the volume of Boric Acid to be added on Attachment 11.8, Calculation of Boric Acid Volume for VCT Acid/Water Batches.

6.11.3 Verify Direct Boration Valve, BAM-143, control switch in CLOSE.

6.11.4 Set Boric Acid Makeup Batch Counter to volume of Boric Acid desired.

6.11.5 Verify Boric Acid Makeup Pumps selector switch aligned to desired Boric Acid Makeup Pump A(B).

6.11.6 Place Makeup Mode selector switch to BORATE 6.11.7 Verify selected Boric Acid Makeup Pump A(B) Starts.

6.11.8 Open VCT Makeup Valve, CVC-510.

46

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 NOTE The Boric Acid Flow Totalizer will not register below 3 GPM. The Boric Acid Flow Totalizer is most accurate in the range of 10 - 25 GPM.

6.11.9 If manual control of Boric Acid flow is desired, then perform the following:

6.11.9.1 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual.

6.11.9.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, output to >3 GPM flow rate.

6.11.10 If automatic control of Boric Acid flow is desired, then perform the following:

6.11.10.1 Place Boric Acid Flow controller, BAM-IFIC-0210Y, in Auto.

6.11.10.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, setpoint potentiometer to

>3 GPM flow rate.

6.11.11 Verify Boric Acid Makeup Control Valve, BAM-141, Intermediate or Open.

6.11.12 Observe Boric Acid flow rate for proper indication.

6.11.13 When Boric Acid Makeup Batch Counter has counted down to desired value, then verify Boric Acid Makeup Control Valve, BAM-141, Closed.

6.11.14 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual.

6.11.15 Verify both Boric Acid Flow controller, BAM-IFIC-0210Y, output and setpoint potentiometer set to zero.

6.11.16 Place Makeup Mode selector switch to MANUAL.

6.11.17 Verify selected Boric Acid Makeup Pump A(B) Stops.

6.11.18 Set Primary Makeup Water Batch Counter to volume of Primary Makeup water desired.

6.11.19 Place Makeup Mode selector switch to DILUTE.

47

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 035 NOTE The Dilution Flow Totalizer will not register below 5 GPM. The Dilution Flow Totalizer is most accurate at >10 GPM.

CAUTION DILUTION SHALL IMMEDIATELY BE STOPPED IF PRE-POWER DEPENDENT INSERTION LIMIT (H-9, CABINET H) ALARM IS INITIATED OR ANY UNEXPECTED REACTIVITY CHANGE OCCURS.

6.11.20 If manual control of Primary Makeup Water flow is desired, then perform the following:

6.11.20.1 Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Manual.

6.11.20.2 Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, output to

>5 GPM flow rate.

6.11.21 If automatic control of Primary Makeup Water flow is desired, then perform the following:

6.11.21.1 Place Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Auto.

6.11.21.2 Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, setpoint potentiometer to >5 GPM flow rate.

6.11.22 Verify Primary Makeup Water Control Valve, PMU-144, Intermediate or Open.

6.11.23 Observe Primary Makeup water flow rate for proper indication.

6.11.24 When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary Makeup Water Control Valve, PMU-144, Closed.

6.11.25 Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Manual.

6.11.26 Verify both Primary Makeup Water Flow controller, PMU-IFIC-0210X, output and setpoint potentiometer set to zero.

6.11.27 Place Makeup Mode selector switch to MANUAL.

6.11.28 Repeat Steps 6.11.4 through 6.11.27 to obtain desired VCT level.

6.11.29 When desired VCT level is achieved, then Close VCT Makeup Valve, CVC-510.

48

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S3 Secure Shutdown Cooling Train B Applicant:

Examiner:

JPM S3 JOB PERFORMANCE MEASURE DATA PAGE Task: Secure Shutdown Cooling Train B Task Standard: Applicant secures Shutdown Cooling Train B in accordance with OP-009-005, Shutdown Cooling.

References:

OP-009-005, Shutdown Cooling Alternate Path: No Time Critical: No Validation Time: 15 mins.

K/A 005 A4.01 Controls and indications for RHR Importance Rating 3.6 / 3.4 pumps RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 9 2011 NRC Exam

JPM S3 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-009-005, Shutdown Cooling

==

Description:==

Shutdown Cooling Trains A and B will be running in this IC. The applicant will be directed to secure Shutdown Cooling Train B. All manipulations occur at CP-8.

There are no malfunctions associated with this JPM.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 9 2011 NRC Exam

JPM S3 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

Shutdown Cooling Trains A and B are operating.

Reactor Coolant Pumps 1B and 2B are operating.

TCOLD is 240 °F and stable Shutdown Cooling Train B is scheduled to be tagged out next shift.

INITIATING CUES:

The CRS directs you to secure Shutdown Cooling Train B in accordance with OP-009-005, Shutdown Cooling.

Revision 1 Page 4 of 9 2011 NRC Exam

JPM S3 Evaluator Note Cue the Simulator Operator to ensure the Simulator in RUN.

TASK ELEMENT 1 STANDARD Procedure Note (1) This section provides instructions to restore Shutdown cooling Train B to Standby status. Note reviewed.

(2) Refer to Technical Specifications 3.4.1.3, 3.4.1.4, 3.4.1.5, 3.9.8.1, 3.9.8.2 prior to removal of a Shutdown Cooling Train.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 6.4.1 Set Computer Point PID K43201, SDCS/LPSI PMP B LOW FLOW Provide cue.

LIMIT, to Zero in accordance with OP-004-012, Plant Computer System.

Comment:

Cue the applicant that another operator will update the PMC.

SAT / UNSAT TASK ELEMENT 3 STANDARD 6.4.2 Verify Shutdown HX B CCW Flow Control, CC-963B, control switch Switch placed in at SETPOINT. SETPOINT.

Comment: Critical SAT / UNSAT Revision 1 Page 5 of 9 2011 NRC Exam

JPM S3 TASK ELEMENT 4 STANDARD 6.4.3 Place LPSI Header Flow Controller 1A/1B, SI-IFIC-0306, to Manual, Controller is in manual with and adjust output to 10%. 0% output.

Comment: Critical SAT / UNSAT TASK ELEMENT 5 STANDARD 6.4.4 Open RC Loop 1 Shdn Cooling Warmup, SI-135B. SI-135 B is open.

Comment: Critical SAT / UNSAT TASK ELEMENT 6 STANDARD 6.4.5 Close the following valves:

SI-139 B & Si-138 B are SI-139B LPSI Header to RC Loop 1A Flow Control closed.

SI-138B LPSI Header to RC Loop 1B Flow Control Comment: Critical SAT / UNSAT Revision 1 Page 6 of 9 2011 NRC Exam

JPM S3 TASK ELEMENT 7 STANDARD 6.4.6 If Shutdown Cooling Train B temperature, as indicated by LPSI Pump B Discharge Header Temperature Indicator, SI-ITI-0352X, is Step reviewed.

200°F, then allow Shutdown Cooling B temperature to drop to <200°F prior to continuing with this Section.

Comment:

Temperature should be < 200 °F when this step is reached.

SAT / UNSAT TASK ELEMENT 8 STANDARD 6.4.7 Stop LPSI Pump B by performing the following:

LPSI Pump B is secured 6.4.7.1 Place LPSI Pump B control switch to OFF.

and the control switch is in 6.4.7.2 When LPSI Pump B has Stopped, then place LPSI Pump B NORMAL.

control switch to NORMAL.

Comment: Critical SAT / UNSAT TASK ELEMENT 9 STANDARD 6.4.8 Close Shutdown Cooling HX B Temperature Control, SI-415B. SI-415 B is closed.

Comment: Critical SAT / UNSAT TASK ELEMENT 10 STANDARD 6.4.9 Place Shutdown HX B CCW Flow Control, CC-963B, control switch CC-963 B is closed.

to CLOSE.

Comment: Critical SAT / UNSAT Revision 1 Page 7 of 9 2011 NRC Exam

JPM S3 TASK ELEMENT 11 STANDARD 6.4.10 Close and Lock RC Loop 1 SDC Suction Outside Containment Isol, SI-407 B is locked closed.

SI-407B.

Comment: Critical SAT / UNSAT TASK ELEMENT 12 STANDARD 6.4.11 Notify Radiation Protection Department that Shutdown Cooling Notification made.

Train B has been secured.

Comment:

SAT / UNSAT TASK ELEMENT 13 STANDARD 6.4.12 Complete Attachment 11.4, Restoration Lineup for SDC Train B, Step reviewed.

for applicable locked valves operated.

Comment:

End task when this step is reached.

SAT / UNSAT END OF TASK Revision 1 Page 8 of 9 2011 NRC Exam

JPM S3 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-198 There are no malfunctions or overrides associated with this JPM.

Revision 1 Page 9 of 9 2011 NRC Exam

System Operating Procedure OP-009-005 Shutdown Cooling Revision 026 3.0 PRECAUTIONS AND LIMITATIONS 3.1 PRECAUTIONS 3.1.1 The maximum temperature for Purification Ion Exchanger(s) is 140 F.

3.1.2 If Letdown to Ion Exchangers Inlet/Bypass, CVC-140, is in AUTO then Purification Ion Exchangers will automatically bypass at 140 F.

3.1.3 The following applies to Shutdown Cooling flow:

3.1.3.1 A total minimum Shutdown Cooling flow necessary to remove decay heat and prevent boron stratification should be maintained at all times.

3.1.3.2 When considering the minimum Shutdown Cooling flow required to adequately remove decay heat and prevent boron stratification, then the flow from the operating Shutdown Cooling train or the combined flow of both operating Shutdown Cooling trains may be used.

3.1.3.3 The required minimum Shutdown Cooling flow for Modes 5 and 6 are as follows:

TIME AFTER SHUTDOWN (HOURS) REQUIRED FLOW (GPM) 0 - <175 hours 4000 GPM 175 - <375 hours 3000 GPM 375 hours0.00434 days 0.104 hours 6.200397e-4 weeks 1.426875e-4 months 2000 GPM If the Reactor has been shutdown <175 hours, then Shutdown Cooling flow may be reduced to 3000 GPM, if RCS temperature is verified to be

<135 F at least once per hour.

3.1.3.4 Changes to the Shutdown Cooling flow rate will cause the Alternate Shutdown Cooling Purification flow rate to change with SI-424 throttled or full open. For example, with Alternate Shutdown Cooling Purification in service via SI-424, if Shutdown Cooling flow rate is lowered, then the Alternate Shutdown Cooling Purification flow rate will increase due to increased LPSI Discharge Header pressure. This change in flow rate may require readjustment of SI-424 so the 250 GPM limit is not exceeded.

3.1.4 LPSI pumps shall not be run for >3 hours in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period on recirculation flow only.

3.1.5 If LPSI Pump flow is limited to >100 gpm and <2000 gpm for 3 consecutive hours, then contact the System Engineer for guidance on LPSI Pump monitoring.

[CR-W3-2000-1376]

5

System Operating Procedure OP-009-005 Shutdown Cooling Revision 026 3.2 LIMITATIONS 3.2.1 Shutdown Cooling shall not be initiated until Reactor Coolant System (RCS) temperature <350 F and RCS Pressure <392 PSIA. [P-4055]

3.2.2 RCS temperature changes shall be limited by the following:

A maximum heatup rate of 60 F per hour A maximum cooldown rate of 100 F per hour 3.2.3 RCS temperature and pressure shall be limited in accordance with the limit lines shown on Technical Specification Figures 3.4-2 and 3.4-3 with instrument uncertainty incorporated for pressure and temperature as follows:

[ER-W3-2004-00439]

Subtract 30 F from indicated temperature Add 35 PSI to the indicated pressure from the following instruments:

CP-2 RC IPI0103 (100-750 PSIA)

RC IPI0104 (100-750 PSIA)

CP-4 RC IPI0105 (100-750 PSIA)

RC IPI0106 (100-750 PSIA)

CP-7 RC IPI0101A(B,C,D) (1500-2500 PSIA)

LCP-43 RC IPI0105-1 (100-750 PSIA)

RC IPI0106-1 (100-750 PSIA)

Add 110 PSI to the indicated pressure from the following instruments:

CP-2 RC IPI0102A3 (B3) (0-3000 PSIA)

CP-4 RC IPI0102A2 (B2) (0-3000 PSIA)

CP-7 RC IPI0102A (B,C,D) (0-3000 PSIA)

LCP-43 RC IPI0102A1 (B1,C1,D1) (0-3000 PSIA) 3.2.4 Maximum flow through a Purification Ion Exchanger is 126 GPM.

6

System Operating Procedure OP-009-005 Shutdown Cooling Revision 026 3.2.5 When RCS is in Mode 4 and any RCS Cold Leg temperature is <230 F or in Mode 5 or Mode 6 with Reactor Vessel Head on, then Low Temperature Overpressure Protection (Tech. Spec. 3.4.8.3) shall be provided by one of the following: [P-5804]

Both Shutdown Cooling Suction Header Relief Valves aligned to RCS or RCS depressurized with an RCS vent 5.6 in2 3.2.6 In Mode 4 with RCS pressure >400 PSIA, Both Containment Spray Trains shall be operable in accordance with Tech Spec 3.6.2.1.

3.2.7 The Shutdown Cooling Train placed in service should be on the Protected Train.

3.2.8 Scaffolding will be required to Vent CS Header A, when restoring CS to operation after securing Shutdown Cooling.

3.2.9 Shutdown Cooling requires one Operable Low Pressure Safety Injection Flow Control Valve per train for Shutdown Cooling to be Operable.

3.2.10 The Shutdown Cooling suction line piping upstream of SI-407A(B) should not be filled from an external water source due to the potential to cause thermal binding of SI-405A(B). There is a possibility of air intrusion into and voiding of the Shutdown Cooling suction line piping upstream of SI-407A(B) following a period of operation in Modes 1 - 4, when the Safety Injection and Containment Spray Systems are aligned for the normal injection mode. This condition has been evaluated and is described in the Design Basis Document for the Safety Injection System (W3-DBD-001). The Safety Injection System retains the capability of performing its design safety function with the described voiding and no actions are required to fill the subject section of piping.

[ER-W3-2002-0283-002, CR-WF3-2004-01300, ECM03-003, W3-DBD-01]

3.2.11 Each pump start stresses motor windings both thermally and mechanically. A start means motor comes up to rated speed. Starts for LPSI pumps should be limited as follows:

3.2.11.1 LPSI Pump A:

3.2.11.1.1 With motor at ambient temperature, do not attempt more than 6 consecutive starts.

3.2.11.1.2 With motor at operating temperature, do not attempt more than 4 consecutive starts.

3.2.11.1.3 Allowed time between additional starts is 15 minutes with motor at operating temperature or 30 minutes with motor at ambient temperature.

7

System Operating Procedure OP-009-005 Shutdown Cooling Revision 026 3.2.11.2 LPSI Pump B:

3.2.11.2.1 With motor at ambient temperature, do not attempt more than 2 consecutive starts.

3.2.11.2.2 With motor at operating temperature, do not attempt more than 1 consecutive starts.

3.2.11.2.3 Allowed time between additional starts is 15 minutes with motor at operating temperature or 45 minutes with motor at ambient temperature.

3.2.12 Following a Design Basis Tornado Event, delaying the initiation of Shutdown Cooling (SDC) for up to 7 days will be required to ensure the Component Cooling Water system is capable of removing Reactor Coolant System decay heat. The actual delay time will depend on UHS damage and ambient temperature and will be determined by Engineering. Emergency Feedwater supports decay heat removal until SDC can be initiated. [EC-530]

3.2.13 To assure that the RCS does not drop to the minimum RCS bolt up Temperature indicated on TS figures 3.4-2 and 3.4-3, the limits on the following instruments apply:

Instrument Description Min Value Location SI IT7114 / SI IT7115 RWSP Temperature 67.3°F PMC SI IT0351 X / SI IT0352 X LPSI Pump Outlet 66.09°F QSPDS Temperature CS IT0303 X AND Y Shutdown Cooling Outlet 66.09°F QSPDS Temperature 3.2.14 When Alternate Shutdown Cooling Purification is aligned with parallel Purification Ion Exchangers and SI-424 is throttled or full open, then:

A maximum Total Shutdown Cooling flow of 250 GPM should not be exceeded.

A minimum Total Shutdown Cooling flow of 50 GPM should be minimized.

If Total Shutdown Cooling flow is to be <50 GPM, then single Purification Ion Exchanger operation is required.

PMC PID S39205, SDC/LD Purification Full Range Flow is the only available indication when Letdown flow is >150 GPM.

8

System Operating Procedure OP-009-005 Shutdown Cooling Revision 026 6.4 SECURING SHUTDOWN COOLING TRAIN B NOTE (1) This section provides instructions to restore Shutdown cooling Train B to Standby status.

(2) Refer to Technical Specifications 3.4.1.3, 3.4.1.4, 3.4.1.5, 3.9.8.1, 3.9.8.2 prior to removal of a Shutdown Cooling Train.

6.4.1 Set Computer Point PID K43201, SDCS/LPSI PMP B LOW FLOW LIMIT, to Zero in accordance with OP-004-012, Plant Computer System.

6.4.2 Verify Shutdown HX B CCW Flow Control, CC-963B, control switch at SETPOINT.

6.4.3 Place LPSI Header Flow Controller 1A/1B, SI-IFIC-0306, to Manual, and adjust output to 10%.

6.4.4 Open RC Loop 1 Shdn Cooling Warmup, SI-135B.

6.4.5 Close the following valves:

SI-139B LPSI Header to RC Loop 1A Flow Control SI-138B LPSI Header to RC Loop 1B Flow Control 6.4.6 If Shutdown Cooling Train B temperature, as indicated by LPSI Pump B Discharge Header Temperature Indicator, SI-ITI-0352X, is 200 F, then allow Shutdown Cooling B temperature to drop to <200 F prior to continuing with this Section.

6.4.7 Stop LPSI Pump B by performing the following:

6.4.7.1 Place LPSI Pump B control switch to OFF.

6.4.7.2 When LPSI Pump B has Stopped, then place LPSI Pump B control switch to NORMAL.

6.4.8 Close Shutdown Cooling HX B Temperature Control, SI-415B.

6.4.9 Place Shutdown HX B CCW Flow Control, CC-963B, control switch to CLOSE.

6.4.10 Close and Lock RC Loop 1 SDC Suction Outside Containment Isol, SI-407B.

6.4.11 Notify Radiation Protection Department that Shutdown Cooling Train B has been secured.

6.4.12 Complete Attachment 11.4, Restoration Lineup for SDC Train B, for applicable locked valves operated.

31

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S4 Balance of Plant Operator Immediate Operator Actions on Control Room Evacuation Applicant:

Examiner:

JPM S4 JOB PERFORMANCE MEASURE DATA PAGE Task: Perform balance of plant operators immediate operator action on Control Room evacuation with fire conditions.

Task Standard: Applicant performed immediate operator actions for BOP position for a fire in the Control Room in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

Applicant manually closed Atmospheric Dump Valve #2 due to setpoint failure.

References:

OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown Alternate Path: Yes Time Critical: No Validation Time: 5 mins.

K/A 039 A4.01 Main Steam Supply Valves Importance Rating 2.9 / 2.8 RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 7 2011 NRC Exam

JPM S4 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

None

==

Description:==

The applicant will be cued that there is a fire in CP-33. The CRS will direct him to carry out his immediate operator actions as BOP operator. ADV #2 setpoint will fail to 666 PSIG, spuriously opening ADV #2. The applicant will be required to place that controller to manual and set it to 0% output to close ADV #2. The task will end when the applicant goes to the key locker.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 7 2011 NRC Exam

JPM S4 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is at 100% power A fire has started in CP-33 The CRS has entered OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown The ATC is performing his immediate operator actions.

INITIATING CUES:

The CRS directs you to perform the BOP immediate operator actions.

Revision 1 Page 4 of 7 2011 NRC Exam

JPM S4 Evaluator Note Cue the Simulator Operator to place the Simulator in RUN and initiate Trigger 1.

TASK ELEMENT 1 STANDARD 2.1 Verify Turbine trip:

Governor valves Closed Verification complete.

Throttle valves Closed Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 2.2 Verify Generator trip:

Exciter Field Breaker Tripped Verification complete.

Generator Breaker A Tripped Generator Breaker B Tripped Comment:

SAT / UNSAT TASK ELEMENT 3 STANDARD 2.3 RESET Moisture Separator Reheater controls. Reset was depressed.

Comment: Critical SAT / UNSAT Revision 1 Page 5 of 7 2011 NRC Exam

JPM S4 2.4 IF evacuating the Control Room due to fire, THEN perform the following:

Evaluator Note The Alternate Path becomes applicable when the applicant addresses the Atmospheric Dump Valves. ADV #2 will have spuriously opened.

TASK ELEMENT 4 STANDARD 2.4.1 IF EITHER of the following valves has spuriously Opened, THEN place the applicable controller(s) in MANUAL AND lower the output to ADV #2 controller is placed zero:

in manual and output MS-116A SG 1 Atmospheric Dump lowered to 0%.

MS-116B SG 2 Atmospheric Dump Comment: Critical ADV #2 setpoint fails, driving ADV #2 open with normal pressure in S/G

2. SAT / UNSAT TASK ELEMENT 5 STANDARD 2.4.2 Close the following valves:

MS-124 A & MS-124 B are MS-124A Main Steam Isol Valve #1 closed.

MS-124B Main Steam Isol Valve #2 Comment: Critical SAT / UNSAT TASK ELEMENT 6 STANDARD 2.5 Obtain Operations Security Key Ring AND proceed to RAB +35 Relay Keys obtained.

Room.

Comment:

SAT / UNSAT END OF TASK Revision 1 Page 6 of 7 2011 NRC Exam

JPM S4 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-196 Verify the following Malfunctions:

o MS23b on Trigger 1 with a 5 second delay Verify the following Overrides:

o Di-02a02s02-0 on Trigger 1 o Di-02a02s01-0 on Trigger 1 Coordinate with examiner to initiate Trigger 1 on his cue. This will trip the reactor and insert the ADV #2 malfunction.

Revision 1 Page 7 of 7 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S5 Start Hydrogen Recombiner A Applicant:

Examiner:

JPM S5 JOB PERFORMANCE MEASURE DATA PAGE Task: Start Hydrogen Recombiner A Task Standard: Hydrogen Recombiner A is operating in accordance with OP-008-006, Hydrogen Recombiner.

References:

OP-008-006, Hydrogen Recombiner.

Alternate Path: No Time Critical: No Validation Time: 20 mins.

K/A 028 A4.01 HRPS controls Importance Rating 4.0 / 4.0 RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 11 2011 NRC Exam

JPM S5 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-008-006, Hydrogen Recombiner

==

Description:==

Applicant will be directed to start Hydrogen Recombiner A using OP-008-006.

All manipulations will occur at CP-33. One reading will be required to be obtained from CP-8. Step 6.1.10 will direct monitoring for temperature to rise to > 1225 °F. The task will be stopped when this step is reached.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 11 2011 NRC Exam

JPM S5 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

A Loss of Coolant Accident is in progress.

Before the LOCA, Containment Temperature was recorded as 105 °F in OP-903-001, Technical Specification Logs.

INITIATING CUES:

The CRS directs you to place Hydrogen Recombiner A in service using OP-008-006, Hydrogen Recombiner.

Revision 1 Page 4 of 11 2011 NRC Exam

JPM S5 Evaluator Note Cue the Simulator Operator to place the Simulator in RUN.

TASK ELEMENT 1 STANDARD 6.1.1.1 Record present Post-LOCA Containment pressure from Containment Atmosphere Pressure Indicator, ESF-IPI-6750A, on Value recorded. 1.2, Hydrogen Recombiner Power Control Setting Data Sheet.

Comment:

This value can vary based on the time taken to complete JPM S2. Value should be 18.7 - 18.0 psia. SAT / UNSAT TASK ELEMENT 2 STANDARD 6.1.1.2 Record Pre-LOCA Containment Average Temperature from OP-Value recorded.

903-001, Technical Specification Surveillance Logs, on Attachment 11.2.

Comment:

Given in the cue.

SAT / UNSAT TASK ELEMENT 3 STANDARD 6.1.1.3 Determine Pressure Factor (Cp) from Attachment 11.4, Dry Correction Factor Containment Recombiner Power Correction Factor Graph. determined.

Comment: Critical Will vary based on pressure recorded in step 1. Should determine a value of 1.150 to 1.195. SAT / UNSAT TASK ELEMENT 4 STANDARD 6.1.1.3.1 Record Pressure Factor (Cp) on Attachment 11.2. Value recorded.

Comment:

SAT / UNSAT Revision 1 Page 5 of 11 2011 NRC Exam

JPM S5 TASK ELEMENT 5 STANDARD 6.1.1.4 Determine Hydrogen Recombiner Power Control Setting by Setting determined.

multiplying a reference power of 48 KW by Cp.

Comment: Critical Equals value determined in step 6.1.1.3

48. Should be 55.2 - 57.36.

SAT / UNSAT TASK ELEMENT 6 STANDARD 6.1.1.4.1 Record Hydrogen Recombiner Power Control Setting on Value recorded. 1.2.

Comment:

SAT / UNSAT TASK ELEMENT 7 STANDARD 6.1.2 Continuously monitor the Hydrogen Recombiner A Power Meter, Step reviewed.

HRA-EM-960, when power level is being changed.

Comment:

SAT / UNSAT TASK ELEMENT 8 STANDARD Procedure Note Adjusting the Hydrogen Recombiner Power Control potentiometer slowly Note reviewed.

compensates for the lag between the meter and the potentiometer adjustments.

Comment:

SAT / UNSAT Revision 1 Page 6 of 11 2011 NRC Exam

JPM S5 TASK ELEMENT 9 STANDARD 6.1.3 Verify Hydrogen Recombiner A Power Control potentiometer is set Setting verified.

at zero (000).

Comment:

SAT / UNSAT TASK ELEMENT 10 STANDARD 6.1.4 Place Hydrogen Recombiner A Power control switch, HRA-001A, to HRA-001A is ON.

ON.

Comment: Critical SAT / UNSAT TASK ELEMENT 11 STANDARD 6.1.5 Slowly adjust Hydrogen Recombiner Power Control potentiometer 5 KW is indicated on for Hydrogen Recombiner A until 5 KW is indicated on Hydrogen Hydrogen Recombiner A Recombiner A Power Meter, HRA-EM-960. Power Meter.

Comment: Critical SAT / UNSAT TASK ELEMENT 12 STANDARD 6.1.5.1 Hold reading for 10 minutes. Acknowledges hold.

Comment:

Prompt applicant that 10 minutes has elapsed.

SAT / UNSAT Revision 1 Page 7 of 11 2011 NRC Exam

JPM S5 TASK ELEMENT 13 STANDARD 6.1.6 Verify Hydrogen Thermocouple Temperatures trend upward when adjusting Power Control Potentiometer, as indicated on Hydrogen Temperatures monitored.

Recombiner A Outlet Air Temperature Indicator, HRA-ITI-0001A. Use Temperature Select switch to read thermocouple temperatures.

Comment:

SAT / UNSAT TASK ELEMENT 14 STANDARD 6.1.7 Adjust Hydrogen Recombiner Power Control potentiometer for 10 KW is indicated on Hydrogen Recombiner A until 10 KW indicated on Hydrogen Recombiner Hydrogen Recombiner A A Power Meter, HRA-EM-960. Power Meter.

Comment: Critical SAT / UNSAT Upward temperature trend should be monitored by applicant, but temperature change will be very slow.

TASK ELEMENT 15 STANDARD 6.1.7.1 Hold reading for 10 minutes. Acknowledges hold.

Comment:

Prompt applicant that 10 minutes has elapsed.

SAT / UNSAT TASK ELEMENT 16 STANDARD 6.1.8 Adjust Hydrogen Recombiner Power Control potentiometer for 20 KW is indicated on Hydrogen Recombiner A until 20 KW indicated on Hydrogen Recombiner Hydrogen Recombiner A A Power Meter, HRA-EM-960. Power Meter.

Comment: Critical SAT / UNSAT Upward temperature trend should be monitored by applicant, but temperature change will be very slow.

Revision 1 Page 8 of 11 2011 NRC Exam

JPM S5 TASK ELEMENT 17 STANDARD 6.1.8.1 Hold reading for 10 minutes. Acknowledges hold.

Comment:

Prompt applicant that 10 minutes has elapsed.

SAT / UNSAT TASK ELEMENT 18 STANDARD Procedure Caution Caution Reviewed.

DO NOT EXCEED 75 KW.

Comment:

SAT / UNSAT TASK ELEMENT 19 STANDARD 6.1.9 Adjust Hydrogen Recombiner Power Control Potentiometer for Potentiometer adjusted to Hydrogen Recombiner A to setting calculated on Attachment 11.2. 56.6 - 57.4.

Comment: Critical SAT / UNSAT TASK ELEMENT 20 STANDARD Procedure Caution Caution reviewed.

DO NOT EXCEED 1400 °F.

Comment:

SAT / UNSAT Revision 1 Page 9 of 11 2011 NRC Exam

JPM S5 TASK ELEMENT 21 STANDARD 6.1.10 Adjust Hydrogen Recombiner Power Control potentiometer as necessary, within the following guidelines, to maintain heater temperature

> 1225 °F to 1400 °F, as read on Hydrogen Recombiner A Outlet Air Temperature Indicator, HRA-ITI-0001A:

Use the average of all three thermocouples temperatures to obtain a heater temperature. Example: 1200, 1210, and 1220, use 1210°F.

If only two thermocouples are within 50°F of each other, then use average of the closest two temperatures. Examples: 1200, 1210, and Step reviewed.

1270, use 1205°F.

The following computer points can be used to trend operation of the Hydrogen Recombiner Operation:

A42700 - Temp 1 A42701 - Temp 2 A42702 - Temp 3 Comment:

Temperature will be < 300 °F at this time. End task when applicant reads all 3 thermocouple temperatures. SAT / UNSAT END OF TASK Revision 1 Page 10 of 11 2011 NRC Exam

JPM S5 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-199 Verify the following Malfunctions:

o Rc24a 3%

Place the Simulator in Run on the lead examiners cue.

Revision 1 Page 11 of 11 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S6 Start and Load Emergency Diesel Generator A Applicant:

Examiner:

JPM S6 JOB PERFORMANCE MEASURE DATA PAGE Task: Start and Load Emergency Diesel Generator A Task Standard: Applicant starts and commences loading EDG A in accordance with OP-009-002. Applicant must trip EDG A when load starts rising without manipulation.

References:

OP-009-002, Emergency Diesel Generator OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification Alternate Path: Yes Time Critical: No Validation Time: 20 mins.

K/A 064 A4.06 Manual start, loading, and Importance Rating 3.9 / 3.9 stopping of the ED/G RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 16 2011 NRC Exam

JPM S6 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-009-002, Emergency Diesel Generator OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification

==

Description:==

The applicant will be directed to start Emergency Diesel Generator A in accordance with OP-903-068 and OP-009-002. When the EDG A load is raised to 1 MW, load will begin rising on its own. The applicant must trip the EDG from CP-1. No action is necessary by the simulator booth operator.

Prompts will be required to inform the applicant that the EDG A pre-start checks have been completed and that other operators will be gathering start time information for the EDG Start Evaluation.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 16 2011 NRC Exam

JPM S6 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The shift is scheduled to perform a remote start of Emergency Diesel Generator A in accordance with OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification OP-903-068 Section 7.1 is complete through step 7.1.4.

This is not for ESFAS testing.

All pre-startup checks have been completed and the RAB Watch is standing by.

2 additional operators are standing by with stop watches to time the start of EDG A. They will time the start and fill out Attachment 11.8 Emergency Diesel Generator Start Evaluation.

INITIATING CUES:

The CRS directs you to perform a manual start of Emergency Diesel Generator A for OP-903-068 using OP-009-002.

Revision 1 Page 4 of 16 2011 NRC Exam

JPM S6 Evaluator Note The applicant will should start the task in OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification.

TASK ELEMENT 1 STANDARD Procedure Note:

(1) Timing shall be started when the Emergency Diesel Generator is started. Timing shall be stopped when the EDG voltage reaches 3920 Reviews Notes VAC and frequency reaches 58.8 Hz.

(2) A Chart Recorder may be substituted, under an approved WA, for recording EDG start time(s) at the discretion of the SM/CRS.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 7.1.5 Perform a start of the Emergency Diesel Generator using one of the following methods: Refers to OP-009-002 Section Manual, in accordance with OP-009-002, Emergency Diesel 6.3 Generator.

Comment:

SAT / UNSAT Evaluator Note The applicant will should transition to OP-009-002, Emergency Diesel Generator.

TASK ELEMENT 3 STANDARD Verifies RTGB light is lit for 6.3.1 Verify RTGB light Illuminated on CP-1.

EDG A.

Comment:

SAT / UNSAT Revision 1 Page 5 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 4 STANDARD 6.3.2 Dispatch an Operator to the Diesel Room to observe Start Goes to next step sequence.

Comment:

SAT / UNSAT Evaluator Cue: An operator is standing by in EDG Room A.

TASK ELEMENT 5 STANDARD 6.3.3 Perform Emergency Diesel Generator pre-startup checks in accordance with Subsection 6.1, Emergency Diesel Generator Pre- Reviews step Startup Checks.

Comment:

Evaluator Cue: Pre-startup Checks have been completed SAT / UNSAT TASK ELEMENT 6 STANDARD Procedure Note:

(1) The Starting sequence is listed in Attachment 11.6, Starting Sequence, for reference.

(2) Receipt of Starting Air System Malfunction alarm during start sequence may be indication of failed starting air solenoid or starting air Note reviewed control valves.

(3) During engine operation, oil or water may be observed dripping from EG A(B) Left and Right Intercooler Drains, EGA-201A(B) and EGA-202A(B).

Comment:

SAT / UNSAT Revision 1 Page 6 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 7 STANDARD Procedure Caution (1) MINIMIZE THE TIME THE DIESEL GENERATOR IS OPERATED AT UNLOADED OR LOW LOAD CONDITIONS.

(2) BOTH EMERGENCY DIESEL GENERATOR'S SHALL NOT BE OPERATED IN THE TEST MODE SIMULTANEOUSLY, EXCEPT WHEN PERFORMING TESTING PURSUANT TO TECHNICAL Caution reviewed SPECIFICATION SURVEILLANCE REQUIREMENT 4.8.1.1.2.G.

(3) STEP 6.3.4 REQUIRES THE USE OF TWO STOPWATCHES AND OPERATORS FOR TWO DIFFERENT TIME REQUIREMENTS. M&TE DATES ARE RECORDED ON ATTACHMENT11.9, DIESEL GENERATOR RUNNING LOG.

Comment:

Evaluator Cue: Additional operators are in position and standing by to SAT / UNSAT perform timing duties and complete start evaluation attachment and diesel running log.

TASK ELEMENT 8 STANDARD 6.3.4 Simultaneously position the Engine Control Switch to Start on Positions switch to START CP-1, and start timing:

Comment: Critical Timing not critical, performed by other operators. SAT / UNSAT TASK ELEMENT 9 STANDARD 6.3.5 Verify Diesel Generator A Exh Fan, HVR-0025A, has Started and Verifies fan starts and damper EG A Room Outside Air Intake Damper, HVR-501A, Open. Refer to opens computer mimic HVR3 for damper position indication.

Comment:

SAT / UNSAT TASK ELEMENT 10 STANDARD 6.3.6 Verify Emergency Diesel Generator parameters are within acceptable ranges 15 minutes after start and every 30 minutes Refers to OP-903-068 thereafter in accordance with Attachment 11.9, Diesel Generator Running Log.

Comment:

Evaluator Cue: Additional operators are standing by to record SAT / UNSAT readings and monitor diesel parameters.

Revision 1 Page 7 of 16 2011 NRC Exam

JPM S6 Evaluator Note The applicant will should transition back to OP-903-068, Emergency Diesel Generator and Subgroup Relay Operability Verification.

TASK ELEMENT 11 STANDARD 7.1.6 Verify that the Emergency Diesel Generator steady state voltage and frequency are maintained between 3920 to 4580 VAC and 58.8 to Verification complete.

61.2 Hz respectively.

Comment:

SAT / UNSAT TASK ELEMENT 12 STANDARD 7.1.7 Record start times on Attachment 10.1. Performed by other operators.

Comment:

Timing not critical, performed by other operators. SAT / UNSAT TASK ELEMENT 13 STANDARD 7.1.8 Operate Emergency Diesel Generator unloaded for 5 minutes. N/A Comment:

Cue the applicant that 5 minutes has passed.

SAT / UNSAT TASK ELEMENT 14 STANDARD Procedure Caution:

DO NOT EXCEED 4.84 MW FOR MORE THAN TWO HOURS OUT Caution reviewed.

OF ANY 24 HOUR PERIOD.

Comment:

SAT / UNSAT Revision 1 Page 8 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 15 STANDARD 7.1.9 Synchronize the Emergency Diesel Generator to Offsite Power and load to > 1.0 MW and < 1.2 MW, in accordance with OP-009-002, Refers to OP-009-002 Emergency Diesel Generator.

7.1.9.1 Maintain this load for 5 minutes.

Comment:

SAT / UNSAT Evaluator Note The applicant will should transition to OP-009-002, Emergency Diesel Generator, section 6.4.

TASK ELEMENT 16 STANDARD Procedure Note:

(1) Diesel Generator load changes can be accomplished by performing the following:

Manual voltage control, when in parallel, will raise or lower reactive load.

Manual voltage control, when not in parallel, will raise or lower generator voltage.

Note reviewed.

While in parallel engine speed control is used to raise or lower generator load.

(2) The operations necessary to synchronize the Diesel Generator either from the Control Room or locally are identical. The point of control is determined by whether the Control mode is selected for Local or RTGB (Control Room). Switch positions for the local control panel are in parentheses.

Comment:

SAT / UNSAT Revision 1 Page 9 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 17 STANDARD Procedure Caution:

WHENEVER POSSIBLE THE EMERGENCY DIESEL GENERATOR SHOULD BE OPERATED FOR 5 MINUTES PRIOR TO LOADING.

Caution reviewed.

THIS WILL HELP TO MINIMIZE THERMAL STRESS ON THE ENGINE TO ENSURE OPTIMUM ENGINE LIFE AND PERFORMANCE.

Comment:

SAT / UNSAT TASK ELEMENT 18 STANDARD 6.4.1 Verify Emergency Diesel Generator operating with voltage 3920 -

Verification complete.

4580 VAC and frequency 58.8 - 61.2 Hz.

Comment:

SAT / UNSAT TASK ELEMENT 19 STANDARD 6.4.2 Verify Volt Regulator Mode Select (Sevr Manual/Auto) Switch is Verifies switch position in Auto.

Comment:

SAT / UNSAT TASK ELEMENT 20 STANDARD Procedure Caution:

RELAY DAMAGE MAY RESULT IF SYNCHRONIZER IS ENERGIZED Reviews note FOR LONGER THAN 5 MINUTES Comment:

SAT / UNSAT Revision 1 Page 10 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 21 STANDARD Procedure Warning EMERGENCY DIESEL GENERATOR B SHOULD NOT BE OPERATED IN PARALLEL WITH THE MAIN GENERATOR WHEN MAIN GENERATOR VOLTAGE IS >25.95 KV AS INDICATED BY PID A58003. REACTIVE LOAD (MVAR) MAY BE LOWERED TO REDUCE Note Reviewed.

MAIN GENERATOR VOLTAGE. OPERATING EDG B IN PARALLEL WITH THE MAIN GENERATOR WHEN MAIN GENERATOR VOLTAGE IS >25.95 KV HAS THE POTENTIAL TO CAUSE THE 3B32 BUS BREAKERS, UPON A FAULT, TO STRUCTURALLY DECOMPOSE AND EXPLODE.

Comment:

Note refers to Emergency Diesel Generator B, EDG A is being started.

SAT / UNSAT TASK ELEMENT 22 STANDARD 6.4.3 Position the Emergency Diesel Generator A Synchronizer Switch Positions Switch to MAN (Man/Off/Auto Synch Switch) to Gen Man (Man).

Comment: Critical SAT / UNSAT TASK ELEMENT 23 STANDARD Raises and lowers EDG A 6.4.4 Verify proper voltage response using the Volt Adjust (Sevr voltage using Volt Adjust switch Potentiometer Adjust), then adjust Emergency Diesel Generator A Verifies EDG A Voltage slightly voltage to slightly higher than system voltage. higher than bus voltage and between 3920-4580 V Comment:

SAT / UNSAT Revision 1 Page 11 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 24 STANDARD 6.4.5 Verify proper frequency response using the Speed Adjust Raises and lowers EDG A (Engine Speed Adjustment), then adjust engine speed until the Speed and verifies EDG A synchroscope is rotating slowly in the clockwise direction. frequency responds Comment:

SAT / UNSAT TASK ELEMENT 25 STANDARD Procedure Note:

If the Red Start light is out, then the Emergency Diesel Generator control circuit may not be lined up to automatically shift to the Test Note reviewed.

Mode of operation when the Emergency Diesel Generator output breaker is Closed. This may make the Emergency Diesel Generator trip when the Emergency Diesel Generator output breaker is closed.

Comment:

SAT / UNSAT TASK ELEMENT 26 STANDARD Verifies light is lit on EDG A 6.4.6 Verify Emergency Diesel Generator A Red Start Light Illuminated.

Start Switch.

Comment:

SAT / UNSAT TASK ELEMENT 27 STANDARD Procedure Note:

Do not simultaneously connect both Emergency Diesel Generator A(B)

Note reviewed.

to their respective busses during non-emergency conditions or with offsite power available.

Comment:

SAT / UNSAT Revision 1 Page 12 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 28 STANDARD Procedure Caution:

WHEN EMERGENCY DIESEL GENERATOR IS CONNECTED TO THE GRID, MAINTAIN OUTGOING REACTIVE LOAD (MVAR) AND Caution reviewed.

AT LEAST 0.1 MW REAL LOAD TO PREVENT A REVERSE POWER TRIP.

Comment:

SAT / UNSAT TASK ELEMENT 29 STANDARD 6.4.7 Observing Synchroscope rotating slowly in the clockwise direction, Close the Diesel Generator output breaker at the 5 minutes Breaker Closed to twelve position on the synchroscope.

Comment: Critical SAT / UNSAT TASK ELEMENT 30 STANDARD 6.4.8 Immediately apply a small load, approximately 0.1 MW, to the EDG does not trip on Reverse Emergency Diesel Generator A using the Speed Adjust (Engine Speed Power Adjustment) Control Switch.

Comment:

SAT / UNSAT TASK ELEMENT 31 STANDARD 6.4.9 Position the Emergency Diesel Generator A Synchronizer Switch Switch placed in OFF (Man/Off/Auto Synch Switch) to Off.

Comment: Critical Total time energized should be < 5 minutes SAT / UNSAT Revision 1 Page 13 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 32 STANDARD Procedure Caution:

WHILE ADJUSTING MVAR DO NOT EXCEED BUS VOLTAGE OF Caution reviewed.

4470 VAC.

Comment:

SAT / UNSAT TASK ELEMENT 33 STANDARD Adjusts MVAR to obtain ~ 1 6.4.10 Adjust the Volt Adjust to obtain 1 MVAR.

MVAR out Comment:

SAT / UNSAT Evaluator Note The Alternate Path is inserted at this point. When the applicant raises load, the EDG load will continue rising after the Speed Adjust switch is released.

TASK ELEMENT 34 STANDARD Adjusts EDG A Load to between 1.0 and 1.2 MWe using Speed Adjust Releases switch when load is switch as needed per step 7.1.9 of OP-902-068. in required range.

Comment: Critical SAT / UNSAT Revision 1 Page 14 of 16 2011 NRC Exam

JPM S6 TASK ELEMENT 35 STANDARD Notes that EDG load does not stop increasing when Speed Adjust switch is released and secures EDG A by any of the following:

DIESEL GEN A TRIP pushbutton on CP-1 Trips EDG A or Opens Output Breaker prior to exceeding 4.84 Opens EDG A Output Breaker at CP-1 MWe.

Takes the EDG A control switch to stop Directs the local NAO to pull the EDG overspeed.

Comment: Critical SAT / UNSAT END OF TASK Revision 1 Page 15 of 16 2011 NRC Exam

JPM S6 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC - 195 Verify the following using the Event button:

o ZAOEGEM2328CS > 0.1 set to initiate Trigger 5 Verify the following Overrides:

o Di-01a07s02-1 set to raise on Trigger 5.

Revision 1 Page 16 of 16 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S7 Perform Controlled Ventilation Area System Operability Check Applicant:

Examiner:

JPM S7 JOB PERFORMANCE MEASURE DATA PAGE Task: Perform OP-903-052, Controlled Ventilation Area System Operability Check Task Standard: Applicant secured RAB Normal Ventilation, started CVAS Train A, and restarted RAB Normal Ventilation in accordance with OP-903-052, Controlled Ventilation Area System Operability Check.

References:

OP-903-052, Controlled Ventilation Area System Operability Check Alternate Path: No Time Critical: No Validation Time: 15 mins.

K/A 029 K1.03 Engineering safeguards Importance Rating 3.6 / 3.8 RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 12 2011 NRC Exam

JPM S7 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-052, Controlled Ventilation Area System Operability Check

==

Description:==

Applicant will be directed to perform OP-903-052, OP-903-052, Controlled Ventilation Area System Operability Check, on Train A. The applicant will secure RAB Normal Ventilation, start CVAS Train A, and re-start RAB Normal Ventilation. All manipulations will occur on CP-18.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 12 2011 NRC Exam

JPM S7 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

RAB Normal Ventilation Train B is in operation.

No welding, burning, painting, or chemical cleaning activities have been performed within the CVAS boundary or RAB +46 HVAC room for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

There is no paint or chemical cleaning solvents that are not completely dry.

INITIATING CUES:

The CRS directs you to perform OP-903-052, Controlled Ventilation Area System Operability Check, for CVAS Train A.

Revision 1 Page 4 of 12 2011 NRC Exam

JPM S7 Evaluator Note Cue the Simulator Operator to place the Simulator in RUN.

TASK ELEMENT 1 STANDARD 7.1.1 Obtain permission to perform this test from SM/CRS and record Provided.

permission on Attachment 10.1.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 7.1.2 Indicate train to be tested on Attachment 10.1. Data recorded.

Comment:

SAT / UNSAT TASK ELEMENT 3 STANDARD Procedure Caution DO NOT OPERATE CVAS FILTER UNITS FOR NON-EMERGENCY PURPOSES IF WELDING, BURNING, PAINTING, OR CHEMICAL CLEANING ACTIVITIES ARE BEING PERFORMED WITHIN THE CVAS Caution reviewed.

BOUNDARY OR RAB +46 HVAC ROOM. ENSURE ANY PAINTING OR CHEMICAL CLEANING IS COMPLETELY DRY BEFORE RUNNNG THE CVAS UNIT.

Comment:

SAT / UNSAT This is provided on the cue sheet.

TASK ELEMENT 4 STANDARD 7.1.3 Verify Containment Pressure Control is secured in accordance with Verification complete.

OP-008-002, Containment Atmosphere Release.

Comment:

SAT / UNSAT Revision 1 Page 5 of 12 2011 NRC Exam

JPM S7 TASK ELEMENT 5 STANDARD 7.1.4 Notify Radiation Protection that RAB Normal Ventilation will be Notification made.

secured for approximately 10 minutes.

Comment:

SAT / UNSAT TASK ELEMENT 6 STANDARD 7.1.5 Notify Chemistry to suspend Fume Hood operation, prior to securing Notification made.

RAB Normal Ventilation (approximate duration of 10 minutes).

Comment:

SAT / UNSAT TASK ELEMENT 7 STANDARD 7.1.6 Notify Security Department that RAB Normal Ventilation will be Notification made.

secured for approximately 10 minutes.

Comment:

SAT / UNSAT TASK ELEMENT 8 STANDARD CAUTION DO NOT SECURE RAB NORMAL VENTILATION IF CONTAINMENT PRESSURE REDUCTION IS IN PROGRESS, TO PREVENT AN Caution reviewed.

UNMONITORED RELEASE FROM CONTAINMENT INTO THE +46 HVAC ROOM AS WELL AS THE +46 HVAC PENETRATION AREA.

Comment:

SAT / UNSAT Revision 1 Page 6 of 12 2011 NRC Exam

JPM S7 TASK ELEMENT 9 STANDARD 7.1.7 Make an announcement on the Plant Page stating No welding, burning, chemical cleaning or painting in the RAB or RCA while running Page made.

CVAS Train A(B).

Comment:

SAT / UNSAT TASK ELEMENT 10 STANDARD WARNING FOR PERSONEL SAFETY, THE RAB EXHAUST FAN, HVR0009 A(B), Warning reviewed.

SHALL BE SECURED IF THE RAB SUPPLY FAN HVR0002 A(B), IS SECURED.

Comment:

SAT / UNSAT TASK ELEMENT 11 STANDARD 7.1.8 Secure the RAB Normal Supply Fan, HVR0002 B. Fan secured.

Comment: Critical SAT / UNSAT TASK ELEMENT 12 STANDARD 7.1.9 Secure the RAB Normal Exhaust Fan, HVR0009 B. Fan secured Comment: Critical SAT / UNSAT Revision 1 Page 7 of 12 2011 NRC Exam

JPM S7 TASK ELEMENT 13 STANDARD NOTE HVAC pig B may give erroneous readings and alarms when CVAS Note reviewed.

filtration units are operating.

Comment:

SAT / UNSAT TASK ELEMENT 14 STANDARD 7.1.10 Start Controlled Vent Area Exh Fan A(B), HVR0021A B. Fan started.

Comment: Critical SAT / UNSAT TASK ELEMENT 15 STANDARD 7.1.10.1 Verify the following:

HVR-104 (D52434) ....................Closed

HVR-105 (D52432) ....................Closed

HVR-106 (D52438) ................... Closed

HVR-107 (D52439) ................... Closed

HVR-108 (D52430) ................... Closed Verification complete.

HVR-109 (D52428) ................... Closed

HVR-110 (D52424) ................... Closed

HVR-111 (D52426) ................... Closed

HVR-301 (D52435) ....................Open

HVR-302 (D52421) ....................Open Comment:

SAT / UNSAT Revision 1 Page 8 of 12 2011 NRC Exam

JPM S7 TASK ELEMENT 16 STANDARD 7.1.10.2 Record the time and date on Attachment 10.1. Data recorded.

Comment:

SAT / UNSAT TASK ELEMENT 17 STANDARD NOTE Controlled Vent Area Exh Fan A(B) DP is measured for trending Note reviewed.

purposes.

Comment:

SAT / UNSAT TASK ELEMENT 18 STANDARD 7.1.11 When the system is stable, then record DP on Attachment 10.1. Data recorded.

Comment:

SAT / UNSAT TASK ELEMENT 19 STANDARD NOTE When RAB Normal Exhaust Fans, HVR0009A(B), are started, HVAC B Note reviewed.

Radiation Monitor may alarm depending on gas levels in the RAB.

Comment:

SAT / UNSAT Revision 1 Page 9 of 12 2011 NRC Exam

JPM S7 TASK ELEMENT 20 STANDARD 7.1.12 Start RAB Normal Exhaust Fan, HVR0009 B. Fan started.

Comment: Critical SAT / UNSAT TASK ELEMENT 21 STANDARD 7.1.12.1 Verify the following on the operating fan:

HVR-122B (D52027) ..........................Open Verification complete.

HVR-121B (D52024) ..........................Not Closed Comment:

SAT / UNSAT TASK ELEMENT 22 STANDARD 7.1.12.2 Verify >69,000 scfm as indicated on PID S52432. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 23 STANDARD NOTE Always start RAB Normal Supply Fan, HVR0002A(B), of same Note reviewed.

designation as operating Exhaust Fan, HVR0009A(B).

Comment:

SAT / UNSAT Revision 1 Page 10 of 12 2011 NRC Exam

JPM S7 TASK ELEMENT 24 STANDARD 7.1.13 Start RAB Normal Supply Fan, HVR0002 B. Fan Started Comment: Critical SAT / UNSAT TASK ELEMENT 25 STANDARD 7.1.14 Operate CVAS Unit A(B) continuously for a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months End of Task with heaters energized.

Comment:

SAT / UNSAT END OF TASK Revision 1 Page 11 of 12 2011 NRC Exam

JPM S7 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-196 There are no malfunctions for this JPM.

Revision 1 Page 12 of 12 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE S8 Place Reactor Power Cutback in Service Applicant:

Examiner:

JPM S8 JOB PERFORMANCE MEASURE DATA PAGE Task: Place Reactor Power Cutback in Service.

Task Standard: The applicant placed Reactor Power Cutback in service in accordance with OP-004-015, Reactor Power Cutback System.

References:

OP-004-015, Reactor Power Cutback System Alternate Path: No Time Critical: No Validation Time: 20 mins.

K/A 016 A4.01 NNI channel select controls Importance Rating 2.9 / 2.8 RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 12 2011 NRC Exam

JPM S8 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-004-015, Reactor Power Cutback System

==

Description:==

This task is performed at CP-2 and CP-7. The applicant must perform OP-004-015 section 6.1 and place Reactor Power Cutback in service. This will require selecting the proper sub-groups for both Reactor Power Cutback events and removing Reactor Trip on Turbine Trip from service.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 0 Page 3 of 12 2011 NRC Exam

JPM S8 APPLICANT CUE SHEET (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is at 100% power.

Reactor trip on Turbine trip is in service.

OP-004-015, section 5.1, Reactor Power Cutback System Standby Alignment, has been completed.

Attachment 11.1, Manual CEA Subgroup Selection has been completed and reviewed.

Subgroups 5 and 11 are both required for current plant conditions.

INITIATING CUES:

The CRS directs you to place Reactor Power Cutback in service.

Revision 0 Page 4 of 12 2011 NRC Exam

JPM S8 Evaluator Note Cue the Simulator Operator to place the Simulator in RUN.

TASK ELEMENT 1 STANDARD 6.1.1: Verify Section 5.1, Reactor Power Cutback System Standby Included in cue sheet.

Alignment, completed.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 6.1.1.1: Depress LAMP TEST pushbutton and verify all pushbuttons on Lamps tested.

panel illuminate.

Comment:

SAT / UNSAT TASK ELEMENT 3 STANDARD 6.1.1.1.2 Release LAMP TEST pushbutton. Button released.

Comment:

SAT / UNSAT TASK ELEMENT 4 STANDARD 6.1.2: Verify AUTO ACTUATE OUT OF SERVICE pushbutton Verification complete.

Illuminated.

Comment:

SAT / UNSAT Revision 0 Page 5 of 12 2011 NRC Exam

JPM S8 TASK ELEMENT 5 STANDARD 6.1.3: If the TEST RESET pushbutton is illuminated, then depress the Verification complete.

TEST RESET pushbutton and verify pushbutton extinguishes.

Comment:

SAT / UNSAT TASK ELEMENT 6 STANDARD 6.1.4: Verify Reactor Pwr Cutback Single Chnl Trouble (L-5, Cabinet Verification complete.

H) annunciator Clear.

Comment:

Alarm will be clear.

SAT / UNSAT TASK ELEMENT 7 STANDARD 6.1.5: Verify MANUAL SELECT Illuminated on AUTO SELECT Verification complete.

/MANUAL SELECT pushbutton.

Comment:

SAT / UNSAT TASK ELEMENT 8 STANDARD 6.1.6: Determine the appropriate CEA subgroup selection by This Attachment has already performing Attachment 11.1, Manual CEA Subgroup Selection. been completed.

Comment:

Evaluator: If asked, inform the applicant that Attachment 11.1 concluded that subgroups 5 and 11 were necessary for both RXC SAT / UNSAT events.

Revision 0 Page 6 of 12 2011 NRC Exam

JPM S8 TASK ELEMENT 9 STANDARD 6.1.7.1: Depress ENTER MANUAL SUBGRPS SELECT pushbutton Manipulation completed.

and verify pushbutton Illuminates.

Comment: Critical SAT / UNSAT TASK ELEMENT 10 STANDARD 6.1.7.2: Establish CEA subgroup pattern by Depressing desired SUBGROUP SELECT pushbuttons and verifying each selected Subgroup 5 and 11 pressed.

pushbutton Illuminates.

Comment: Critical On Reactor Power Cutback, Regulating Group 5 and 6 drop. This corresponds to sub-groups 5 and 11. 5 and 11 should be pressed, not SAT / UNSAT 5 and 6.

TASK ELEMENT 11 STANDARD 6.1.7.3: Depress LARGE LOAD REJECT pushbutton and verify Manipulation completed.

pushbutton Illuminates.

Comment: Critical SAT / UNSAT TASK ELEMENT 12 STANDARD 6.1.7.4.1: When the SUBGROUP SELECT and LARGE LOAD REJECT pushbuttons have Extinguished (after approximately 60 Manipulation completed.

seconds), then depress DISPLAY SUBGRP SELECT pushbutton and verify pushbutton Illuminates.

Comment:

SAT / UNSAT Revision 0 Page 7 of 12 2011 NRC Exam

JPM S8 TASK ELEMENT 13 STANDARD 6.1.7.4.2: Depress LARGE LOAD REJECT pushbutton and verify Manipulation completed.

pushbutton Illuminates.

Comment:

SAT / UNSAT TASK ELEMENT 14 STANDARD 6.1.7.4.3: Verify correct CEA subgroup pattern is displayed. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 15 STANDARD 6.1.8.1: Depress ENTER MANUAL SUBGRPS SELECT pushbutton Manipulation completed.

and verify pushbutton Illuminates.

Comment: Critical SAT / UNSAT TASK ELEMENT 16 STANDARD 6.1.8.2: Establish CEA subgroup pattern by Depressing desired SUBGROUP SELECT pushbuttons and verifying each selected Subgroup 5 and 11 pressed.

pushbutton Illuminates.

Comment: Critical On Reactor Power Cutback, Regulating Group 5 and 6 drop. This corresponds to sub-groups 5 and 11. 5 and 11 should be pressed, not SAT / UNSAT 5 and 6.

Revision 0 Page 8 of 12 2011 NRC Exam

JPM S8 TASK ELEMENT 17 STANDARD 6.1.8.3: Depress LOSS OF FEED PUMP pushbutton and verify Manipulation completed.

pushbutton Illuminates.

Comment: Critical SAT / UNSAT TASK ELEMENT 18 STANDARD 6.1.8.4.1: When the SUBGROUP SELECT and LOSS OF FEED PUMP pushbuttons have Extinguished (after approximately 60 Manipulation completed.

seconds), then depress DISPLAY SUBGRP SELECT pushbutton and verify pushbutton Illuminates.

Comment:

SAT / UNSAT TASK ELEMENT 19 STANDARD 6.1.8.4.2: Depress LOSS OF FEED PUMP pushbutton and verify Manipulation completed.

pushbutton Illuminates.

Comment:

SAT / UNSAT Revision 0 Page 9 of 12 2011 NRC Exam

JPM S8 TASK ELEMENT 20 STANDARD 6.1.8.4.3: Verify correct CEA subgroup pattern is displayed. Verification complete.

Comment:

Lights extinguish after ~ 60 seconds.

SAT / UNSAT TASK ELEMENT 21 STANDARD Procedure Note: Turbine DEH System Program has a minimum floor of 20% power. A Reactor Cutback rod configuration should not be Note reviewed.

selected that would drop Reactor Power below 20% in the event of a Reactor Power Cutback.

Comment:

SAT / UNSAT TASK ELEMENT 22 STANDARD 6.1.9: Verify both Main Feedwater Pumps operating. Verification complete.

Comment:

SAT / UNSAT TASK ELEMENT 23 STANDARD 6.1.10: Depress AUTO ACTUATE OUT OF SERVICE pushbutton and Manipulation complete.

verify pushbutton Extinguishes.

Comment: Critical SAT / UNSAT Revision 0 Page 10 of 12 2011 NRC Exam

JPM S8 TASK ELEMENT 24 STANDARD 6.1.11.1 On CP-2, place LOSS OF LOAD keyswitch to RPC. Switch was placed in RPC.

Comment: Critical SAT / UNSAT TASK ELEMENT 25 STANDARD 6.1.11.2 On CP-7, place all four LOSS OF TURB BYPASS Switches were all placed in keyswitches to BYPASS and verify and all four red BYPASS lamps bypass.

Illuminate.

Comment: Critical SAT / UNSAT TASK ELEMENT 26 STANDARD 6.1.11.3 On CP-2, place LOSS OF TURBINE TRIP keyswitch to Keyswitch was placed in DISABLE. DISABLE.

Comment: Critical SAT / UNSAT END OF TASK Revision 0 Page 11 of 12 2011 NRC Exam

JPM S8 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-196 Keys required to setup this scenario:

Keys 179 and 180 for CP-2 Keys 173 - 176 for CP-7 There are no malfunctions or overrides for this JPM.

Revision 0 Page 12 of 12 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE P1 Reset Emergency Feedwater Pump AB Applicant:

Examiner:

JPM P1 JOB PERFORMANCE MEASURE DATA PAGE Task: Reset overspeed device on Emergency Feedwater Pump AB during a Station Blackout.

Task Standard: The overspeed device on Emergency Feedwater Pump AB was reset in accordance with OP-902-005, Station Blackout Recovery.

References:

OP-902-005, Station Blackout Recovery Alternate Path: No Time Critical: No Validation Time: 10 mins.

K/A 061 A2.04, Pump failure or improper Importance Rating 3.4 / 3.8 operation RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 9 2011 NRC Exam

JPM P1 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-902-005, Station Blackout Recovery

==

Description:==

This task takes place in the RCA. The task takes place during a simulated Station Blackout. This requires the applicant to simulate manual operation of MS-416, Emergency Feedwater Pump AB Trip Valve. All components are located on the -35 elevation. This JPM takes place in a location with numerous pipes and supports that are bump hazards, but no special PPE will be required for this task. There are no special radiological requirements to perform this task.

This JPM can also be run using the EFW Pump AB mockup. If this device is used, the step numbers and actions are the same. When in the RCA, the applicant will still be required to demonstrate location of the components. The applicant must wear hand and eye PPE when operating the mockup.

When performing JPM validation, actions are necessary to ensure exam security is maintained.

Prior to commencing in plant JPM validation, contact Health Physics and direct them to disable all cameras in the CAA in a manner that prevents anyone from viewing any of the CAA cameras.

After all in plant JPMs are complete, contact Health Physics to restore the disabled cameras.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All steps for this JPM will be simulated, do not manipulate any plant components. Make all necessary communications to me. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 9 2011 NRC Exam

JPM P1 APPLICANT CUE SHEET Do Not Manipulate Any Plant Components (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

A Station Blackout event is in progress Emergency Feedwater Pump AB has tripped on an overspeed condition The Control Room has closed MS-401 A and MS-401 B.

INITIATING CUES:

The CRS directs you to reset Emergency Feedwater Pump AB in accordance with OP-902-005, Station Blackout Recovery, step 10.

Revision 1 Page 4 of 9 2011 NRC Exam

JPM P1 Evaluator Note This JPM can be performed by simulating the task in the field or performing the task on the EFW Pump AB mockup. All task elements should be executed in the same manner regardless of the method used. Gloves and eye protection should be available for the applicant if the mockup is being used.

Evaluator Note If the EFW Pump AB mockup is used, then the applicants are required to bring the examiner to EFW Pump AB during the other RCA in plant JPM.

TASK ELEMENT 1 STANDARD 10.1.a. Close BOTH PUMP AB TURB STM SUPPLY valves: MS-401 A Provided in the cue.

and MS-401 B.

Comment:

SAT / UNSAT Evaluator Note MS-407 is located at ground level against the wall when approaching EFW Pump AB.

TASK ELEMENT 2 STANDARD 10.1.b. Locally verify the steam supply header depressurized using MS-Verification complete.

407, EFW Pump AB Drip Pot Normal Drain Bypass.

Comment:

Examiner cue: MS-407 fails open on a loss of power. Provide cue to the applicant that MS-407 is open (upper limit switch engaged), stem fully up, SAT / UNSAT with no sound of steam flow.

There is a handwheel on MS-407. Closing MS-407 after verification is not required. If the applicant does close MS-407, provide the cue that as the handwheel is turned clockwise, the MS-407 stem moves down to the closed position.

Evaluator Note MS-416 trips closed on an overpseed. In a Station Blackout, MS-416 MOV must be manually operated in the closed direction to position the valve for reset. The physical indication as this operation is performed would be the trip bar moving from a low, tripped position up to the trip latch.

Revision 1 Page 5 of 9 2011 NRC Exam

JPM P1 Evaluator Note MS-416 in the tripped position before and after MOV is manually operated. As the applicant turns the handwheel in the clockwise direction, the bar will move from the lower position to the upper position.

TASK ELEMENT 3 STANDARD MS-416 is operated in the 10.1.c. Locally close MS 416, EFW Pump AB Turbine Stop Valve.

closed direction.

Comment: Critical Examiner cue: Latch bar moves from the lower position to the upper position as MS-416 handwheel is turned in the closed/clockwise direction. SAT / UNSAT Revision 1 Page 6 of 9 2011 NRC Exam

JPM P1 Evaluator Note The overspeed device is reset by moving the trip bar towards MS-416, by holding the rod and moving it from right to left.

The tappet nut drops down when bar is moved right to left.

TASK ELEMENT 4 STANDARD 10.1.d. Verify mechanical Overspeed reset. Overspeed device is reset.

Comment: Critical Examiner cue: Bar moves to the left, tappet nut drops down and stays down, bar maintains its position when released. SAT / UNSAT Revision 1 Page 7 of 9 2011 NRC Exam

JPM P1 Evaluator Note MS-416 in the closed position and the open position, before and after the MOV is manually operated. As the applicant turns the handwheel in the counter clockwise direction, the stem and gland assembly will move from the lower position to the upper position.

TASK ELEMENT 5 STANDARD MS-416 is manually 10.1.e. Locally open MS 416, EFW Pump AB Turbine Stop Valve.

opened.

Comment: Critical Examiner cue: Stem and gland assembly move up as the handwheel is turned in the counter clockwise direction. SAT / UNSAT TASK ELEMENT 6 STANDARD 10.1.f. Open BOTH PUMP AB TURB STM SUPPLY valves.

Communication made to MS 401A the Control Room.

MS 401B Comment:

SAT / UNSAT END OF TASK Revision 1 Page 8 of 9 2011 NRC Exam

JPM P1 TRAINER INSTRUCTIONS Position the EFW Pump AB mockup in the desired location. Position the cart in a manner that allows the applicant to access the mockup with MS-416 to the left and the manual trip tappet to the right.

Stage gloves and eye protection for the applicant.

Revision 1 Page 9 of 9 2011 NRC Exam

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE P2 Trip Emergency Diesel Generator B Locally Applicant:

Examiner:

JPM P2 JOB PERFORMANCE MEASURE DATA PAGE Task: Trip Emergency Diesel Generator B locally.

Task Standard: Applicant trips EDG B after initial efforts to trip the diesel fail.

References:

OP-009-002, Emergency Diesel Generator Alternate Path: Yes Time Critical: No Validation Time: 5 mins.

K/A 064 K4.02 Trips for EDG while operating Importance Rating 3.9 / 4.2 (normal or emergency) RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 7 2011 NRC Exam

JPM P2 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-009-002, Emergency Diesel Generator

==

Description:==

The applicant will be directed to trip EDG B locally. This JPM will require entry into the RCA, +21 level. The first method the applicant uses to trip the diesel will not function and the applicant will be required to trip the diesel using another method.

The reason for tripping EDG B will be due to a fuel oil leak, so depressing the System Reset pushbutton on the local control panel is required.

When performing JPM validation, actions are necessary to ensure exam security is maintained.

Prior to commencing in plant JPM validation, contact Health Physics and direct them to disable all cameras in the CAA in a manner that prevents anyone from viewing any of the CAA cameras.

After all in plant JPMs are complete, contact Health Physics to restore the disabled cameras.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All steps for this JPM will be simulated, do not manipulate any plant components. Make all necessary communications to me. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 7 2011 NRC Exam

JPM P2 APPLICANT CUE SHEET Do Not Manipulate Any Plant Components (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

Emergency Diesel Generator B is running in Emergency Mode A fuel oil leak has developed on the EDG B.

INITIATING CUES:

The CRS directs you to locally trip EDG B.

Revision 1 Page 4 of 7 2011 NRC Exam

JPM P2 TASK ELEMENT 1 STANDARD NOTE If control air is lost during any EDG run, then the Fuel Rack Override lever Note reviewed.

must be used to shutdown the EDG.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD CAUTION SUBSECTION 8.7 IS FOR EMERGENCY CONDITIONS, WHEN Caution reviewed.

NORMAL SHUTDOWN IS INOPERATIVE OR IT IS NECESSARY TO RAPIDLY STOP THE EMERGENCY DIESEL GENERATOR.

Comment:

SAT / UNSAT TASK ELEMENT 3 STANDARD NOTE Two Operators are required to secure the EDG in Emergency Mode using Note reviewed.

method 2 of step 8.7.2.

Comment:

SAT / UNSAT TASK ELEMENT 4 STANDARD CAUTION THE EDG WILL RESTART IF THE FUEL RACK OVERRIDE LEVER IS Caution reviewed.

RELEASED AND THE UNDERVOLTAGE OR SIAS SIGNAL IS PRESENT PRIOR TO COMPLETING STEP 8.7.2.2.2.

Comment:

SAT / UNSAT Revision 1 Page 5 of 7 2011 NRC Exam

JPM P2 Evaluator Note There are 2 acceptable methods to trip an EDG running in emergency mode.

The applicant could choose either method to perform first. Which ever method used first will fail, requiring the applicant to exercise the other method.

8.7.2 With the Emergency Diesel Generator B in Emergency Mode, Stop the Emergency Diesel Generator B by performing one of the following methods:

TASK ELEMENT 5 STANDARD 8.7.2.1 Method 1: Pull the manual Fuel Oil Overspeed Trip on the Overspeed pulled.

Overspeed Governor.

Comment: Critical This plunger is located on the upper level of the EDG. If this is attempted first, cue that the EDG B is still running. SAT / UNSAT TASK ELEMENT 6 STANDARD 8.7.2.2 Method 2:

8.7.2.2.1 Pull down and hold the Fuel Rack Override Lever on the North Handle is held down.

side of the Emergency Diesel Generator A(B) engine.

Comment: Critical If this is attempted first, cue the applicant that the lever did not move and that the EDG B is still running. SAT / UNSAT Evaluator Note If the applicant uses the fuel lever second, cue him that another operator has arrived to assist. After the applicant is holding the fuel rack lever, cue that another operator is now holding the lever.

TASK ELEMENT 7 STANDARD 8.7.2.2.2 To prevent the Emergency Diesel Generator A(B) from Starting, Unlock and Close the following valves:

Valves are closed.

EGA-152A(B) A(B) Air Receiver A2(B2) Outlet Isolation EGA-153A(B) A(B) Air Receiver A1(B1) Outlet Isolation Comment: Critical SAT / UNSAT Revision 1 Page 6 of 7 2011 NRC Exam

JPM P2 TASK ELEMENT 8 STANDARD 8.7.2.2.3 Release Fuel Rack Override Lever. Lever released.

Comment:

SAT / UNSAT TASK ELEMENT 9 STANDARD NOTE Depressing the System Reset pushbutton after the EDG has stopped will Note reviewed.

secure the Standby Fuel Oil Booster Pump, which may help mitigate the fuel oil leak.

Comment:

SAT / UNSAT TASK ELEMENT 10 STANDARD 8.7.3 If a fuel oil leak is in progress, when the Emergency Diesel Pushbutton depressed.

Generator A(B) has stopped then depress the System Reset pushbutton.

Comment: Critical Cue the applicant that EDG B has stopped rotating after the second trip is used. SAT / UNSAT TASK ELEMENT 11 STANDARD 8.7.4 Verify steps applicable or directed by the SM/CRS in Subsection 6.5, Control Room informed.

Unloading, Stopping and Returning EDG A(B) to Standby, are completed.

Comment:

SAT / UNSAT END OF TASK Revision 1 Page 7 of 7 2011 NRC Exam

System Operating Procedure OP-009-002 Emergency Diesel Generator Revision 312 8.7 PERFORMING AN EMERGENCY SHUTDOWN OF THE EMERGENCY DIESEL GENERATOR NOTE If control air is lost during any EDG run, then the Fuel Rack Override lever must be used to shutdown the EDG.

CAUTION SUBSECTION 8.7 IS FOR EMERGENCY CONDITIONS, WHEN NORMAL SHUTDOWN IS INOPERATIVE OR IT IS NECESSARY TO RAPIDLY STOP THE EMERGENCY DIESEL GENERATOR.

8.7.1 With Emergency Diesel Generator A(B) in Test Mode:

8.7.1.1 For no Cooldown cycle, then perform one of the following:

Depress Emergency Stop Pushbutton on CP-1 Depress the Emergency Stop Push Button on Emergency Diesel Generator A(B) Control Panel Pull manual Fuel Oil Overspeed Trip on Overspeed Governor Pull down and hold Fuel Rack Override Lever on North side of Emergency Diesel Generator A(B) engine until the engine comes to a complete Stop 8.7.1.2 For a 5 minute Cooldown cycle, then perform one of the following:

Position Diesel Cranking Control Switch at CP-1 to Stop Position Control Switch on Emergency Diesel Generator A(B) Control Panel to Stop 45

System Operating Procedure OP-009-002 Emergency Diesel Generator Revision 312 NOTE Two Operators are required to secure the EDG in Emergency Mode using method 2 of step 8.7.2.

CAUTION THE EDG WILL RESTART IF THE FUEL RACK OVERRIDE LEVER IS RELEASED AND THE UNDERVOLTAGE OR SIAS SIGNAL IS PRESENT PRIOR TO COMPLETING STEP 8.7.2.2.2.

8.7.2 With the Emergency Diesel Generator A(B) in Emergency Mode, Stop the Emergency Diesel Generator A(B) by performing one of the following methods:

8.7.2.1 Method 1: Pull the manual Fuel Oil Overspeed Trip on the Overspeed Governor.

or 8.7.2.2 Method 2:

8.7.2.2.1 Pull down and hold the Fuel Rack Override Lever on the North side of the Emergency Diesel Generator A(B) engine.

8.7.2.2.2 To prevent the Emergency Diesel Generator A(B) from Starting, Unlock and Close the following valves:

EGA-152A(B) A(B) Air Receiver A2(B2) Outlet Isolation EGA-153A(B) A(B) Air Receiver A1(B1) Outlet Isolation 8.7.2.2.3 Release Fuel Rack Override Lever.

NOTE Depressing the System Reset pushbutton after the EDG has stopped will secure the Standby Fuel Oil Booster Pump, which may help mitigate the fuel oil leak.

8.7.3 If a fuel oil leak is in progress, when the Emergency Diesel Generator A(B) has stopped then depress the System Reset pushbutton.

8.7.4 Verify steps applicable or directed by the SM/CRS in Subsection 6.5, Unloading, Stopping and Returning EDG A(B) to Standby, are completed.

46

Waterford 3 2011 NRC Exam JOB PERFORMANCE MEASURE P3 Close Safety Injection Tank Outlet Isolation during Control Room Evacuation Applicant:

Examiner:

JPM P3 JOB PERFORMANCE MEASURE DATA PAGE Task: Close Train B Safety Injection Tank Outlet valves during Control Room evacuation in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

Task Standard: Applicant closes SI-331 B and SI-332 B in accordance with OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown, Attachment 19, SIT OUTLET ISOLATION VALVE CLOSURE.

References:

OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown Alternate Path: No Time Critical: No Validation Time: 15 mins.

K/A 068 AA1.28 PZR level control and pressure Importance Rating 3.8 / 4.0 control RO / SRO Applicant:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 1 Page 2 of 6 2011 NRC Exam

JPM P3 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown

==

Description:==

The applicant will be cued that the Control Room has been evacuated due to a fire in CP-33. OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown, has been entered. He will be directed to close SI-331 B and SI-332 B, SIT 1B and 2B outlet isolation valves. This will require operating a permissive keyswitch at each breaker, closing the breakers, and closing the associated valves from the Remote Shutdown Panel. Only the first Train B valve will be performed due to protected train considerations combined with the similarity of the actions.

When performing JPM validation, actions are necessary to ensure exam security is maintained.

Prior to commencing in plant JPM validation, contact Health Physics and direct them to disable all cameras in the CAA in a manner that prevents anyone from viewing any of the CAA cameras.

After all in plant JPMs are complete, contact Health Physics to restore the disabled cameras.

DIRECTION TO APPLICANT:

I will explain the initial conditions, and state the task to be performed. All steps for this JPM will be simulated, do not manipulate any plant components. Make all necessary communications to me. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

Read the Initial Condition and Cues from the colored Applicant Cue Sheet, and then give the cue sheet to the applicant.

Revision 1 Page 3 of 6 2011 NRC Exam

JPM P3 APPLICANT CUE SHEET Do Not Manipulate Any Plant Components (TO BE RETURNED TO EXAMINER UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The Control Room has been evacuated due to a fire in CP-33.

The CRS has entered OP-901-502, Evacuation of Control Room and Subsequent Plant Shutdown.

RCS cooldown is in progress.

RCS pressure is 690 PSIA INITIATING CUES:

The CRS directs you to perform Attachment 19, SIT Outlet Isolation Valve Closure, commencing with step 3 for Train B SIT Tank Outlet, SI-331 B and SI-332 B.

There are no other operators to assist, perform (simulate) all procedure steps.

Revision 1 Page 4 of 6 2011 NRC Exam

JPM P3

3. To Close SI-331B, Safety Injection Tank 1B Outlet Isol, perform the following:

TASK ELEMENT 1 STANDARD 3.1 At SI-EBKR-311B-8H, Safety Injection Tank 1B Outlet Isol breaker, perform the following: Given in cue.

3.1.1 Verify Pressurizer pressure is between 700 and 685 psia.

Comment:

SAT / UNSAT TASK ELEMENT 2 STANDARD 3.1.2 Place in BYPASS, SIT 1B Isol Press Bypass Close Permissive key Keyswitch is in bypass.

switch at SI-EBKR-311B-8H.

Comment: Critical SAT / UNSAT TASK ELEMENT 3 STANDARD 3.1.3 Unlock AND Close breaker, SI-EBKR-311B-8H. Breaker is closed.

Comment: Critical SAT / UNSAT Revision 1 Page 5 of 6 2011 NRC Exam

JPM P3 Evaluator Note When the applicant gets to the Remote Shutdown Panel Room, knock on the door for access; the applicant does not have to obtain a set of keys for access.

When in front of the Remote Shutdown Panel, cue the applicant that all lights are energized as expected.

TASK ELEMENT 4 STANDARD 3.2 Close SI-331B, Safety Injection Tank 1B Outlet Isol, from LCP-43. SI-331 B is closed.

Comment: Critical Cue the applicant that the control switch for SI-331 B is red before operation, red and green during the valve stroke, and green when the SAT / UNSAT valve is full closed, about 1 minute later.

Evaluator Note After the applicant has closed SI-331 B from the Remote Shutdown Panel, end the JPM. Performance of step 4 is not required.

END OF TASK Revision 1 Page 6 of 6 2011 NRC Exam

Evacuation of Control Room OP-901-502 and Subsequent Plant Shutdown Revision 019 Page 1 of 2 ATTACHMENT 19 SIT OUTLET ISOLATION VALVE CLOSURE

1. To Close SI-331A, Safety Injection Tank 1A Outlet Isol, perform the following:

1.1 At SI-EBKR-311A-8H, Safety Injection Tank 1A Outlet Isol breaker, perform the following:

1.1.1 Verify Pressurizer pressure is between 700 and 685 psia.

1.1.2 Place in BYPASS, SIT 1A Isol Press Bypass Close Permissive key switch at SI-EBKR-311A 8H.

1.1.3 Unlock AND Close SI-EBKR-311A-8H.

1.2 Close SI-331A, Safety Injection Tank 1A Outlet Isol, from LCP-43.

2. To Close SI-332A, Safety Injection Tank 2A Outlet Isol, perform the following:

2.1 At SI-EBKR-311A-8M, Safety Injection Tank 2A Outlet Isol breaker, perform the following:

2.1.1 Verify Pressurizer pressure is between 700 and 685 psia.

2.1.2 Place in BYPASS, SIT 2A Isol Press Bypass Close Permissive key switch at SI-EBKR-311A-8M.

2.1.3 Unlock AND Close, SI-EBKR-311A-8M.

2.2 Close SI-332A, Safety Injection Tank 2A Outlet Isol, from LCP-43.

140

Evacuation of Control Room OP-901-502 and Subsequent Plant Shutdown Revision 019 Page 2 of 2 ATTACHMENT 19 SIT OUTLET ISOLATION VALVE CLOSURE (CONTD)

3. To Close SI-331B, Safety Injection Tank 1B Outlet Isol, perform the following:

3.1 At SI-EBKR-311B-8H, Safety Injection Tank 1B Outlet Isol breaker, perform the following:

3.1.1 Verify Pressurizer pressure is between 700 and 685 psia.

3.1.2 Place in BYPASS, SIT 1B Isol Press Bypass Close Permissive key switch at SI-EBKR-311B-8H.

3.1.3 Unlock AND Close breaker, SI-EBKR-311B-8H.

3.2 Close SI-331B, Safety Injection Tank 1B Outlet Isol, from LCP-43.

4. To Close SI-332B, Safety Injection Tank 2B Outlet Isol perform the following:

4.1 At SI-EBKR-311B-8M, Safety Injection Tank 2B Outlet Isol breaker, perform the following:

4.1.1 Verify Pressurizer pressure is between 700 and 685 psia.

4.1.2 Place in Bypass, SIT 2B Isol Press Bypass Close Permissive key switch at SI-EBKR-311B-8M.

4.1.3 Unlock AND Close SI-EBKR-311B-8M.

4.2 Close SI-332B, Safety Injection Tank 2B Outlet Isol, from LCP-43.

141

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 1 Op Test No.: NRC Examiners: Operators:

Initial Conditions: Reactor power is 100%

Protected Train is A AB Bus is aligned to Train A Turnover: Maintain 100% power Event Malf. No. Event Type* Event No. Description I - ATC Pressurizer level instrument RC-ILI-0110 X 1 RC15A2 I - SRO fails low. OP-901-110, Pressurizer Level TS - SRO Control Malfunction.

Steam Generator #1 Feedwater flow I - BOP instrument FW-IFR-1111 fails low. OP-901-2 FW26A I - SRO 201, Steam Generator Level Control TS - SRO Malfunction.

C - BOP CEA 52 Drops into the core 3 RD02A52 C - SRP OP-901-102, CEA or CEDMCS Malfunction TS - SRO R - ATC OP-901-212, Rapid Plant Power Reduction.

3 N/A R - BOP N - SRO Loss of Coolant Accident, OP-902-002, Loss RC23A of Coolant Accident Recovery.

4 M - All CS04A CS-125 A fails closed Secure RCPs (Critical Task 1)

C - ATC Charging Pump A fails to auto-start.

5 CV02A C - SRO C - BOP Low Pressure Safety Injection Pump A fails 6 SI02D to auto start on SIAS requiring manual start C - SRO Containment Spray Pump B trip, OP-902-C - BOP 008, Safety Function Recovery Procedure 7 CS01B C - SRO Alignment of LPSI Pump B to replace CS Pump B. (Critical Task 2)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 1 Rev 2

Scenario Event Description NRC Scenario 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After taking the shift, Pressurizer level instrument RC-ILI-0110X fails low. Due to the failure, Letdown flow goes to minimum flow and both backup Charging Pumps start.

The SRO should enter OP-901-110, Pressurizer Level Control Malfunction. The crew should utilize sub section E1, Pressurizer Level Control Channel Malfunction. The ATC should take manual control of Pressurizer level and select the non-faulted channel.

Using Tech Specs and OP-903-013, Monthly Channel Checks, the SRO should enter Tech Spec 3.3.3.5, a 7 day action requirement, and determine Tech Spec 3.3.3.6 entry is not required since QSPDS is operable and meeting the Pressurizer level channel check. SPDS indication of Pressurizer level on the Plant Monitoring Computer is affected by this failure.

After the non-faulted channel is selected and Tech Specs are addressed, Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low. The Feedwater Control System will respond by raising Feedwater flow to Steam Generator #1. The SRO should enter OP-901-201, Steam Generator Level Control Malfunction. The BOP will be required to take manual control and match Feedwater and Main Steam flow. The Ultrasonic Flow Meter will fail as a result of the instrument failure and require entry into TRM 3.3.5. The Feedwater controls for Steam Generator #1 will remain in manual as a result of this failure.

After the crew has addressed the Feedwater instrument failure, CEA 52 drops into the core. Off normal procedure OP-901-102, CEA or CEDMCS Malfunction, should be entered. The dropped CEA will require a rapid plant power reduction. The SRO should enter OP-901-212, Rapid Plant Power Reduction. Direct Boration should commence within 15 minutes of the dropped CEA. For the power reduction, the ATC will perform direct Boration to the RCS as well as ASI control with CEAs and Pressurizer Boron Equalization. The BOP will manipulate the controls to reduce Main Turbine load and manipulate Feedwater to Steam Generator #1 in manual. The SRO should enter Tech Specs 3.2.3, 3.1.3.1, and 3.1.3.5.

Once the crew has commenced the power reduction and lowered power to ~ 90%, or at the lead examiners discretion, a loss of coolant accident will occur. Charging Pump A will fail to start on the lowering Pressurizer level. The crew should diagnose the Pressurizer level dropping with all available Charging Pumps operating, trip the Reactor, and initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS). When Containment Spray is actuated, either manually or automatically, CS-125 A will fail to automatically open and will not open using the control switch. This does not create a need for action at this time, but Containment Spray flow will only be provided from Train B with CS-125 A failed closed. Low Pressure Safety Injection Pump A will fail to automatically start on SIAS, requiring the BOP operator to manually start LPSI Pump A.

Scenario 1 Rev 2

Scenario Event Description NRC Scenario 1 After the crew completes OP-902-000, Standard Post Trip Actions and diagnoses into OP-902-002, Loss of Coolant Accident Recovery, Containment Spray Pump B will trip, resulting in no Containment Spray flow. The crew should recognize that they are not meeting the Safety Function Status Checklist of OP-902-002 and transition to OP-902-008, Safety function Recovery Procedure.

Prioritization in OP-902-008 should result in Containment Isolation being priority 1 and Containment Temperature and Pressure Control being priority 2. The crew should address Containment Isolation by overriding CS-125 B closed. The crew should address Containment Temperature and Pressure Control by aligning Low Pressure Safety Injection Pump B to replace the failed Containment Spray Pump B. It is acceptable to pursue these tasks in parallel, since establishing flow with LPSI B to the Containment Spray header will also satisfy Containment Isolation concerns.

The scenario can be terminated after Low Pressure Safety Injection Pump B is aligned for Containment Spray, or after the CRS gives the order to perform that alignment, at the lead examiners discretion.

Scenario 1 Rev 2

NRC Scenario 1 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. This task becomes applicable after Containment Spray is initiated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling.

2. Establish Containment temperature and pressure control.

This task is satisfied by aligning LPSI Pump B to replace CS Pump B prior to exiting the Containment Temperature and Pressure Control safety function in OP-902-008. This task becomes applicable following the failure of Containment Spray Pump B. The Functional Recovery procedure utilized following this failure will direct this activity to satisfy the Containment Pressure and Temperature Control safety function.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 7

2. Malfunctions after EOP entry (1-2) 2

3. Abnormal events (2-4) 3

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 2

6. EOP contingencies requiring substantive actions (0-2) 1

7. Critical tasks (2-3) 2 Scenario 1 Rev 2

NRC Scenario 1 Scenario Notes:

A. Reset Simulator to IC-191.

B. Verify the following Scenario Malfunctions:

1. rc15a for Pressurizer level

2. fw26a for Steam Generator #1 Feedwater flow

3. rd02a52 for CEA 52

4. rc23a for LOCA

5. cv02a for Charging Pump A

6. si02d for Low Pressure Safety Injection Pump A

7. cs01b for Containment Spray Pump B

8. cs04a for CS-125 A C. Verify the following Override:

1. di-08a04s22-1 for CS-125 A D. Ensure Protected Train A sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

G. Start DCS, Record Data, select file PlantParameters.txt.

Scenario 1 Rev 2

NRC Scenario 1 Simulator Booth Instructions Event 1 Pressurizer Level Instrument RC-ILI-0110X Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 1.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If sent to LCP-43, report RC-ILI-0110 X1 is failed low.

Event 2 Steam Generator #1 Feedwater Flow Instrument FW-IFR-1111 Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 CEA 52 Drops, Rapid Plant Power Reduction

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If called to remove Condensate Polishers from service, acknowledge communication and report that you will perform actions requested.

3. If Work Week Manager or I&C is called, inform the caller that a and a team will be sent to the CEDMCS Alley to investigate.

Event 4 LOCA Inside Containment

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If called as RCA watch report CS-125 A appears to be mechanically bound, the stem looks bent.

3. If called as RAB watch to check the Emergency Diesel Generators, use remote EGR26 and 27. When EDG A & B Trouble alarms clear, report they are running satisfactorily.

4. If the Duty Plant Manager is called, inform the caller that he will make the necessary calls.

Event 5 Low Pressure Safety Injection Pump A fails to start

1. If called to check the LPSI Pump A breaker, report all indications are normal.

2. If called to check the LPSI Pump A locally, report all indications are normal.

Scenario 1 Rev 2

NRC Scenario 1 Event 6 Containment Spray Pump B Trips

1. After the crew has entered OP-902-002 and on the Lead Examiner's cue, initiate Event Trigger 7.

2. If called to check the Containment Spray Pump B breaker, report over-current flags are picked up on all 3 phases.

3. If called to check the Containment Spray Pump B, report that there are visible charring on the motor with an acrid smell, but no indications of a fire or smoke.

4. If called for TSC concurrence, report SM/EC has granted concurrence.

5. If called as RAB watch to come to the Control Room for over-ride key for CS-125 B, acknowledge communication. Report to the Control Room on lead examiners cue.

6. If crew does obtain key and over-rides CS-125 B closed, use remote CSR13B for the local key operation.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario1.cdf. Save the file into the folder for the appropriate crew.

Scenario 1 Rev 2

NRC Scenario 1 Scenario Timeline:

Ramp Event Malfunction Severity Delay Trigger HH:MM:SS 1 RC15A2 0 N/A N/A 1 Pressurizer level instrument RC-ILI-0110 X fails low 2 FW26A 0 N/A N/A 2 Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low 3 RD02A52 N/A N/A N/A 3 CEA 52 Drops into the core 4 RC23A 3.0 % 8:00 N/A 4 Loss of Coolant Accident 5 CV02A N/A N/A N/A 5 Charging Pump A fails to auto-start 6 SI02D N/A N/A N/A N/A Low Pressure Safety Injection Pump A fails to auto start 7 CS04A N/A N/A N/A N/A DI-08a04s22-1 CS-125 A Fails to open, will not open manually.

7 CS01B N/A N/A N/A 7 Containment Spray Pump B trip Scenario 1 Rev 2

NRC Scenario 1

REFERENCES:

Event Procedures 1 OP-901-110, Pressurizer Level Control Malfunction OP-903-013, Monthly Channel Checks Tech Spec 3.3.3.5 2 OP-901-201, Steam Generator Level Control Malfunction Tech Requirement Manual 3.3.5 3 OP-901-102, CEA or CEDMCS Malfunction OP-901-212, Rapid Plant Power Reduction OP-004-004, Control Element Drive Tech Spec 3.2.3, 3.1.3.1, 3.1.3.5 4 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-002, Loss of Coolant Accident Recovery 5 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 6 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 7 OP-902-008, Safety Function Recovery Procedure OP-902-009, Standard Appendices, Appendix 28, Aligning LPSI to Replace CS Scenario 1 Rev 2

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 2 Op Test No.: NRC Examiners: Operators:

Initial Conditions: Reactor power is 77%

Protected Train is B AB Bus is aligned to Train B Turnover: Charging Pumps A & B are operating Boron Equalization is in progress Event Malf. No. Event Type* Event No. Description Pressurizer pressure instrument RC-IPR-I - ATC 1 RX14A 0100 X fails low, OP-901-120, Pressurizer I - SRO Pressure Control Malfunction I - BOP RCP 1A speed instrument failure, Channel B, 2 RC16B I - SRO Core Protection Calculator B trip TS - SRO Letdown Back Pressure controller CVC-IPIC-I - ATC 3 DI-04a3a02e-5 0201 setpoint fails to 100% output. OP-901-I - SRO 112, Charging or Letdown Malfunction.

Dry Cooling Tower Fan 8B failure 4 N/A TS - SRO DI-07a8s06-1 I - BOP Inadvertent Containment Spray Actuation 5 OP-901-504, Inadvertent ESFAS Actuation DI-07a8s12-1 I - SRO Main Steam line break inside Containment, S/G #2, OP-902-004, Excess Steam Demand 6 MS11B M - All Recovery (Critical Task 1, 3, and 4)

C - BOP Initiate Containment Spray flow 7 N/A (Critical Task 2)

C - SRO C - ATC Relay K301 failure, BAM-113 A and CVC-8 RP09E 183 fail to position on Safety Injection C - SRO

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 2 Rev 2

Scenario Event Description NRC Scenario 2 The crew assumes the shift at ~77% power with instructions to hold power pending planned Chemistry Department hold.

After assuming the shift, Pressurizer pressure instrument RC-IPR-0100 X will fail low.

Since Boron Equalization is in progress, the Main Spray valves will close. The SRO will enter OP-901-120, Pressurizer Pressure Control Malfunction, and select the non-faulted pressure channel.

After Channel Y has been selected for Pressurizer pressure control, Reactor Coolant Pump 1A speed sensor for Core Protection Calculator B will fail. CPC B will trip as a result of the failure. The SRO should enter Tech Spec 3.3.1 and have the BOP operator bypass bistables 3 and 4 on Channel B.

After the bypass operation is complete, the Letdown Back Pressure controller, CVC-IPIC-0201, setpoint fails to 700 PSIA, 100% scale. This causes the in service Letdown Back Pressure control valve to close and Letdown flow to go to 0 gpm. The CRS should enter OP-901-112, Charging or Letdown Malfunction, and use sub-section E2 to address the failure. The Letdown Flow controller and the Back Pressure controller will be placed in manual to control Letdown flow.

After the ATC has control of the Letdown System in manual, the Outside Watch will call and report an oil failure on Dry Cooling Tower Fan 8B. The SRO should enter Tech Spec 3.7.4 action d. His review of ambient temperature and Tech Spec 3.7.4 should conclude that Train B Ultimate Heat Sink remains operable and that Tech Spec 3.8.1.1 is being complied with.

After the Tech Spec review is complete, an inadvertent Containment Spray Actuation will occur. Component Cooling Water flow to the Reactor Coolant Pumps will be secured. The SRO should enter OP-901-504, Inadvertent ESFAS Actuation. The Containment Spray Pumps should be secured. If the Component Cooling Water Isolations to the Reactor Coolant Pumps are not restored within 3 minutes, the reactor should be tripped and the Reactor Coolant Pumps secured.

A Main Steam line break will develop on Steam Generator #2 after the preceding event.

If the crew restored CCW to the Reactor Coolant Pumps, the crew should perform a manual reactor trip due to the excess steam demand. If the crew tripped the reactor and secured Reactor Coolant Pumps on the previous event, then the Main Steam line break will ramp in after the reactor trip. Because the Containment Spray Pumps control switches maintain off, the BOP should re-start Containment Spray Pumps A and B after Containment pressure rises above 17.7 psia.

Relay K301 will not actuate and BAM-113 A will fail to open and CVC-183 will fail to close on the Safety Injection Actuation. The ATC operator should position these valves to ensure Emergency Boration. After Steam Generator #2 blows down, the crew will take action to maintain RCS temperature and pressure. The scenario can be terminated after these actions are complete.

Scenario 2 Rev 2

NRC Scenario 2 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. The required task becomes applicable after Containment Spray has been actuated. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling. If the crew does not restore CCW flow to the RCPs after the inadvertent CSAS, then the 3 minute criteria starts at the time of that CSAS. If the crew restores CCW flow to the RCPs following the inadvertent CSAS, then the 3 minute criteria starts after the Main Steam line break.

2. Establish Containment temperature and pressure control This task is satisfied by manually starting at least 1 Containment Spray Pump following the Main Steam line break. This should be completed before completing the review of OP-902-000, Standard Post Trip Actions.

3. Establish RCS temperature control This task is satisfied by taking action to stabilize RCS temperature within the limits of the RCS P/T curve using ADV #1 and establishing EFW flow to Steam Generator #1.

Action to address this task should commence prior to RCS temperature exceeding 550 °F.

4. Establish RCS pressure control This task is satisfied by taking action to stabilize RCS pressure within the limits of the RCS P/T curve and additionally maintain RCS pressure within 1500-1600 psid of the faulted steam generator. Action to address this task should commence prior to RCS pressure exceeding 2250 PSIA.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 7

2. Malfunctions after EOP entry (1-2) 2

3. Abnormal events (2-4) 3

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3) 4 Scenario 2 Rev 2

NRC Scenario 2 Scenario Notes:

A. Reset Simulator to IC-192.

B. Verify the following Scenario Malfunctions:

1. rx14-A for Pressurizer pressure instrument RC-IPT-0100 X

2. rc16b for RCP 1A speed

3. ms11b for Main Steam line break S/G #2

4. rp09e for Relay K301 C. Verify the following Overrides:

1. di-07a08s06-1 and di-07a08s12-1 for CSAS

2. di-04a3a02e-5 for Letdown Back Pressure Controller D. Ensure Protected Train B sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

G. Start DCS, Record Data, select file PlantParameters.txt.

Scenario 2 Rev 2

NRC Scenario 2 Simulator Booth Instructions Event 1 Pressurizer Pressure Instrument Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 1.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 RCP 1A Speed Instrument Failure

3. On Lead Examiner's cue, initiate Event Trigger 2.

4. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 Letdown Back Pressure Controller Setpoint Failure

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If called as the RCA Watch to locally monitor the following Letdown DP indications, report that all DP indications are normal.

3. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 Dry Cooling Tower Fan 8B Fan Failure

1. On Lead Examiner's cue, call the CRS as the Outside Watch and report that Dry Cooling Tower Fan 8B has no oil in the reduction gear sightglass. There is oil on the ground under the fan. This discovery is made during rounds.

2. If Work Week Manager or PMM is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 Inadvertent CSAS

1. On Lead Examiner's cue, initiate Event Trigger 5.

2. No communications should occur for this evolution.

Event 6 Main Steam Line Break S/G #2

1. On the Lead Examiner's cue, or after the reactor is manually tripped in the previous event, initiate Event Trigger 6.

2. When called as the Outside Watch to check Main Steam Safeties not lifting, report that no safety valves are lifting.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario 2.cdf. Save the file into the folder for the appropriate crew.

Scenario 2 Rev 2

NRC Scenario 2 Scenario Timeline:

Ramp Event Malfunction Severity Delay Trigger HH:MM:SS 1 RX14A 0% N/A N/A 1 Pressurizer pressure RC-IPR-0100 X fails low 2 RC16B N/A N/A N/A 2 RCP 1A Speed failure, Channel B 3 Di-04a3a02e-5 Push N/A N/A 3 Letdown Back Pressure controller setpoint failure 4 N/A N/A N/A N/A N/A Dry Cooling Tower Fan 8B failure 5 Di-07a8a06-1 N/A N/A N/A 5 DI-07a8s12-1 Inadvertent Containment Spray 6 MS11B 10% 3:00 N/A 6 Main Steam line break, S/G #2 7 N/A N/A N/A N/A N/A Initiate Containment Spray flow 8 RP09E N/A N/A N/A N/A Relay K301 failure Scenario 2 Rev 2

NRC Scenario 2

REFERENCES:

Event Procedures 1 OP-901-120, Pressurizer Pressure Control Malfunction 2 OP-009-007, Plant Protection System Tech Spec 3.3.1 3 OP-901-112, Charging or Letdown Malfunction 4 Tech Spec 3.7.4 and 3.8.1.1 OP-100-014, Technical Specification and Technical Requirements Compliance 5 OP-901-504, Inadvertent ESFAS Actuation 6 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-004, Excess Steam Demand Recovery 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 8 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance Scenario 2 Rev 2

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 4 Op Test No.: NRC Examiners: Operators:

Initial Conditions: Reactor power is 100%

Protected Train is B AB Bus is aligned to Train A Turnover: Maintain 100% power Event Malf. No. Event Type* Event No. Description SG10D C - BOP Steam Generator #1 level instrument SG-ILI-1 C - SRO 1113 D fails high.

TS - SRO FW03A C - ATC Main Feedwater Pump A trips, Reactor 2 C - SRO Power Cutback TS - SRO OP-901-101, Reactor Power Cutback RD07D R - ATC Regulating Group 4 CEAs fail to insert in 3 automatic following Reactor Power Cutback TP01A C - BOP Turbine Cooling Water Pump A trips, Turbine 4 TP08B C - SRO Cooling Water Pump B fails to auto start OP-901-512, Loss of Turbine Cooling Water FW03B M - All Main Feedwater Pump B trips, manual 5 FW07A N - SRO reactor trip, Emergency Feedwater Pump A fails to run RP03 C - BOP Main Turbine fails to trip following the reactor 6 C - SRO trip RD11A C - ATC 3 CEAs fail to insert following the reactor trip, 7 28, 37, 79 C - SRO Emergency Boration (Critical task 1)

FW05 C - BOP Emergency Feedwater Pump AB trip on 8 C - ATC overspeed C - SRO (Critical task 2)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 4 Rev 2

Scenario Event Description NRC Scenario 4 The crew assumes the shift at 100% power with instructions to maintain 100% power.

After assuming the shift, Steam Generator #1 level instrument SG-ILI-1113 D fails high.

The SRO should review Tech Specs and enter Tech Spec 3.3.1 and 3.3.2 and TRM 3.3.1. The SRO should direct the BOP operator to bypass the bistables for low Steam Generator #1 level, high level and Steam Generator #1 differential pressure for channel D. This instrument does apply to Tech Spec 3.3.3.6 for Accident Monitoring, but the minimum channel requirements are met using other channels.

After the proper bistables are bypassed, Main Feedwater Pump A will trip. A Reactor Power Cutback will occur. The ATC should perform the immediate operator actions.

The SRO should enter OP-901-101, Reactor Power Cutback. Following the Cutback, Regulating Group 4 CEAs will fail to insert in automatic. The SRO should enter Tech Spec 3.2.4 for DNBR and 3.2.7 for ASI. The crew should take action to address the DNBR power operating limit within 15 minutes by performing ASI control with Group P CEAs.

After the crew has addressed Tech Specs and commenced ASI control with Group P CEAs, Turbine Cooling Water Pump A trips. Turbine Cooling Water Pump B fails to start. The SRO should enter OP-901-512, Loss of Turbine Cooling Water, and start Turbine Cooling Water Pump B. The Plant Monitoring Computer will display an overload condition for TCW Pump A After Turbine Cooling Water Pump B is running, Main Feedwater Pump B will trip. The crew should perform a manual reactor trip based on this failure. On the Emergency Feedwater Actuation, Emergency Feedwater Pump A will fail to start and will not start manually. The Main Turbine will fail to trip on the reactor trip. The BOP should manually trip the Main Turbine. 3 CEAs will fail to insert on the reactor trip. The ATC operator should perform Emergency Boration due to this condition. The SRO should enter OP-902-006, Loss of Main Feedwater Recovery. The ATC operator should secure 2 Reactor Coolant Pumps.

After 2 Reactor Coolant Pumps are secured, Emergency Feedwater Pump AB will trip due to operator error locally. The crew should remain in OP-902-006 and secure the remaining Reactor Coolant Pumps. On investigation, the local watchstander will report Emergency Feedwater Pump AB is ready to be reset. The BOP operator should perform the necessary actions for resetting Emergency Feedwater Pump AB.

The scenario can be terminated after Emergency Feedwater Pump AB is reset.

Scenario 4 Rev 2

NRC Scenario 4 Critical Tasks

1. Establish reactivity control.

This task is satisfied by establishing Emergency Boration prior to completing Standard Post Trip Actions Reactivity Control verification. The required task becomes applicable after the Reactor is tripped and 3 CEAs remain stuck out.

2. Establish a primary to secondary heat sink This task is satisfied by securing all RCPs after Emergency Feedwater Pump AB trips. With Emergency Feedwater Pump A off, Emergency Feedwater Pump B does not have the capacity to provide necessary Emergency Feedwater flow. The requirement is that all RCPs be secured within 30 minutes of the loss of Main Feedwater, the time of the reactor trip.

Scenario Quantitative Attributes

1. Total malfunctions (5-8) 8

2. Malfunctions after EOP entry (1-2) 3

3. Abnormal events (2-4) 2

4. Major transients (1-2) 1

5. EOPs entered/requiring substantive actions (1-2) 1

6. EOP contingencies requiring substantive actions (0-2) 0

7. Critical tasks (2-3) 2 Scenario 4 Rev 2

NRC Scenario 4 Scenario Notes:

A. Reset Simulator to IC-194.

B. Verify the following Scenario Malfunctions:

1. sg10d for S/G #1 level instrument

2. tp01a for TCW Pump A

3. tp08b for TCW Pump B

4. fw03a for Main Feedwater Pump A

5. rd07d for Regulating Group 4 CEAs

6. fw03b for Main Feedwater Pump B

7. fw07a for EFW Pump A

8. rp03 for the Main Turbine failure

9. rd11a28, 37, and 79 for CEAs 28, 37, and 79

10. fw05 for EFW Pump AB C. Verify the following Override:

1. di-08a04s09-1 for EFW Pump A D. Ensure Protected Train B sign is placed in SM office window.

E. Verify EOOS is 10.0 Green F. Complete the simulator setup checklist.

G. Start DCS, Record Data, select file PlantParameters.txt.

Scenario 4 Rev 2

NRC Scenario 4 Simulator Booth Instructions Event 1 Steam Generator #1 level instrument failure

1. On the Lead Examiner's cue, initiate Event Trigger 1.

2. If directed to check the remote shutdown panel, report that Channel D S/G #1 level reads 67%.

3. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2/3 Main Feedwater Pump A trip, Reactor Power Cutback/Reg Group 4 Failure

1. On the Lead Examiner's cue, initiate Event Trigger 3.

2. If directed to check Main Feedwater Pump A locally, report there are no abnormal indications locally.

Event 4 Turbine Cooling Water Pump A trip

1. On the Lead Examiner's cue, initiate Event Trigger 2.

2. If directed to check Turbine Cooling Water Pumps locally, report TCW Pump A has over-current flags tripped and that TCW Pump B looks normal.

3. If Work Week Manager is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 MFW Pump B trip, Reactor trip, Emergency Feedwater Pump A trip

1. On the Lead Examiner's cue, initiate Event Trigger 5.

2. If directed to check Main Feedwater Pump B locally, report indications of broken linkages on the governor assembly.

3. If directed to check EFW Pump A locally, report indications of a broken breaker for EFW Pump A at Switchgear 3A.

Event 8 Emergency Feedwater Pump AB trip

1. On the Lead Examiner's cue, initiate Event Trigger 8.

2. After the remaining Reactor Coolant Pumps are tripped, call as the RCA watch and report that the Emergency Feedwater Pump AB tripped on overspeed due to his activities while checking the pump. Recommend performing actions to reset EFW Pump AB.

At the end of the scenario, before resetting, complete data collection by stopping recording and saving the file as 2011 SRO Scenario 4.cdf. Save the file into the folder for the appropriate crew.

Scenario 4 Rev 2

NRC Scenario 4 Scenario Timeline:

Ramp Event Malfunction Severity Delay Trigger HH:MM:SS 1 SG10D 100% N/A N/A 1 S/G #1 level instrument channel D fails high FW03A N/A N/A N/A 2 MFW Pump A trips 3 RD07D N/A N/A N/A N/A Regulating Group 4 fails to auto insert 4 TP01A N/A N/A N/A 4 TP08B TCW Pump A trips, TCW Pump B fails to auto-start 5 FW03B N/A N/A N/A 5 FW07A DI-08a04s09-1 MFW Pump B trips, EFW Pump A fails to run 6 RP03 N/A N/A N/A N/A Main Turbine fails to trip on reactor trip 7 RD11A N/A N/A N/A N/A 28, 37, 79 CEAs 28, 37, 79 fail to insert 8 FW05 N/A N/A N/A 8 EFW Pump AB trips Scenario 4 Rev 2

NRC Scenario 4

REFERENCES:

Event Procedures 1 OP-009-007, Plant Protection System OP-903-013, Monthly Channel Checks Tech Spec 3.3.1 and 3.3.2 3&4 OP-901-101, Reactor Power Cutback Tech Spec 3.2.1 2 OP-901-512, Loss of Turbine Cooling Water 5 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-006, Loss of Main Feedwater Recovery 6 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 7 OP-902-000, Standard Post Trip Actions OI-038-000, Emergency Operating Procedures Operations Expectations /

Guidance 8 OP-902-006, Loss of Main Feedwater Recovery Scenario 4 Rev 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 1 of 20 Event

Description:

Pressurizer level instrument RC-ILI-0110 X fails low.

Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

PRESSURIZER LEVEL HI/LO (Cabinet H, B-1)

PRESSURIZER LEVEL LO-LO (Cabinet H, C-1)

LETDOWN FLOW HI/LO (Cabinet G, C-1)

Indications Mismatch between Charging (CVC-IFI-0212) AND Letdown (CVC-IFI-0202) flow indications Low level indicated on Pressurizer level indicator RC-ILI-0110 X Deviation between actual level AND programmed level as indicated on Pressurizer level recorder (RC-ILR-0110)

CRS Enter and direct the implementation of OP-901-110, Pressurizer Level Control Malfunction.

CRS In section E0, General, of OP-901-110, direct use of sub-section E1.

CRS There is a note at the start of sub-section E1 that reads:

Selecting the non-faulted channel may cause automatic actions to occur if actual level is not at program level.

The CRS should evaluate this note and either wait for Pressurizer level to return to setpoint before selecting Channel Y or acknowledge that additional Charging Pumps may start as a result of selecting Channel Y.

If he chooses to wait, it will take several minutes for Pressurizer level to return to setpoint. In this case, consider moving to malfunction 2 during the wait.

All ATC manipulations are located on CP-2 ATC 1: Place Pressurizer Level Controller (RC-ILIC-0110) in MAN AND adjust OUTPUT to slowly adjust letdown flow to restore Pressurizer level.

ATC 2: Transfer Pressurizer Level Control CHANNEL SELECT switch to non-faulted channel.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 2 of 20 Event

Description:

Pressurizer level instrument RC-ILI-0110 X fails low.

Time Position Applicants Actions or Behavior ATC 3: Transfer Pressurizer CHANNEL SELECT LO LEVEL HEATER CUTOFF switch to non-faulted channel.

ATC / CRS 4: Verify desired backup Charging pumps in AUTO.

ATC 5: Verify ALL PROPORTIONAL AND BACKUP HEATER BANKS reset.

ATC 6: Place Pressurizer Level Controller (RC-ILIC-0110) in AUTO and verify Pressurizer Level is being restored to setpoint.

ATC / CRS 7: Verify Pressurizer level controlling at program setpoint in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

CRS 8: Refer to Technical Specifications 3.3.3.5 and 3.3.3.6 for Remote Shutdown and Accident Monitoring operability determination.

OP-903-013, Monthly Channel Checks should be used to implement Tech Specs 3.3.3.5 and 3.3.3.6.

CRS Enter Tech Spec 3.3.3.5.a.

Since QSPDS meets the channel check requirement of OP-903-013, entry into Tech Spec 3.3.3.6 is not appropriate.

Note If Pressurizer level exceeds 62.5% during this malfunction, tech Spec 3.4.3.1 would also apply.

If Pressurizer pressure exceeds 2275 PSIA during this malfunction, tech Spec 3.2.8 would also apply.

There will be a RCS pressure rise and drop during this malfunction. The ATC may discuss the reactivity effects associated with this pressure rise.

Examiner Note This event is complete when Channel Y is selected for Pressurizer Level Control and for Pressurizer Lo Level Heater Cutout and after Tech Specs 3.3.3.5 and 3.3.3.6 have been addressed Or As directed by the Lead Evaluator Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 3 of 20 Event

Description:

Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low Time Position Applicants Actions or Behavior BOP Recognize and report indications of failed instrument.

Alarms:

SG 1 Steam/FW Flow Signal Dev (Cabinet F, T-17)

SG 1 Level Hi/Lo (Cabinet F, U-14)

COLSS MASTER (Cabinet L, A-6)

Indications Feedwater Flow indicator FW-IFR-1111 fails low Actual Feedwater flow to S/G #1 rises Actual S/G #1 level rises CRS / BOP Direct the BOP to take manual control and match Main Steam flow and Main Feedwater flow for S/G#1.

CRS Enter and direct the implementation of OP-901-201, Steam Generator Level Control Malfunction.

N/A 1. If Steam Generator level is <41% NR and there is no Feedwater flow to the Steam Generator, then perform the following:

1.1 Trip the Reactor.

CRS 2. Go to Attachment 1, General Actions.

Note The CRS and the BOP will work through the flow chart. The CRS will ask the BOP questions about the status of Feedwater to S/G #1.

Observe the affected Steam Generator FWCS controllers AND note ANY controllers that are behaving erratically None are behaving erratically Place appropriate controllers for the affected FWCS in manual AND establish control of S /G level (See Notes 1 & 2).

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 4 of 20 Event

Description:

Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low Time Position Applicants Actions or Behavior Is the output of the affected FWCS Master Controller behaving erratically?

No Verify SGFP Discharge pressure for BOTH SGFP 's is matched AND is greater than S/G pressures .

Stop turbine load changes except to match Tave and Tref .

Review the following guidelines AND restore S /G level to 50-70% NR:

1. IF one SGFP Speed controller is in auto , THEN use its output to help set the SGFP Speed controller that is in manual .

2. Momentary taps on the raise AND lower buttons of the Main Feedwater Reg Valve Controller have a noticable impact on associated Steam Generator level .

3. Use the Startup Feedwater Reg Valve Controller to control Steam Generator level at low power levels .

4. Use indications on the unaffected FWCS controllers to help set affected FWCS controllers .

Check the following Control Channel indicators to determine if a Control Channel has failed : (See Note 3)

FW IFR 1111, Steam Generator 1 Feedwater Flow (green pen ) is failed Control Channel level deviation of >7%?

No Main Feedwater Pump Speed Controller malfunction ?

No Is Feedwater flow for the affected S /G abnormally high ?

No Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 5 of 20 Event

Description:

Steam Generator #1 Feedwater flow instrument FW-IFR-1111 fails low Time Position Applicants Actions or Behavior Determine AND correct the cause of the malfunction .

Note At this point, the CRS and BOP would discuss the need to control S/G #1 level in manual. The crew should brief at this point in the event.

CRS The Ultrasonic Flow Meter quality goes to BAD on this malfunction. The ATC should not disrupt the CRS and the BOP when trying to stabilize S/G

1 level. The ATC should inform the CRS after the plant is stable and the flow chart is complete. TRM 3.3.5 entry is required on a failure of the UFM.

Examiner Note This event is complete when S/G #1 level is stabilized and the brief is completed Or As directed by the Lead Evaluator Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 6 of 20 Event

Description:

CEA 52 Drops into the core; Rapid Plant Power Reduction.

Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of dropped CEA Alarms Multiple alarms will come on Panel K Multiple alarms will come on Panel H CPC TILT EXCEEDED (Cabinet L, B-6)

EXERCISE LIMIT SHUTDOWN GROUP A(Cabinet L, B-7)

TECH SPEC TILT EXCEEDED (Cabinet L, C-6)

Indications Rod bottom light for CEA 52 TCOLD dropping LPD and DNBR trips on Channel B (targeted channel)

CRS Enter and direct the implementation of OP-901-102, CEA or CEDMCS Malfunction.

ATC 1. If in Mode 1 and two or more Control Element Assemblies drop or are misaligned by >19 inches, then manually trip the Reactor and go to OP-902-000, Standard Post Trip Actions.

This is the immediate actions for this conditions. They are not applicable to this scenario.

ATC 1. Place CEDMCS Mode Select switch to OFF.

N/A 2. If any of the following occur, then manually trip the Reactor and go to OP-902-000, Standard Post Trip Actions:

CRS 3. If Control Element Assembly is misaligned >7 inches, then go to section E1, CEA Misalignment Greater Than 7 Inches.

CRS 1. Match Tavg and Tref by performing the following:

Adjust Turbine load in accordance with OP-010-004, Power Operations, Adjust RCS boron concentration in accordance with OP-002-005, Chemical and Volume Control.

The CRS may direct a Main Turbine load reduction before entering the off normal procedure to raise TCOLD. It is not required to do this before entering OP-901-102.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 7 of 20 Event

Description:

CEA 52 Drops into the core; Rapid Plant Power Reduction.

Time Position Applicants Actions or Behavior CRS 2. Notify Duty Plant Manager and Duty Engineering.

CRS 3. Record time of CEA misalignment >7 inches in Station Log.

CRS 4. If CEA misalignment >19 inches, then go to step 8.

CRS Procedure Caution A power reduction must be started within 15 minutes of CEA misalignment

>7 inches to comply with tech spec 3.1.3.1.

CRS 8. If misalignment >19 inches or affected CEA is not aligned to within 7 inches of all other CEAs in the same group within 15 minutes, then perform the following:

Reduce power in accordance with OP-901-212, Rapid Plant Power Reduction to comply with Technical Specification 3.1.3.1.

Maintain TAVG at TREF by adjusting turbine load If PMC is Operable, then verify CEA Pulse Counter indication is correct or enter the correct CEA position in the PMC database.

Declare COLSS Inoperable and enter OP-901-501, PMC or COLSS Inoperable and perform concurrently with this procedure due to COLSS being Inoperable.

Use SEC CAL PWR (C24230), CBTFSP (C24102), BDELT (C24104),

CBDELT (C24103), or TURB PWR (C24101) for indication during power reduction.

Note The CRS typically directs performance of OP-901-501 to the STA. The CRS may direct this action. Determination of Tech Specs should not come before commencing the Rapid Power Reduction. The CRS should document the applicable Tech Specs during the power reduction.

3.1.3.5, SHUTDOWN CEA INSERTION LIMIT 3.1.3.1, CEA POSITION 3.2.3, AZIMUTHAL POWER TILT Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 8 of 20 Event

Description:

CEA 52 Drops into the core; Rapid Plant Power Reduction.

Time Position Applicants Actions or Behavior CRS Enter and direct the implementation of OP-901-212, Rapid Down Power.

CRS NOTE (1) A rapid power reduction is defined as approximately 30 MW/minute load reduction on the main turbine.

(2) Power Reduction may be stopped at any point.

(3) Some Steps of this procedure may not be applicable due to plant conditions. In these cases SM/CRS may NA the step.

(4) Steps within this procedure may be performed concurrently or out of sequence with SM/CRS concurrence.

ATC Begin RCS Boration by either of the following methods as directed by the CRS:

Direct Boration Emergency Boration using one Charging Pump Step 1: Begin RCS Boration by one of the following methods:

> 340 EFPD:

Direct Boration or Emergency Boration using one Charging Pump Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 9 of 20 Event

Description:

CEA 52 Drops into the core; Rapid Plant Power Reduction.

Time Position Applicants Actions or Behavior ATC Steps for Direct Boration:

CAUTION (1) This section affects reactivity. This evolution should be crosschecked and completed prior to leaving CP-4.

(2) At least one reactor coolant pump in each loop should be operating prior to performing direct boration operations to ensure proper chemical mixing.

6.7.1 Inform SM/CRS that this Section is being performed.

NOTE When performing a Plant down power where final RCS Boron Concentration needs to be determined, the following Plant Data Book figure(s) will assist the Operator in determining the required RCS Boron PPM change.

1.2.1.1 Power Defect Vs Power Level 1.4.3.1 Inverse Boron Worth Vs. Tmod at BOC (<30 EFPD) 1.4.4.1 Inverse Boron Worth Vs. Tmod at Peak Boron (30 EFPD up to 170 EFPD) 1.4.5.1 Inverse Boron Worth Vs. Tmod at MOC (170 EFPD up to 340 EFPD) 1.4.6.1 Inverse Boron Worth Vs. Tmod at EOC (>340 EFPD) 6.7.2 At SM/CRS discretion, calculate volume of Boric Acid to be added on Attachment 11.6, Calculation of Boric Acid Volume for Direct Boration or VCT Borate Makeup Mode.

Should use Reactor Engineering Reactivity Worksheet 6.7.3 Set Boric Acid Makeup Batch Counter to volume of Boric Acid desired.

6.7.4 Verify Boric Acid Makeup Pumps selector switch aligned to desired Boric Acid Makeup Pump A(B).

6.7.5 Place Direct Boration Valve, BAM-143, control switch to AUTO.

6.7.6 Place Makeup Mode selector switch to BORATE.

6.7.7 Verify selected Boric Acid Makeup Pump A(B) Starts.

6.7.8 Verify Direct Boration Valve, BAM-143, Opens.

NOTE The Boric Acid Flow Totalizer will not register below 3 GPM. The Boric Acid Flow Totalizer is most accurate in the range of 10 - 25 GPM.

6.7.9 If manual control of Boric Acid flow is desired, then perform the following:

6.7.9.1 Verify Boric Acid Flow controller, BAM-IFIC-0210Y, in Manual.

6.7.9.2 Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, output to >3 GPM flow rate.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 10 of 20 Event

Description:

CEA 52 Drops into the core; Rapid Plant Power Reduction.

Time Position Applicants Actions or Behavior 6.7.11 Verify Boric Acid Makeup Control Valve, BAM-141, Intermediate or Open.

6.7.12 Observe Boric Acid flow rate for proper indication.

Note This manipulation is performed at CP-4. The ATC should use the Reactivity Worksheet to recommend a boron quantity to the CRS.

ATC Perform Boron Equalization as follows:

Place available Pressurizer Pressure Backup Heater Control Switches to ON.

Reduce Pressurizer Spray Valve Controller (RC-IHIC-0100) setpoint potentiometer to establish spray flow and maintain RCS pressure 2250 PSIA (2175 - 2265).

This manipulation is performed at CP-2.

ATC Operate CEAs to maintain ASI using CEA Reg. Group 6 or Group P Control Element Assemblies.

Operate CEAs in Manual Group mode as follows:

6.7.1 Verify Plant Monitoring Computer operable in accordance with OP-004-012, Plant Monitoring Computer.

6.7.2 Position Group Select switch to desired group.

6.7.3 Place Mode Select switch to MG and verify the following:

White lights Illuminated on Group Selection Matrix for selected group MG light Illuminates 6.7.4 Operate CEA Manual Shim switch to WITHDRAW or INSERT group to desired height while monitoring the following:

CEA Position Indicator selected CEA group is moving in desired direction If Reactor is critical, then monitor the following:

Reactor Power Reactor Coolant System (RCS) temperature Axial Shape Index (ASI)

NOTE The Operator should remain in the area in front of the CEA Drive Mechanism Control Panel when the Mode Select switch is not in OFF.

6.7.5 When desired set of moves have been completed, then place Mode Select switch to OFF.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 11 of 20 Event

Description:

CEA 52 Drops into the core; Rapid Plant Power Reduction.

Time Position Applicants Actions or Behavior CEA Group P should be used first to a low limit of 120 inches, followed by CEA Group 6 to a low limit of 120 inches to comply with Tech Spec 3.1.3.6.

This manipulation is performed at CP-2.

CRS Notify the Load Dispatcher (Woodlands) that a rapid power reduction is in progress.

CRS Announce to Station Personnel over the Plant Paging System that a rapid plant power reduction is in progress.

Crew Maintain RCS Cold Leg Temperature 536°F to 549°F.

BOP Commence Turbine load reduction by performing the following:

Depress LOAD RATE MW/MIN pushbutton.

Set selected rate in Display Demand Window.

Depress ENTER pushbutton.

Depress REFERENCE pushbutton.

Set desired load in Reference Demand Window.

Depress ENTER pushbutton.

Depress GO pushbutton.

This manipulation is performed at CP-1. The BOP will set up the Main Turbine controls. The ATC will direct the BOP when to commence unloading the Main Turbine based on the drop in RCS Cold Leg temperature.

BOP Maintain S/G #1 level in manual during power reduction.

CRS When Reactor Power consistently indicates less than 98% power, as indicated on PMC PID C24631 [MAIN STEAM RAW POWER (MSBSRAW)],

or an alternate point provided by Reactor Engineering, then verify the value of C24648 [BSCAL SMOOTHING VAL. APPLD (DUMOUT17)] automatically changes to 1.

Examiner Note This event is complete when the desired power reduction has been accomplished Or As directed by the Lead Evaluator.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4/5/6 Page 12 of 20 Event

Description:

Loss of Coolant Accident; CS-125 A fails closed; Charging Pump A fails to auto-start; Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of RCS Leak.

Alarms Containment Water Leakage Hi (Cabinet N, L-20)

Containment Water Leakage Hi-Hi (Cabinet N, K-20)

Class 1E Rad Monitoring Sys Activity Hi-Hi (Cabinet SA, K-4)

Indications Lowering Pressurizer level.

Lowering Pressurizer pressure.

Backup Charging Pump AB auto-starts.

ATC If discovered at this point, the ATC may start Charging Pump A at this time.

ATC If directed by CRS, trip Reactor using 2 Reactor Trip pushbuttons at CP-2.

ATC If directed by CRS, initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS) at CP-7.

CRS Direct ATC and BOP to carry out Standard Post trip Actions.

ATC Determine Reactivity Control acceptance criteria are met:

Check reactor power is dropping.

Check startup rate is negative.

Check less than TWO CEAs are NOT fully inserted.

BOP Determine Maintenance of Vital Auxiliaries acceptance criteria are met:

Check the Main Turbine is tripped:

Governor valves closed Throttle valves closed Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4/5/6 Page 13 of 20 Event

Description:

Loss of Coolant Accident; CS-125 A fails closed; Charging Pump A fails to auto-start; Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start Time Position Applicants Actions or Behavior BOP Check the Main Generator is tripped:

GENERATOR BREAKER A tripped GENERATOR BREAKER B tripped EXCITER FIELD BREAKER tripped BOP Check station loads are energized from offsite electrical power as follows:

Train A A1, 6.9 KV non safety bus A2, 4.16 KV non safety bus A3, 4.16 KV safety bus A-DC electrical bus A or C vital AC Instrument Channel Train B B1, 6.9 KV non safety bus B2, 4.16 KV non safety bus B3, 4.16 KV safety bus B-DC electrical bus B or D vital AC Instrument Channel ATC Determine RCS Inventory Control acceptance criteria are met:

Check that the following conditions exist:

Pressurizer level is 7% to 60%

Pressurizer level is trending to 33% to 60%

Check RCS subcooling is greater than or equal to 28ºF.

ATC Determine RCS Pressure Control acceptance criteria are met by checking that BOTH of the following conditions exist:

Pressurizer pressure is 1750 psia to 2300 psia Pressurizer pressure is trending to 2125 psia to 2275 psia Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4/5/6 Page 14 of 20 Event

Description:

Loss of Coolant Accident; CS-125 A fails closed; Charging Pump A fails to auto-start; Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start Time Position Applicants Actions or Behavior IF pressurizer pressure is less than 1684 psia, THEN verify the following have initiated.

SIAS

CIAS BOP Start Low Pressure Safety Injection Pump A at CP-8.

ATC Start Charging Pump A if not already started.

ATC Determine Core Heat Removal acceptance criteria are met:

Check at least one RCP is operating.

Check operating loop T is less than 13ºF.

Check RCS subcooling is greater than or equal to 28ºF.

This could be N/A since all RCPs will be manually secured on Containment Spray actuation.

BOP Determine RCS Heat Removal acceptance criteria are met:

Check that at least one steam generator has BOTH of the following:

Steam generator level is 5% to 80% NR Main Feedwater is available to restore level within 50%-70% NR.

ATC Check RCS TC is 530ºF to 550ºF BOP Check steam generator pressure is 885 psia to 1040 psia.

BOP Check Feedwater Control in Reactor Trip Override:

MAIN FW REG valves are closed STARTUP FW REG valves are 13% to 21% open Operating main Feedwater pumps are 3800 rpm to 4000 rpm Note With MSIS in, MAIN FW REG valves and STARTUP FW REG valves will be closed. MFW Pumps will be coasting down.

BOP Reset moisture separator reheaters, and check the temperature control valves closed.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4/5/6 Page 15 of 20 Event

Description:

Loss of Coolant Accident; CS-125 A fails closed; Charging Pump A fails to auto-start; Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start Time Position Applicants Actions or Behavior ATC Determine Containment Isolation acceptance criteria are met:

Check containment pressure is less than 16.4 psia.

Check NO containment area radiation monitor alarms OR unexplained rise in activity.

Check NO steam plant activity monitor alarms OR unexplained rise in activity.

IF containment pressure is greater than or equal to 17.1 psia, THEN verify the following:

CIAS is initiated

SIAS is initiated

MSIS is initiated BOP Determine Containment Temperature and Pressure Control acceptance criteria are met:

Check containment temperature is less than or equal to 120ºF.

Check containment pressure is less than 16.4 psia.

IF containment pressure is greater than or equal to 17.7 psia, THEN verify ALL of the following:

CSAS is initiated

ALL available containment spray pumps are delivering flow greater than 1750 gpm

ALL RCPs are secured BOP After Containment Spray is initiated (CSAS), attempt to open CS-125 A.

Valve will not open when this action is taken.

Verify > 1750 GPM Containment Spray flow on Train B.

This action is taken at CP-8.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4/5/6 Page 16 of 20 Event

Description:

Loss of Coolant Accident; CS-125 A fails closed; Charging Pump A fails to auto-start; Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start Time Position Applicants Actions or Behavior Critical Task Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow.

ATC Following initiation of Containment Spray (auto or manual) secure all running Reactor Coolant Pumps as follows:

Place each RCP control switch to stop.

BOP Secure AH-12 A or B on CRS direction after initiation of SIAS at CP-18.

CRS After review of Standard Post Trip Actions, use Appendix 1, Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-002, Loss of Coolant Accident Recovery CRS 1. Confirm diagnosis of a LOCA :

a. Check Safety Function Status Check Acceptance criteria are satisfied.

b. IF Steam Generator sample path is available, THEN direct Chemistry to sample BOTH Steam Generators for activity.

Crew 2. Announce a Loss of Coolant Accident is in progress using the plant page.

CRS 3. Advise the Shift Manager to REFER TO EP-001-001, "Recognition &

Classification of Emergency Condition" and implement the Emergency Plan.

N/A 4. IF power has been interrupted to either 3A or 3B safety buses, THEN perform Appendix 20, "Operation of DCT Sump Pumps".

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4/5/6 Page 17 of 20 Event

Description:

Loss of Coolant Accident; CS-125 A fails closed; Charging Pump A fails to auto-start; Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start Time Position Applicants Actions or Behavior CRS 5. REFER TO Section 6.0, "Placekeeper" and record the time of the reactor trip.

Note The CRS will typically perform a brief at this point in the EOP.

CRS/ATC 6. IF pressurizer pressure is less than 1684 psia, THEN check SIAS has initiated.

CRS/BOP 7. IF SIAS has initiated, THEN:

a. Verify safety injection pumps have started.

b. Check safety injection flow is within the following:

Appendix 2-E, "HPSI Flow Curve"

Appendix 2-F, "LPSI Flow Curve"

c. Verify ALL available charging pumps are operating.

CRS 19. Cooldown the RCS to less than 350 ºF TH or CET temperature using the steam bypass control valves.

The CRS should discuss this step during the brief. It is not necessary for the crew to begin this cooldown before moving to the last malfunction.

Crew When Containment Temperature rises above 200 F, update crew on need to use bracketed parameters due to harsh environment in Containment.

CRS During brief in OP-902-002, should discuss necessary strategy of using Steam Generators to cool RCS.

Examiner Note This event is complete after entry into OP-902-002. It is not necessary to allow a brief to occur at this point. SIAS initiation verification and CSAS verification should occur before moving forward. ATC should have secured RCPs and the BOP should have started LPSI Pump A.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 18 of 20 Event

Description:

Containment Spray Pump B trip; Safety Function Recovery; Alignment of LPSI Pump B to replace CS Pump B Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of Containment Spray Pump B trip.

Alarms CNTMT Spray Pump B Unavailable (Cabinet N, A-14)

CNTMT Spray Pump B Trip/Trouble (Cabinet N, B-14)

Indications Amber light on Containment Spray Pump B control switch.

No Containment Spray flow indicated on CS-IFI-7122 B.

CRS Recognize the Containment Temperature and Pressure Control safety function is not met. Exit OP-902-002 and enter OP-902-008, Functional Recovery procedure.

BOP Place Hydrogen Analyzers in service as follows:

Train A o Place Train A H2 ANALYZER CNTMT ISOL VALVE keyswitch to OPEN.

o Place H2 ANALYZER A POWER to ON.

o Check H2 ANALYZER A Pumps indicate ON.

Train B o Place Train B H2 ANALYZER CNTMT ISOL VALVE keyswitch to OPEN.

o Place H2 ANALYZER B POWER to ON.

o Check H2 ANALYZER B Pumps indicate ON.

CRS Identify success paths to be used and prioritize Safety Functions.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 19 of 20 Event

Description:

Containment Spray Pump B trip; Safety Function Recovery; Alignment of LPSI Pump B to replace CS Pump B Time Position Applicants Actions or Behavior Proper prioritization will result in Containment Isolation being priority 1 and Containment Temperature and Pressure Control being priority 2.

The CRS may request plant data from the ATC and BOP operators during his prioritization. The CRS may ask for the BOP to verify Safety Injection Pumps are meeting the flow curves of OP-902-009, Appendix

2. Both High and Low Pressure Safety Injection Pumps are meeting their flow curves.

The CRS should address Containment Isolation by overriding CS-125 B closed. OP-902-008, section CI-1 will direct this to be accomplished in accordance with OP-902-009, Standard Appendices, Attachment 21-A.

This requires the RAB Watch to come to the Control Room and obtain a key from the Shift Managers office.

The CRS should address Containment Temperature and Pressure Control by aligning Low Pressure Safety Injection Pump B to replace the failed Containment Spray Pump B. OP-902-008, section CTPC, Continuing Actions will direct this to be accomplished in accordance with OP-902-009, Standard Appendices, Attachment 28.

These actions should be pursued in parallel. The CRS may choose to prepare for, but not close, CS-125 B, pending attempts to accomplish aligning LPSI Pump B. Establishing flow through CS-125 B with LPSI Pump B will remove the requirement to close CS-125 B.

Critical Task Establish Containment temperature and pressure control.

This task is satisfied by aligning LPSI Pump B to replace CS Pump B prior to exiting the Containment Temperature and Pressure Control safety function in OP-902-008.

BOP Align LPSI Pump B to replace CS Pump B as follows:

Verify LPSI Pump B control switch in OFF.

Verify Containment Spray Pump B control switch in OFF.

Place SI-129 B, LPSI FLOW CONTROL VALVE to AUTO.

Place SI-IFIC-0306, LPSI FLOW CONTROLLERS HEADER 1A/1B in MAN.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 20 of 20 Event

Description:

Containment Spray Pump B trip; Safety Function Recovery; Alignment of LPSI Pump B to replace CS Pump B Time Position Applicants Actions or Behavior Adjust SI-IFIC-0306, LPSI FLOW CONTROLLERS HEADER 1A/1B to 0% output.

Verify the following valves Closed:

o SI-415 B, LPSI SHUTDOWN TEMP CONTROL valve.

o SI-138 B, LPSI FLOW CONTROL COLD LEG 1B.

Control switch must be taken to Open and then to Close to shut valve.

o SI-139 B, LPSI FLOW CONTROL COLD LEG 1A.

Control switch must be taken to Open and then to Close to shut valve.

Open SI-125B/SI-412B, SHDN HX B ISOL valves.

Valve stroke time is ~60 seconds.

Verify CS-125B, CNTMT SPRAY HEADER ISOL valve open.

Start LPSI Pump B.

Examiner Note This event is complete after aligning LPSI Pump B to replace CS Pump B Or As directed by the Lead Evaluator.

Scenario 1, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 1 of 20 Event

Description:

Pressurizer pressure instrument RC-IPR-0100 X fails low Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed pressure instrument.

Alarms PRESSURIZER PRESSURE HI/LO (Cabinet H, E-1)

PRESSURIZER PRESS SIGNAL DEVIATION (Cabinet H, F-1)

Indications Recorder RC-IPR-0100 red pen fails low.

Controller RC-IPIC-0100 process fails low.

Controller RC-IPIC-0100 output goes to 0%.

Both Pressurizer Main Spray Valves go closed.

Note Pressurizer boron equalization will be set up for the power ascension when this failure occurs. The Pressurizer Heaters will already be energized. After the spray valves go closed, RCS pressure will rise until all Heaters secure.

CRS Enter and direct the implementation of OP-901-120, Pressurizer Pressure Malfunction, and use sub-section E1, Pressurizer Pressure Control Channel Instrument Failure.

CRS Procedure Caution Steam Generator pressures dropping concurrently with dropping Pressurizer level may be indicative of an excess steam demand.

CRS 1. IF Pressurizer Pressure and Level are dropping concurrently, OR RCS leakage is otherwise indicated, THEN GO TO OP-901-111, Reactor Coolant System Leak.

CRS 2. IF PRESSURIZER PRESSURE CHANNEL X/Y recorder (RC-IPR-0100) indicates a Pressurizer Pressure Control Channel instrument has failed, THEN GO TO Subsection E1, Pressurizer Pressure Control Channel Instrument Failure.

ATC 1. Verify control channel instrument failure by checking PRESSURIZER PRESSURE CHANNEL X/Y recorder (RC-IPR-0100).

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 2 of 20 Event

Description:

Pressurizer pressure instrument RC-IPR-0100 X fails low Time Position Applicants Actions or Behavior ATC 2. Transfer Pressurizer pressure control to operable channel using Pressurizer Pressure Channel Selector control switch.

Position Y should be selected.

ATC 3. IF Pressurizer Pressure control channel is failed high, THEN perform the following:

Pressurizer pressure is failed low.

Note The pressurizer Proportional Heaters will secure on high RCS pressure.

The CRS may direct the ATC to reset these Heaters.

ATC 4. Verify proper operation of Pressurizer Pressure controller (RC-IPIC-0100)

AND Pressurizer Pressure controlling OR being restored to 2250 PSIA.

CRS Refer to Technical Specification 3.2.8.

Entry required if RCS pressure drops below 2125 PSIA or exceeds 2275 PSIA. Neither condition should be met.

ATC Place Pressurizer Spray Controller RC-IHIC-0100 to Auto.

The CRS should direct this after Channel Y has been selected for Pressurizer pressure control.

Examiner Note This event is complete when Pressurizer Pressure Control has been transferred to Channel Y.

Or As directed by the Lead Evaluator.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 3 of 20 Event

Description:

RCP 1A speed instrument failure, Channel B, Core Protection Calculator B trip Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

RPS CHANNEL TRIP LOCAL PWR DENSITY HI (Cabinet K, A-11)

RPS CHANNEL TRIP DNBR LO (Cabinet K, A-12)

LOCAL PWR DENSITY HI PRETRIP A/C (Cabinet K, C-11)

DNBR LO PRETRIP A/C (Cabinet K, C-12)

RPS CHANNEL B TROUBLE (Cabinet K, F-18)

Indications Trips and pre-trips indicated on Core Protection Calculator B for LPD and DNBR.

CPC Channel B point 415 reads 32768.

ATC/BOP Use CPC Channel B and OP-004-006, Core Protection Calculator System, to decode the trip buffer and determine that RCP 1B speed sensor is the source of the problem.

Steps from OP-004-006:

NOTE Sensor Failure Status Words are stored in each CPC as a troubleshooting aid. Sensor Failure Status Words are PIDs 415, 416, 417 and 418.

6.3.1 If CPC Auxiliary Trip has occurred due to Sensor Failure, then refer to Technical Specification 3.3.1.

6.3.2 Check that the Sensor Failure has Cleared by observing Sens Fail light is extinguished.

6.3.3 If CPC Sensor Failure has caused a Channel Trip, then refer to Technical Specification 3.3.1.

6.3.4 Decode the following Sensor Failure Status Words in accordance with Attachment 11.7, Sensor Failure Status Words:

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 4 of 20 Event

Description:

RCP 1A speed instrument failure, Channel B, Core Protection Calculator B trip Time Position Applicants Actions or Behavior From Attachment 11.7:

Note CPC Point IDs 415, 416, 417, and 418 are used to store Sensor failure events. When Sensor is out of range, a corresponding Sensor Status Word bit is set. The Sensor Status Words store information about the following inputs:

Point 415 Pump Speed Sensors Point 416 Temperature, pressure, and excore detectors, Sensors, and CEACs Point 417 Subgroup 1 - 16 position Sensors Point 418 Subgroup 17 - 22 position Sensors The stored values must be converted to binary in order to decode the Sensor Status Words using the following table:

Value: 32768 BIT: 0 415, bit 0, Coolant Pump Speed 1 CRS Review Tech Specs based on the failed instrument.

Enter Tech Spec 3.3.1 Direct bypassing Channel B bistables 3, and 4 for Low DNBR and Hi Local Power. This is a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action.

Note This instrument is an input to Core Protection Calculator B (CPC B), which is why bistables 3 and 4 need to be bypassed..

BOP Bypass Channel B bistables 3, and 4 for Low DNBR and Hi local Power Located on CP-10 B (rear panel).

BOP CAUTION (1) ATTEMPTING TO PLACE MORE THAN ONE TRIP CHANNEL IN BYPASS REMOVES BOTH TRIP CHANNELS FROM BYPASS.

(2) PRIOR TO PLACING ANY TRIP CHANNEL IN BYPASS, VERIFY BYPASS PUSH BUTTONS ON DE-ENERGIZED PPS BAY NOT DEPRESSED.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 5 of 20 Event

Description:

RCP 1A speed instrument failure, Channel B, Core Protection Calculator B trip Time Position Applicants Actions or Behavior BOP 6.2.1 To Bypass a Trip Channel, perform the following:

6.2.1.1 Verify desired Trip Channel is not Bypassed on another PPS Channel.

BOP 6.2.1.2 Open key-locked portion of BCP in desired PPS Channel.

BOP 6.2.1.3 Depress Bypass push button for desired Trip Channel.

BOP 6.2.1.4 Verify Bypass push button remains in a Depressed state.

BOP 6.2.1.5 Verify Bypass light Illuminates on BCP and ROM for the desired Trip Channel.

Examiner Note This event is complete when the proper bistables have been bypassed Or As directed by the Lead Evaluator Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 6 of 20 Event

Description:

Letdown Back Pressure Controller Setpoint Fails to 100%

Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed setpoint.

Alarms:

Letdown HX Outlet Pressure Hi (Cabinet G, A-2)

Letdown Flow Hi/Lo (Cabinet G, C-1)

Indications Letdown flow goes to 0 gpm In service Back Pressure Control Valve goes shut with 0% output CRS Enter and direct the implementation of OP-901-112, Charging or Letdown Malfunction, and use sub-section E2, Letdown Malfunction.

ATC 1. IF necessary, THEN maintain Pressurizer level by placing LETDOWN FLOW CONTROL VALVES controller (RC-IHIC-0110) in MAN, and control manually.

The crew may place the letdown Flow controller in manual at this time but they will not be able to control Pressurizer level until the Back Pressure failure is addressed. There is another step later that also addresses the Letdown Flow controller.

Crew NOTE If all Charging Pumps are secured, then LETDOWN STOP VALVE (CVC 101) will close on high REGEN HX TUBE OUTLET temperature if RCS is 470°F.

Crew 2. Operate Charging Pumps as necessary to maintain Pressurizer level in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

This should not be applicable. If the crew delays taking action due to the failure, Pressurizer level will rise and they may choose to take action.

N/A 3. IF Pressurizer level falls below the minimum level for operation of Attachment 1, THEN perform the following:

3.1 Trip the Reactor.

3.2 Manually initiate Safety Injection Actuation.

3.3 Go to OP-902-000, STANDARD POST TRIP ACTIONS.

This should not be applicable.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 7 of 20 Event

Description:

Letdown Back Pressure Controller Setpoint Fails to 100%

Time Position Applicants Actions or Behavior N/A 4. IF a leak exists in Letdown System, THEN attempt to locate AND isolate leak.

This is not applicable.

N/A 5. IF leak has been isolated, THEN re-establish Letdown in accordance with OP-002-005, CHEMICAL AND VOLUME CONTROL.

CRS 6. IF the in service Letdown Flow Control valve (CVC 113A) OR (CVC 113B) is NOT controlling, THEN place standby Letdown Flow Control valve in service as follows:

This is not failed.

ATC 7. IF Letdown Backpressure Controller (CVC-IPIC-0201) is NOT operating properly, THEN perform the following:

7.1 Control Letdown flow with Letdown Flow Controller (RC-IHIC-0110) in MAN.

ATC 7.2 Control Letdown backpressure with Letdown Backpressure Controller (CVC-IPIC-0201) in MAN.

CRS 8. IF the in service Letdown Backpressure Control valve (CVC 123A) OR (CVC 123B) is NOT operating properly, THEN place standby Letdown Backpressure Control valve in service as follows:

This is not failed.

CRS 9. Locally monitor the following differential pressure indications:

Purification Ion Exchanger A DP (CVC-IDPI-0207)

Purification Ion Exchanger B DP (CVC-IDPI-0205)

Deborating Ion Exchanger DP (CVC-IDPI-0203)

Purification Filter DP (CVC-IDPI-0202)

Letdown Strainer DP (CVC-IDPI-0204)

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 8 of 20 Event

Description:

Letdown Back Pressure Controller Setpoint Fails to 100%

Time Position Applicants Actions or Behavior CRS 10. IF a high ion exchanger OR filter differential pressure is indicated, THEN remove affected component from service in accordance with OP-002-005, CHEMICAL AND VOLUME CONTROL.

ATC 11. IF the LETDOWN FLOW CONTROL VALVES controller (RC-IHIC-0110) has been placed in manual to control Pressurizer level, THEN match controller process with its output and place in AUTO.

The CRS may choose to leave this controller in manual until the repairs are completed on the Back Pressure control. It is acceptable if the CRS places the controller to Auto at this point.

Examiner Note This event is complete when the Letdown system flow has been restored and controlled in manual Or As directed by the Lead Evaluator Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 9 of 20 Event

Description:

Dry Cooling Tower Fan 8B failure Time Position Applicants Actions or Behavior This failure is called in by the Outside Watch for the purpose of evaluating Tech Spec 3.7.4. The simulator operator will call the CRS on cue from the Lead Examiner.

CRS On receiving information regarding DCT Fan 8B, evaluates Tech Spec 3.7.4.

The CRS may direct the BOP to place DCT Fan 8B control switch to OFF.

Crew A member of the crew may determine that DCT Fan 8B is under the missile shield. This determination is not critical for this situation because of outside air temperature.

CRS Determine that with the atmospheric and ambient conditions present, both Ultimate Heat Sinks are operable, but entry into Tech Spec 3.7.4.d is still required. Since both Ultimate Heat Sinks are still operable, Tech Spec 3.8.1.1 is still being complied with.

The action is a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months action. The CRS should assign an action to the BOP to use a Tech Spec/TRM Addendum sheet and record ambient conditions every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Examiner Note This event is complete when the CRS has completed addressing Tech Specs Or As directed by the Lead Evaluator Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 10 of 20 Event

Description:

Inadvertent Containment Spray Actuation Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of a Containment Spray Actuation.

Alarms Numerous alarms on Panels H, SA, and SB for RCP low CCW flow Numerous alarms on Panel M for Containment Spray Pump A and Train A isolation valve Numerous alarms on Panel M for Containment Spray Pump B and Train B isolation valve Indications Containment Spray Pumps A and B running.

CS-125 A and B open.

Containment Spray Header flow on Trains A and B.

CC-710, CC-713, and CC-641 indicate closed at CP-8.

CRS Enter and direct the implementation of OP-901-504, Inadvertent ESFAS Actuation.

CRS IF an inadvertent CSAS occurs, THEN perform the following:

IF a Reactor Trip occurs, THEN GO TO OP-902-000, STANDARD POST TRIP ACTIONS AND perform concurrently with this procedure.

GO TO Subsection E2, Inadvertent CSAS.

BOP Secure BOTH Containment Spray Pumps by placing each control switch to OFF.

CRS Procedure caution If Component Cooling Water is lost to Reactor Coolant Pump seals for >10 minutes, then restoring Component Cooling Water to Reactor Coolant Pumps may result in seal failure.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 11 of 20 Event

Description:

Inadvertent Containment Spray Actuation Time Position Applicants Actions or Behavior BOP 2.1 Within 3 minutes restore CCW flow to Reactor Coolant Pumps as follows:

Open the following valves:

CC 710 RCP OUTLET INSIDE ISOL.

CC 641 RCP INLET OUTSIDE ISOL CC 713 RCP OUTLET OUTSIDE ISOL ATC NOTE Manual override is accomplished by positioning control switch to CLOSED, then to OPEN. If after 100 seconds CCW Return temperature is NOT less than 145 °F, Seal Cooler will isolate.

ATC 2.2 Verify the following Reactor Coolant Pump CCW Isolation valves Open:

CC-679A(B) and CC-680A(B), RCP Seal Cooler CCW Outlet Isolations, and CC-6651A(B) and CC-666A(B), RCP Seal Cooler CCW Inlet Isolations ATC 2.3 Verify Open CC 200A & CC 727, SUCT & DISCH HEADER TIE VALVES A TO AB.

CRS IF in Mode 1 OR 2 AND CCW flow can NOT be restored to Reactor Coolant Pumps within 3 minutes, THEN perform the following:

Trip the Reactor.

Stop the affected Reactor Coolant Pumps.

GO TO OP-902-000, STANDARD POST TRIP ACTIONS.

Examiner Note This event is not faulted and the crew will be capable of restoring CCW flow to the RCPs. The crew may, however, choose to trip the reactor if the 3 minute limit is challenged.

ATC Trip the reactor on direction from the CRS.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 12 of 20 Event

Description:

Inadvertent Containment Spray Actuation Time Position Applicants Actions or Behavior Critical Task Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow.

ATC Following tripping of the reactor, secure all running Reactor Coolant Pumps as follows:

Place each RCP control switch to stop.

Note This is done at CP-2. At this point, the RCPs will be operating with no CCW flow since the actuation of Containment Spray.

This condition is only allowed for 3 minutes.

CRS Direct ATC and BOP to carry out Standard Post trip Actions.

Standard Post Trip Actions are listed in the next event.

Examiner Note This event is complete after Containment Spray Pumps are secured and CCW has been re-aligned to the RCPs. If the crew chooses to trip the reactor and secure RCPs, then insert the next malfunction during post trip actions.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 13 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications.

Alarms Containment Water Leakage Hi (Cabinet N, L-20)

Containment Water Leakage Hi-Hi (Cabinet N, K-20)

Containment Pressure Hi/Lo (Cabinet M, H-4 and N, H-14)

Containment Fan Cooler B Disch Air Temp Hi (Cabinet SB, B-6)

Containment Fan Cooler D Disch Air Temp Hi (Cabinet SB, C-6)

Indications Rising Containment Pressure Lowering level Steam Generator #2 Note The following steps are applicable after the reactor is tripped and the crew is performing Standard Post Trip Actions.

ATC Determine Reactivity Control acceptance criteria are met:

Check reactor power is dropping.

Check startup rate is negative.

Check less than TWO CEAs are NOT fully inserted.

BOP Determine Maintenance of Vital Auxiliaries acceptance criteria are met:

Check the Main Turbine is tripped:

Governor valves closed Throttle valves closed BOP Check the Main Generator is tripped:

GENERATOR BREAKER A tripped GENERATOR BREAKER B tripped EXCITER FIELD BREAKER tripped Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 14 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior BOP Check station loads are energized from offsite electrical power as follows:

Train A A1, 6.9 KV non safety bus A2, 4.16 KV non safety bus A3, 4.16 KV safety bus A-DC electrical bus A or C vital AC Instrument Channel Train B B1, 6.9 KV non safety bus B2, 4.16 KV non safety bus B3, 4.16 KV safety bus B-DC electrical bus B or D vital AC Instrument Channel ATC Determine RCS Inventory Control acceptance criteria are met:

Check that the following conditions exist:

Pressurizer level is 7% to 60%

Pressurizer level is trending to 33% to 60%

Check RCS subcooling is greater than or equal to 28ºF.

Note This safety function may or may not be met, depending on the speed the crew is working Standard Post Trip Actions. Either way, there are no contingencies necessary for this step.

ATC Determine RCS Pressure Control acceptance criteria are met by checking that BOTH of the following conditions exist:

Pressurizer pressure is 1750 psia to 2300 psia Pressurizer pressure is trending to 2125 psia to 2275 psia Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 15 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior CRS IF pressurizer pressure is less than 1684 psia, THEN verify the following have initiated.

SIAS

CIAS ATC If directed by CRS, initiate Safety Injection Actuation (SIAS), Main Steam Isolation (MSIS) and Containment Isolation Actuation (CIAS) at CP-7.

ATC Close CVC-183 at CP-4 Note This valve malfunctions and fails to auto close. It will close when the control switch is taken to close.

ATC Open BAM-113 A at CP-4 Note This valve malfunctions and fails to auto open. It will open when the control switch is taken to open.

ATC IF pressurizer pressure is less than 1621 psia, THEN verify no more than two RCPs are operating ATC Determine Core Heat Removal acceptance criteria are met:

Check at least one RCP is operating.

Check operating loop T is less than 13ºF.

Check RCS subcooling is greater than or equal to 28ºF.

Note Depending on the speed the crew is working, the ATC may have already secured all RCPs due to Containment Spray initiation. If all RCPs are OFF, then the CRS is permitted to skip this step.

BOP Determine RCS Heat Removal acceptance criteria are met:

Check that at least one steam generator has BOTH of the following:

Steam generator level is 5% to 80% NR Main Feedwater is available to restore level within 50%-70% NR.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 16 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior Note The contingency for this step, since there will be a Main Steam Isolation Signal, is to verify Emergency Feedwater is available. This does not require a manual initiation of EFAS.

ATC Check RCS TC is 530 ºF to 550 ºF CRS IF RCS TC is less than 530 ºF, THEN perform the following:

IF RCS TC is being controlled by an ESD, THEN REFER TO Appendix 13, "Stabilize RCS Temperature" and stabilize RCS temperature using the least affected steam generator.

Note Appendix 13 directs steps to address PTS after Representative CET temperature and Pressurizer pressure have both started to rise. The steps are also contained in the excess steam demand recovery procedure.

BOP Check steam generator pressure is 885 psia to 1040 psia.

BOP IF steam generator pressure is less than 885 psia, THEN perform ALL of the following:

1) Verify steam bypass valves are closed.

2) Verify ADVs are closed.

IF steam generator pressure is less than or equal to 666 psia, THEN verify MSIS is initiated.

BOP Check Feedwater Control in Reactor Trip Override:

MAIN FW REG valves are closed STARTUP FW REG valves are 13% to 21% open Operating main Feedwater pumps are 3800 rpm to 4000 rpm Note With a MSIS, MAIN FW REG valves and STARTUP FW REG valves will be closed. Both MFW Pumps will be coasting down.

BOP Reset moisture separator reheaters, and check the temperature control valves closed.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 17 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior ATC Determine Containment Isolation acceptance criteria are met:

Check containment pressure is less than 16.4 psia.

Check NO containment area radiation monitor alarms OR unexplained rise in activity.

Check NO steam plant activity monitor alarms OR unexplained rise in activity.

IF containment pressure is greater than or equal to 17.1 psia, THEN verify the following:

CIAS is initiated

SIAS is initiated

MSIS is initiated BOP Determine Containment Temperature and Pressure Control acceptance criteria are met:

Check containment temperature is less than or equal to 120ºF.

Check containment pressure is less than 16.4 psia.

IF containment pressure is greater than or equal to 17.7 psia, THEN verify ALL of the following:

CSAS is initiated

ALL available containment spray pumps are delivering flow greater than 1750 gpm

ALL RCPs are secured Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 18 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior Critical Task Establish Containment Temperature and Pressure control.

This task is satisfied by starting at least 1 Containment Spray Pump before the review of OP-902-000, Standard Post Trip Actions.

BOP Start Containment Spray Pumps A and B.

Critical Task Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of CSAS initiation.

ATC Following initiation of CSAS secure all running Reactor Coolant Pumps as follows:

Place each RCP control switch to stop at CP-2 BOP Secure AH-12 A or B on CRS direction after initiation of SIAS at CP-18.

CRS After review of Standard Post Trip Actions, use Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-004, Excess Steam Demand Recovery Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 19 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior CRS After Excess Steam Demand is identified, direct ATC and BOP to monitor for the trigger points for the need to stabilize RCS temperature.

Critical parameters are Pressurizer pressure rising and RCS Representative CET temperature rising.

Steps for stabilizing RCS temperature following an excess steam demand are contained in 2 procedures.

Appendix 13 is used if the critical parameters are both rising before the CRS has entered OP-902-004, Excess Steam Demand Recovery.

Step 16 of OP-902-004 is used if both parameters start rising after the crew has entered OP-902-004.

Critical Task Establish RCS temperature control This task is satisfied by taking action to stabilize RCS temperature within the limits of the RCS P/T curve using ADV #1 and establishing EFW flow to Steam Generator #1. Action to address this task should commence prior to RCS temperature exceeding 550 °F.

BOP When directed by the CRS to take action to stabilize RCS temperature:

Place the ADV for Steam Generator #1 to manual and fully open the ADV #1.

Manually initiate EFAS for Steam Generator #1.

Place EFW Flow Control Valve to manual and commence feeding Steam Generator #1.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 6/7 Page 20 of 20 Event

Description:

Main Steam line break inside Containment; Initiate Containment Spray flow Relay K301 failure, BAM-113 A and CVC-183 fail to position on Safety Injection Time Position Applicants Actions or Behavior Critical Task Establish RCS pressure control This task is satisfied by taking action to stabilize RCS pressure within the limits of the RCS P/T curve and additionally maintain RCS pressure within 1500-1600 psid of the faulted steam generator. Action to address this task should commence prior to RCS pressure exceeding 2250 PSIA.

ATC When directed by the CRS to take action to stabilize RCS temperature:

IF RCS pressure is 1500 psia, THEN stabilize RCS pressure at a value not to exceed 1600 psid between the RCS and the lowest SG pressure.

IF RCS pressure is < 1500 psia, THEN stabilize RCS pressure at >

HPSI shutoff head (1500-1600 psia).

Examiner Note This event is complete after RCS temperature and pressure have been stabilized Or As directed by the Lead Evaluator.

Scenario 2, Revision 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 1 Page 1 of 17 Event

Description:

Steam Generator #1 level instrument SG-ILI-1113 D fails high.

Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

RPS CHANNEL TRIP SG 1 LEVEL HI (Cabinet K, E-11)

SG 1 LEVEL HI PRETRIP B/D (Cabinet K, G-11)

RPS CHANNEL D TROUBLE (Cabinet K, H-18)

Indications SG-ILI-1113 D indicates 100% narrow Range on CP-8 Hi SG-1 Level trip and pre-trip on Channel D CRS Review Tech Specs based on the failed instrument.

Enter Tech Spec 3.3.1 (Table 3.3-1 item 8), 3.3.2.b (Table 3.3-3 item 7), and TRM 3.3.1 (Table 3.3-1 item 1).

Direct bypassing Channel D bistables 7, 9, and 19 for Steam Generator Level Low, Steam Generator Level High, and Steam Generator Differential Pressure for Steam Generator #1. This is a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action.

Note This instrument is listed as an accident monitoring instrument for Tech Spec 3.3.3.6, but the minimum channels for operability are met. Entry is not required.

BOP Bypass Channel D bistables 7, 9, and 19 for Steam Generator Level Low, Steam Generator Level High, and Steam Generator Differential Pressure for Steam Generator #1.

Located on CP-10 C (rear panel).

Examiner Note This event is complete when the proper bistables have been bypassed Or As directed by the Lead Evaluator Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 2/3 Page 2 of 17 Event

Description:

Main Feedwater Pump A trips, Reactor Power Cutback, Regulating Group 4 CEAs fail to insert in automatic following Reactor Power Cutback Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of the tripped Main Feedwater Pump A.

Alarms FWPT A Trip Overspeed (Cabinet F, K-15)

REACTOR PWR CUTBACK ACTUATION (Cabinet E, F-9)

COLSS MASTER (Cabinet L, A-6)

Indications Reactor power dropping Regulating Groups 5 & 6 CEA rod bottom lights AND lower electrical limit lights illuminated TAVE dropping Generator Output MW meter indicates power dropping Runback Oper light illuminated on DEH Control Panel CRS Enter and direct the implementation of OP-901-101, Reactor Power Cutback ATC 1. Place Control Element Drive Mechanism Mode Select switch to AS.

2. Verify selected subgroups dropped.

CRS 1. IF Reactor Power Cutback has NOT occurred OR the correct rod pattern has not dropped, THEN trip the Reactor AND go to OP-902-000, Standard Post Trip Actions.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 2/3 Page 3 of 17 Event

Description:

Main Feedwater Pump A trips, Reactor Power Cutback, Regulating Group 4 CEAs fail to insert in automatic following Reactor Power Cutback Time Position Applicants Actions or Behavior CRS 2. IF BOTH Main Feedwater Pumps are tripped, THEN perform the following:

2.1 Trip the Reactor.

2.2 GO TO OP-902-000, Standard Post Trip Actions.

CRS 3. IF Reactor Power Cutback was due to a Main Turbine trip due to loss of Main Turbine Lube Oil OR Main Turbine High Vibration, THEN perform the following:

3.1 Trip the Reactor.

3.2 GO TO OP-902-000, Standard Post Trip Actions AND perform OP-901-210, TURBINE TRIP, concurrently with OP-902-000.

CRS 4. IF Main Turbine has tripped due to reasons other than stated in step 3, THEN perform OP-901-210, TURBINE TRIP, concurrently with this procedure.

CRS 5. IF Reactor Power Cutback was due to GOBs opening AND the Main Generator is supplying plant loads, THEN perform OP-901-211, GENERATOR MALFUNCTION, concurrently with this procedure.

CRS 6. IF ONE Main Feedwater Pump is tripped, THEN verify Main Turbine has setback to <50% (<610 MW).

BOP 7. Verify Feedwater Control System maintaining Steam Generator levels.

ATC 8. Verify Pressurizer Pressure Control System maintaining OR restoring Pressurizer pressure to 2250 psia.

ATC 9. Verify Pressurizer Level Control System maintaining OR restoring Pressurizer level to program level.

BOP 10. Verify Steam Bypass Control System responding to maintain Steam Generator pressure.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 2/3 Page 4 of 17 Event

Description:

Main Feedwater Pump A trips, Reactor Power Cutback, Regulating Group 4 CEAs fail to insert in automatic following Reactor Power Cutback Time Position Applicants Actions or Behavior Crew 11. Using the Plant Paging system, announce the following twice:

"Attention station personnel, attention station personnel, a Reactor Power Cutback has occurred."

ATC 12. WHEN STEAM BYPASS AUTO MOTION INHIBIT (Cabinet H, K-6) annunciator is received, THEN place CEA Drive Mechanism Mode Select switch to OFF.

CRS Procedure Note Following a Reactor Power Cutback ASI will move to the top of the core.

Performing ASI control using Group P to move ASI more toward the center of the core will assist in meeting POLs.

CRS 13. Review Technical Specification POLs.

The CRS should enter Tech Spec 3.2.4 for DNBR and 3.2.7 for ASI.

DNBR requires corrective action be taken within 15 minutes. The previous note directs performing ASI control with Group P, which will also assis in meeting the POL for DNBR.

Note The failure of Regulating Group 4 to insert is what drives ASI out of spec and the need to insert Regulating Group P.

Note The CRS may choose to enter OP-901-102, CEA or CEDMCS Malfunction From OP-901-102, CEA or CEDMCS Malfunction Place CEDMCS Mode Select switch to OFF.

If failure of controlling group to move in automatic occurs, then go to section E4, Failure of Controlling Group to Move in Automatic.

If Plant up-power or down-power is in progress, then stop load changes and Reactor Coolant System boron concentration changes.

Match TAVG and TREF using CEAs in Manual Sequential or Manual Group.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 2/3 Page 5 of 17 Event

Description:

Main Feedwater Pump A trips, Reactor Power Cutback, Regulating Group 4 CEAs fail to insert in automatic following Reactor Power Cutback Time Position Applicants Actions or Behavior From OP-004-004, Control Element Drive ATC Operate CEAs to maintain ASI using CEA Reg. Group P Control Element Assemblies.

Operate CEAs in Manual Group mode as follows:

6.7.1 Verify Plant Monitoring Computer operable in accordance with OP-004-012, Plant Monitoring Computer.

6.7.2 Position Group Select switch to desired group.

6.7.3 Place Mode Select switch to MG and verify the following:

White lights Illuminated on Group Selection Matrix for selected group MG light Illuminates 6.7.4 Operate CEA Manual Shim switch to WITHDRAW or INSERT group to desired height while monitoring the following:

CEA Position Indicator selected CEA group is moving in desired direction If Reactor is critical, then monitor the following:

Reactor Power Reactor Coolant System (RCS) temperature Axial Shape Index (ASI)

NOTE The Operator should remain in the area in front of the CEA Drive Mechanism Control Panel when the Mode Select switch is not in OFF.

6.7.5 When desired set of moves have been completed, then place Mode Select switch to OFF.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 2/3 Page 6 of 17 Event

Description:

Main Feedwater Pump A trips, Reactor Power Cutback, Regulating Group 4 CEAs fail to insert in automatic following Reactor Power Cutback Time Position Applicants Actions or Behavior CRS Procedure Note Resetting Reactor Power Cutback will block Turbine Setback and Runback signals.

ATC 14. Reset Reactor Power Cutback as follows:

14.1 On CP-2, Reactor Control, depress AUTO ACTUATE OUT OF SERVICE pushbutton AND verify pushbutton illuminates.

14.2 On CP-2, Reactor Control, depress TEST RESET pushbutton AND verify pushbutton illuminates.

Examiner Note This event is complete after ASI with group P has commenced Or As directed by the Lead Evaluator Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 4 Page 7 of 17 Event

Description:

Turbine Cooling Water Pump A trips, Turbine Cooling Water Pump B fails to auto start Time Position Applicants Actions or Behavior BOP Recognize and report indications of tripped TCW Pump A.

Alarms TURBINE CLNG WATER DISCH HDR PRESS LO (Cabinet E, E-9)

TURB CLNG WTR PUMP A TRIP/TROUBLE (Cabinet E, F-9)

HYDROGEN TEMPERATURE HI (Cabinet D, H-8)

Indications Rising temperature on components cooled by TCW TCW Pump A indicates tripped TCW Pump B not running CRS Enter and direct the implementation of OP-901-512, Loss of Turbine Cooling Water CRS 1. IF loss of Turbine Cooling Water is due to loss of Turbine Cooling Water Pumps, THEN go to Subsection E1, Loss of Turbine Cooling Water Pumps.

BOP 1. IF EITHER Turbine Cooling Water Pump is available, THEN attempt to start Turbine Cooling Water Pump to restore system flow.

The CRS could direct this step before entering the off normal procedure with the system temperature alarms that come in after the pump trip.

The controls for TCW are located on CP-1.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 4 Page 8 of 17 Event

Description:

Turbine Cooling Water Pump A trips, Turbine Cooling Water Pump B fails to auto start Time Position Applicants Actions or Behavior CRS / BOP 2. IF ANY of the following equipment is in operation, THEN monitor the associated Turbine Cooling Water temperatures AND secure ANY unnecessary loads:

Condenser Vacuum Pumps

Steam Generator Feed Pump Turbines

Heater Drain Pumps

Turbine EH System

Condensate Pumps

Main Turbine Lube Oil Coolers

Generator Exciter Coolers

Generator Hydrogen Coolers

Instrument Air Compressors

Station Air Compressors

Stator Coil Water System

Hydrogen Seal Oil Coolers

Isophase Bus Coolers

Hydrogen Dryers CRS Procedure Caution AT 100% POWER A TOTAL LOSS OF TURBINE COOLING WATER WILL RESULT IN SIGNIFICANT MAIN TURBINE DAMAGE IN 2-3 MINUTES WITH THE GENERATOR AS THE MOST LIMITING COMPONENT.

CRS 3. IF BOTH Turbine Cooling Water pumps are unavailable AND flow can NOT be restored, THEN perform the following:

3.1 Manually trip the Reactor.

3.2 Verify Main Turbine tripped.

This would only be applicable if the crew does not start TCW Pump B.

Examiner Note This event is complete after TCW Pump B is running and the CRS has reviewed the of normal Or As directed by the Lead Evaluator Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 9 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of the tripped Main Feedwater Pump B.

Alarms FWPT B Trip Overspeed (Cabinet F, K-19)

SG 1 Level Hi/Lo (Cabinet F, U-14)

SG 2 Level Hi/Lo (Cabinet F, U-18)

EFW PUMP A UNAVAILABLE (Cabinet M, D-1)

Indications Steam Generator levels dropping MFW Pump B tripped CRS Recognize plant conditions and direct manually tripping the reactor.

ATC If directed by CRS, trip Reactor using 2 Reactor Trip pushbuttons at CP-2.

CRS Direct ATC and BOP to carry out Standard Post trip Actions.

ATC Determine Reactivity Control acceptance criteria are met:

Check reactor power is dropping.

Check startup rate is negative.

Check less than TWO CEAs are NOT fully inserted.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 10 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior Critical Task Establish reactivity control.

This task is satisfied by establishing Emergency Boration prior to completing Standard Post Trip Actions Reactivity Control verification.

ATC Determine 3 CEAs are stuck out and commence Emergency Boration:

Place Makeup Mode selector switch to MANUAL.

Align borated water source by performing one of the following:

o Initiate Emergency Boration using Boric Acid Pump as follows:

o Open Emergency Boration Valve, BAM-133.

o Start one Boric Acid Pump.

o Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A BAM-126B Boric Acid Makeup Pump Recirc Valve B OR o Initiate Emergency Boration using Gravity Feed as follows:

o Open the following Boric Acid Makeup Gravity Feed valves:

o BAM-113A Boric Acid Makeup Gravity Feed Valve A o BAM-113B Boric Acid Makeup Gravity Feed Valve B

1. Close VCT Disch Valve, CVC-183.

2. Verify at least one Charging Pump operating and Charging Header flow 40 GPM.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 11 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior BOP Determine Maintenance of Vital Auxiliaries acceptance criteria are met:

Check the Main Turbine is tripped:

Governor valves closed Throttle valves closed BOP Manually trip the Main Turbine using the Think & Trip pushbuttons on CP-1.

BOP Check the Main Generator is tripped:

GENERATOR BREAKER A tripped GENERATOR BREAKER B tripped EXCITER FIELD BREAKER tripped BOP Check station loads are energized from offsite electrical power as follows:

Train A A1, 6.9 KV non safety bus A2, 4.16 KV non safety bus A3, 4.16 KV safety bus A-DC electrical bus A or C vital AC Instrument Channel Train B B1, 6.9 KV non safety bus B2, 4.16 KV non safety bus B3, 4.16 KV safety bus B-DC electrical bus B or D vital AC Instrument Channel Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 12 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior ATC Determine RCS Inventory Control acceptance criteria are met:

Check that the following conditions exist:

Pressurizer level is 7% to 60%

Pressurizer level is trending to 33% to 60%

Check RCS subcooling is greater than or equal to 28ºF.

ATC Determine RCS Pressure Control acceptance criteria are met by checking that BOTH of the following conditions exist:

Pressurizer pressure is 1750 psia to 2300 psia Pressurizer pressure is trending to 2125 psia to 2275 psia ATC Determine Core Heat Removal acceptance criteria are met:

Check at least one RCP is operating.

Check operating loop T is less than 13ºF.

Check RCS subcooling is greater than or equal to 28ºF.

BOP Determine RCS Heat Removal acceptance criteria are met:

Check that at least one steam generator has BOTH of the following:

Steam generator level is 5% to 80% NR Main Feedwater is available to restore level within 50%-70% NR.

BOP Attempt to start EFW Pump A. The BOP operator will not be able to run EFW Pump A.

ATC Check RCS TC is 530ºF to 550ºF BOP Check steam generator pressure is 885 psia to 1040 psia.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 13 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior BOP Check Feedwater Control in Reactor Trip Override:

MAIN FW REG valves are closed STARTUP FW REG valves are 13% to 21% open Operating main Feedwater pumps are 3800 rpm to 4000 rpm BOP should report both Main Feedwater Pumps are tripped.

BOP Check Feedwater Control in Reactor Trip Override:

MAIN FW REG valves are closed STARTUP FW REG valves are 13% to 21% open Operating main Feedwater pumps are 3800 rpm to 4000 rpm BOP Reset moisture separator reheaters, and check the temperature control valves closed.

ATC Determine Containment Isolation acceptance criteria are met:

Check containment pressure is less than 16.4 psia.

Check NO containment area radiation monitor alarms OR unexplained rise in activity.

Check NO steam plant activity monitor alarms OR unexplained rise in activity.

BOP Determine Containment Temperature and Pressure Control acceptance criteria are met:

Check containment temperature is less than or equal to 120ºF.

Check containment pressure is less than 16.4 psia.

CRS After review of Standard Post Trip Actions, use Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-006, Loss of Main Feedwater Recovery Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 14 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior CRS 1. Confirm diagnosis of a Loss of Main Feedwater by checking Safety Function Status Check Acceptance Criteria are satisfied.

Crew 2. Announce the Loss of Main Feedwater in progress using the plant page.

CRS 3. Advise the Shift Manager to REFER TO EP-001-001, "Recognition &

Classification of Emergency Condition", and implement the Emergency Plan.

N/A 4. IF power has been interrupted to either 3A or 3B safety buses, THEN perform Appendix 20, "Operation of DCT Sump Pumps".

CRS 5. REFER TO Section 6.0, "Placekeeper", and record the time of the reactor trip.

CRS 6. Verify no more than two RCPs are operating.

ATC When directed, secure Reactor Coolant Pumps 1A and 2A as follows:

Place each RCP control switch on CP-2 to stop.

CRS 7. IF ANY of the following conditions exist, THEN perform the following:

a. IF MFW is lost for greater than 30 minutes AND ONE Motor Driven EFW pump is the only EFW pump available, THEN stop ALL RCPs.

b. IF ALL feedwater is lost, THEN stop ALL RCPs.

At this point in time, the EFW Pump B and EFW Pump AB will be running and securing all RCPs is not required.

BOP 8. Check a CCW pump is operating for each energized 4.16 KV safety bus.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 15 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior BOP 9. Conserve steam generator inventory:

a. Verify the following steam generator blowdown isolation valves are closed:

BD 102A, SG BLOWDOWN ISOL STM GEN 1 (IN)

BD 102B, SG BLOWDOWN ISOL STM GEN 2 (IN)

BD 103A, SG BLOWDOWN ISOL STM GEN 1 (OUT)

BD 103B, SG BLOWDOWN ISOL STM GEN 2 (OUT)

b. Verify the following steam generator sampling valves are closed:

SSL 8006A, SAMPLING ISOLATION SG 1

SSL 8006B, SAMPLING ISOLATION SG 2

SSL 301A, SAMPLING ISOLATION MAIN STM LINE 1

SSL 8004A, SAMPLING ISOLATION SG 1

SSL 8004B, SAMPLING ISOLATION SG 2

SSL 301B, SAMPLING ISOLATION MAIN STM LINE 2 N/A 10. IF MSIS has actuated AND opening the MSIVs or MFIVs will aid in the restoration of feedwater, THEN REFER TO Appendix 5-B, "MSIS Main Steam Pressure Reset Procedure", and reset the MSIS.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 16 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior Note After the BOP has completed the step 9 for conserving Steam Generator inventory, cue Malfunction 8 to trip EFW Pump AB.

The CRS must recognize the need to perform the contingency of step 7.

Critical Task

1. Establish a primary to secondary heat sink This task is satisfied by securing all RCPs after Emergency Feedwater Pump AB trips. With Emergency Feedwater Pump A off, Emergency Feedwater Pump B does not have the capacity to provide necessary Emergency Feedwater flow. The requirement is that all RCPs be secured within 30 minutes of the loss of Main Feedwater, the time of the reactor trip.

CRS 7. IF ANY of the following conditions exist, THEN perform the following:

a. IF MFW is lost for greater than 30 minutes AND ONE Motor Driven EFW pump is the only EFW pump available, THEN stop ALL RCPs.

b. IF ALL feedwater is lost, THEN stop ALL RCPs.

With only EFW Pump B running, all RCPs should be secured at this point, but before 30 minutes post trip, since the 1 motor driven pump is only 50% capacity.

Scenario 4, Revision 1

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7/8 Page 17 of 17 Event

Description:

Main Feedwater Pump B trips, manual reactor trip, Emergency Feedwater Pump A fails to run; Main Turbine fails to trip following the reactor trip; 3 CEAs fail to insert following the reactor trip, Emergency Boration, Emergency Feedwater Pump AB trip on overspeed Time Position Applicants Actions or Behavior BOP 11. Replenish inventory in at least one steam generator by performing ANY of the following:

a. Check emergency feedwater available to at least one steam generator.

b. REFER TO Appendix 32, "Establishing Main Feedwater" and restore main feedwater flow to at least one steam generator.

a.1 IF EFW pump AB has tripped, THEN reset as follows:

1) Close BOTH PUMP AB TURB STM SUPPLY valves:

MS 401A

MS 401B Performed at CP-8 Examiner Note The remainder of the steps for resetting EFW Pump AB is performed in the field. The scenario can terminated when MS-401 A and MS-401 B are closed Or As directed by the Lead Evaluator.

Scenario 4, Revision 1

ES-301 Simulator Scenario Quality Checklist Form ES-301-5 Facility: Waterford 3 Date of Exam: March 21, 2011 Operating Test No. NRC A E Scenarios P V 1 2 3 4 T M P E O I L N CREW CREW CREW CREW T N I T POSITION POSITION POSITION POSITION I

C S A B S A B S A B S A B A M

A T R T O R T O R T O R T O L U N Y O C P O C P O C P O C P M(*)

T P E R I U RX 0 1 1 0 NOR 0 1 1 1 SRO-U I/C 1,2,3, 1,3,4, 12 4 4 2 5,7,8 5,7,8 1&2 MAJ 6 6 2 2 2 1 TS 2,4 1,2,3 5 0 2 2 RX 0 1 1 0 NOR 3 1 1 1 1 SRO-I I/C 1,2,5, 1,3,8 1,4,7, 9 4 4 2 6,7 8 1

MAJ 4 6 6 3 2 2 1 TS 1,2,3 3 0 2 2 RX 3 1 1 1 0 NOR 0 1 1 1 SRO-I I/C 1,5 1,2,3, 1,3,4, 14 4 4 2 5,7,8 5,7,8 2

MAJ 4 6 6 3 2 2 1 TS 2,4 1,2,3 5 0 2 2 RX 0 1 1 0 NOR 3 1 1 1 1 SRO-I I/C 1,2,5, 1,3,8 8 4 4 2 6,7 3&4 MAJ 4 6 2 2 2 1 TS 1,2,3 3 0 2 2 NRC Revision 1

RX 0 1 1 0 NOR 3 1 1 1 1 RO I/C 2,6,7 2,5,7 3,5,8 9 4 4 2 1&3 MAJ 4 6 6 3 2 2 1 TS 0 0 2 2 RX 3 1 1 1 0 NOR 0 1 1 1 RO I/C 1,5 1,4,7, 6 4 4 2 8

2&4 MAJ 4 6 2 2 2 1 TS 0 0 2 2 RX 0 1 1 0 RO NOR 3 1 1 1 1 I/C 2,6,7 2,5,7 3,5,8 9 4 4 2 5 MAJ 4 6 6 3 2 2 1 TS 0 0 2 2 RX 4 NOR 4 Spare I/C 1,2,3, 3,7,8 1,2,6, 6,7,8 8 MAJ 5 5 5 TS 1,3 Instructions:

1. Check the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must serve in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must serve in both the SRO and the ATC positions, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position. If an Instant SRO additionally serves in the BOP position, one I/C malfunction can be credited toward the two I/C malfunctions required for the ATC position.

2. Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1-for-1 basis.

3. Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

NRC Revision 1

ES-301 Competencies Checklist Form ES-301-6 Facility: Waterford 3 Date of Exam: March 21, 2011 Operating NRC Test No.:

APPLICANTS CRS ATC BOP Competencies SCENARIO SCENARIO SCENARIO 1 2 3 1 2 3 1 2 3 4 Interpret/Diagnose 1, 2, 1, 2, 1, 2, 1, 2, 1, 3, 1, 2, 3, 5, 3, 4, 2, 3, 2, 5, 1, 2, 1, 2, 3, 4, 3, 5, 3, 4, 3, 4, 4, 5 3, 5, 6, 8 5, 7 4, 6, 6, 7 4, 6, 5, 6, Events and Conditions 7 6, 7, 5, 6, 5, 6, 6, 8 7 7, 8 8 8 7, 8 7, 8 Comply With and All 1, 3, 1, 3, 3, 5, 3, 4, 2, 3, 2, 5, 1, 4, 1, 2, 4, 5 5, 6, 6, 8 5, 7 4, 6, 6, 7 6, 7, 5, 6, Use Procedures (1) 8 7 8 8 Operate Control N/A 1, 3, 1, 3, 3, 5, 3, 4, 2, 3, 2, 5, 1, 4, 1, 2, 4, 5 5, 6, 6, 8 5, 7 4, 6, 6, 7 6, 7, 5, 6, Boards (2) 8 7 8 8 Communicate All 1, 3, 1, 2, 3, 5, 3, 4, 2, 3, 2, 5, 1, 4, 1, 2, 4, 5 3, 5, 6, 8 5, 7 4, 6, 6, 7 6, 7, 5, 6, and Interact 6, 8 7 8 8 Demonstrate All N/A N/A Supervisory Ability (3)

Comply With and 1, 2,3 2, 4 1, 2,3 1, 3 N/A N/A Use Tech. Specs. (3)

Notes:

(1) Includes Technical Specification compliance for an RO.

(2) Optional for an SRO-U.

(3) Only applicable to SROs.

Instructions:

Check the applicants license type and enter one or more event numbers that will allow the examiners to evaluate every applicable competency for every applicant.

Revision 1

ML110970454

Text

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

References:

REFERENCES:

REFERENCES:

REFERENCES:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description: