ML17192A443

301 Draft Items (ADAMS-2A)- Delay Release 2 Yrs

ML17192A443
Person / Time
Site:	Harris
Issue date:	07/11/2017
From:	NRC/RGN-II
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Download: ML17192A443 (763)

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Appendix C Job Performance Measure Form ES-C-1 Worksheet Page 1 of 11 Facility: Harris Nuclear Plant Task No.: 119015H301 Task

Title:

Perform RCS Average Temperature JPM No.: 2016 NRC Exam Data Sheet and Determine Inverse Admin JPM RO A1-1 Count Rate Ratio (1/M)

K/A

Reference:

G2.1.43 RO 4.1 SRO 4.3 Alternate Path - No Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate, discuss or perform, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

A Reactor trip occurred 7 days ago due to a trip of the A MFW pump.

The pump has been repaired.

The unit is in Mode 2, with a plant startup in progress per GP-004, REACTOR STARTUP (MODE 3 TO MODE 2).

Initial Conditions:

The OATC has just completed the 3rd doubling.

The Reactor Startup spreadsheet cannot be used due to a unsatisfactory Benchmark Test message. This has not as yet been resolved, the startup will continue.

The CRS has assigned you to perform manual calculations and plotting of the 1/M data points per Sheet 2 of Attachment 2 and Attachment 3 to predict when the Reactor will become critical. Another operator will verify your calculations when you have completed the attachments.

Initiating Cue:

Place your predicted rod height for Reactor Criticality in the space provided below and circle the YES / NO response for the question.

When complete return your papers to the Examiner.

2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Page 2 of 11 Task Standard: Determines the predicted criticality will occur ABOVE the 500 pcm ECC rod height of 160 steps on CBD.

Required Materials: Calculator, ruler, pencil General

References:

GP-004, Rev 60 Handouts: JPM Cue Sheets Pages 8 and 9, GP-004 marked up to step 24 and Attachment 2 through step 3, and Attachment 3 (page 31) blow up for ease of plotting (11 x 17 size).

Time Critical Task: No Validation Time: 15 minutes Critical Step Justification Critical to correctly calculate inverse count ratio in order to predict the Step 3 estimated critical rod position following the doubling of source range counts Critical to correctly plot the inverse count ratio and controlling rod Step 4 position in order to predict the estimated critical rod position following the doubling of source range counts Critical to correctly calculate inverse count ratio in order to predict the Step 5 estimated critical rod position following the doubling of source range counts Critical to correctly plot the inverse count ratio and controlling rod Step 6 position in order to predict the estimated critical rod position following the doubling of source range counts Critical to correctly calculate inverse count ratio in order to predict the Step 7 estimated critical rod position following the doubling of source range counts Critical to correctly plot the inverse count ratio and controlling rod Step 8 position in order to predict the estimated critical rod position following the doubling of source range counts Critical to determine the estimated critical rod height of 174 steps Step 9 (166 - 182 steps, tolerance based on curve division readability) and NOT within the required band. (ABOVE the + 500 pcm limit) 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 3 of 11 Start Time: __________.

Performance Step: 1 OBTAIN PROCEDURE (GP-004, marked up and 11 x 17 of Attachment 3 will be provided to allow candidates to review)

Standard: Obtains GP-004 and refers to Attachment 2 and 3.

Comment:

Performance Step: 2 Complete Attachment 3 Startup Information Standard: Transfers the following information from Attachment 2 to Attachment 3:

• Date

• Startup Number

• Rod Insertion Limit

• 500 pcm below ECC

• 500 pcm above ECC

• Control Operator The above information has been provided to the candidate as part of the JPM Cue sheet for Attachment 2.

The transfer of this information from Attachment 2 to Attachment 3 may be performed at any time before the Evaluator Note: completion of the JPM IT IS ONLY REQUIRED TO PLOT OUT THE SOURCE RANGE PLOT. IT IS NOT REQUIRED TO PLOT OUT THE INTERMEDIATE RANGE DETECTOR PLOT Comment:

Performance Step: 3 Calculation of 1/M data point at 1421 Standard: Divides initial source range count rate by source range count reading for 1421

• 250 cps / 490 cps = 0.51 (0.510)

Comment:

- Denotes Critical Steps 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 4 of 11 Performance Step: 4 Plot 1/M data point to determine predicted Criticality position at 1421 Standard: Manually plots 1/M data point on Attachment 3 using data from Attachment 2 and interpolates data to determine predicted critical position is > 220 steps on Control Bank D.

Comment:

Performance Step: 5 Calculation of 1/M data point at 1429 Standard: Divides initial source range count rate by source range count reading for 1429

• 250 cps / 975 cps = 0.26 (0.256)

Comment:

Performance Step: 6 Plot 1/M data point to determine predicted Criticality position at 1429 Standard: Manually plots 1/M data point on Attachment 3 using data from Attachment 2 and interpolates data to determine predicted critical position is still > 220 steps on CB D.

Comment:

Performance Step: 7 Calculation of 1/M data point at 1437 Standard: Divides initial source range count rate by source range count reading for 1437

• 250 cps / 2100 cps = 0.12 (0.119)

Comment:

- Denotes Critical Steps 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 5 of 11 Performance Step: 8 Plot 1/M data point to determine predicted Criticality position at 1437 Standard: Manually plots 1/M data point on Attachment 3 using data from Attachment 2 and interpolates data to determine predicted critical position is 174 steps on Control Bank D.

(166 - 182 steps, tolerance based on curve division readability)

Comment:

Performance Step: 9 Documents Attachment 3 results of predicted rod height for criticality Standard: Documents the predicted rod height is 174 steps on Control Bank D (166 - 182 steps, tolerance based on curve division readability)

Answers: Is the predicted Rod Height for Reactor Criticality between the required band? NO Determines the predicted criticality position is ABOVE the 500 pcm ECC value of 160 steps on Control Bank D Comment:

Candidate determines the predicted rod height is 174 steps (166 - 182 steps, tolerance based on curve division readability) and NOT within the required band. Circles - NO Evaluator Note: (ABOVE the 500 pcm ECC limit)

Returns JPM paper work.

END OF JPM Comment:

Stop Time: _________

- Denotes Critical Steps 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 KEY Page 6 of 11 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 KEY Page 7 of 11 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 VERIFICATION OF COMPLETION Page 8 of 11 Job Performance Measure No.: 2016 NRC Admin JPM RO A1-1 Perform RCS Average Temperature Data Sheet and Determine Inverse Count Rate Ratio (1/M)

IAW GP-004 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 A Reactor trip occurred 7 days ago due to a trip of the A MFW pump.

The pump has been repaired.

The unit is in Mode 2, with a plant startup in progress per GP-004, REACTOR STARTUP (MODE 3 TO MODE 2).

Initial Conditions:

The OATC has just completed the 3rd doubling.

The Reactor Startup spreadsheet cannot be used due to a unsatisfactory Benchmark Test message. This has not as yet been resolved, the startup will continue.

The CRS has assigned you to perform manual calculations and plotting of the 1/M data points per Sheet 2 of Attachment 2 and Attachment 3 to predict when the Reactor will become critical. Another operator will verify your calculations when you have completed the attachments.

Initiating Cue:

Place your predicted rod height for Reactor Criticality in the space provided below and circle the YES / NO response for the question below.

When complete return your papers to the Examiner.

Name:

Date:

My predicted Rod Height for Reactor Criticality is steps on Bank Circle the response to the following question:

Is the predicted Rod Height for Reactor Criticality within the required band? YES / NO 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Admin Exam RO A1-1 Rev. 2

Appendix C Page 1 of 11 Form ES-C-1 WORKSHEET Facility: Harris Nuclear Plant Task No.: 119013H304 Task

Title:

Using Valve Maps And Survey Maps JPM No.: 2016 NRC Exam Determine Stay Times For A Clearance Admin JPM RO/SRO A-3 K/A

Reference:

G.2.3.4 RO 3.2 SRO 3.7 Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

A plant shutdown IAW GP-006, Normal Plant Shutdown From Power Operation To Hot Standby (MODE 1 TO MODE 3) is in progress due to a fuel leak causing elevated RCS radiation readings.

The Letdown demins are currently bypassed and unavailable.

1CS-167, VCT Outlet Check Valve gasket failed causing a body to bonnet leak that must be repaired for continued use of the VCT.

Three operators are tasked to hang a clearance on 1CS-167. The clearance includes the following valves:

• 1CS-165, VCT OUTLET LCV-115C

• 1CS-166, VCT OUTLET LCV-115E

• 1CS-321, Seal Water Return Isolation Valve to CSIP Suction Initial

• 1CS-793, ECCS Inner Vent Valve VCT Discharge Downstream 1CS-167 Conditions:

• 1CS-794, ECCS Outer Vent Valve VCT Discharge Downstream 1CS-167 Operator accumulated Whole Body dose for 2016 Operator 1 = 1640 mrem Operator 2 = 1600 mrem Operator 3 = 1580 mrem AND worked at the Surry Nuclear Plant this year from January through March where he has accumulated a whole body does of1540 mrem.

In accordance with PD-RP-ALL-0001, the Radiation Protection Manager has authorized Operator 3 a dose extension to the limit that his signature authority is authorized to administer.

The ALARA group has determined that additional shielding is not warranted for this work.

2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Page 2 of 11 Form ES-C-1 WORKSHEET (Initiating Cue on next page)

Using the supplied valve maps and survey map, determine the maximum allowable individual stay times for each Operator that would prevent exceeding the Duke Energy Annual Administrative dose limit while performing the clearance.

Do not consider dose received during transit. The calculated dose should be Initiating ONLY what they would receive while working at the valves for the clearance.

Cue:

Individual stay times for each Operator that would prevent exceeding the Duke Energy Annual Administrative dose limit plus the extension provided to Operator 3 while performing these activities.

Complete the information below and return to the evaluator when complete.

2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Page 3 of 11 Form ES-C-1 WORKSHEET Task Standard: Calculation of stay times based on survey maps, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 15 minutes for Operator 1, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 minutes for Operator 2, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes for Operator 3.

Required Materials: Survey map A78 RAB 261' Volume Control Tank Valve Gallery Map 23 General

References:

PD-RP-ALL-0001, Radiation Worker Responsibilities, Rev 3 LIMIT = 2 rem but a Dose Extension individual limit up to 3.4 rem can be authorized by the Radiation Protection Manger - RPM (or designee).

Time Critical Task: No Validation Time: 15 minutes Critical Task Justification Step 1 Must determine dose rates in order to calculate stay time Step 2 Must determine available dose to determine stay time.

IF incorrect calculation of stay time is made the individuals could exceed Step 3 their dose limits.

2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Page 4 of 11 Form ES-C-1 PERFORMANCE INFORMATION START TIME:

The order of performance does not matter Evaluator Note: IF THE APPLICANT ASKS FOR IT: PD-RP-ALL-0001, Radiation Worker Responsibilities, Rev 3 Performance Step: 1 Using Radiological Survey Record Map A78 and RAB 261' Volume Control Tank Valve Gallery Map 23, determines dose rates in the area where the clearance will be applied Standard: Identifies that General Area Dose Rates are 160 mrem/hr Comment:

Performance Step: 2 Determine the remaining dose for the year for each individual Standard: Operator 1: 360 mrem 2000 mrem - 1640 mrem = 360 mrem Operator 2: 400 mrem 2000 mrem - 1600 mrem = 400 mrem Operator 3: 280 mrem 3400 mrem - 1580 mrem (DEP) - 1540 mrem (Surry) = 280 mrem Comment: IAW PD-RP-ALL-001, Section 5.2.5 Dose Extension and Reduction, Rev. 3 Page 18. The RPM is authorized to extend an individuals limit up to 3.4 rem (3400 mrem).

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Page 5 of 11 Form ES-C-1 PERFORMANCE INFORMATION Performance Step: 3 Determine stay time for each operator (based on 1st Operator reaching 2 Rem, the 2nd Operator reaching 2 Rem and the 3rd Operator reaching 3.4 Rem - for the year)

Standard: Operator 1: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 15 minutes 360 mrem ÷ 160 mrem/hr = 2 hrs ( 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 15 minutes)

Operator 2: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 minutes 400 mrem ÷ 160 mrem/hr = 2.5 hrs ( 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 30 minutes)

Operator 3: 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes 280 mrem ÷ 160 mrem/hr = 1.75 hrs (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes)

Comment:

Terminating Cue: After the stay time has been calculated, this JPM is complete.

END OF JPM STOP TIME:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Page 6 of 11 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Admin JPM SRO A Using Valve Maps And Survey Maps Determine Stay Times For A Clearance PD-RP-ALL-0001 Rev. 3 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Form ES-C-1 JPM CUE SHEET A plant shutdown IAW GP-006, Normal Plant Shutdown From Power Operation To Hot Standby (MODE 1 TO MODE 3) is in progress due to a fuel leak causing elevated RCS radiation readings.

The Letdown demins are currently bypassed and unavailable.

1CS-167, VCT Outlet Check Valve gasket failed causing a body to bonnet leak that must be repaired for continued use of the VCT.

Three operators are tasked to hang a clearance on 1CS-167. The clearance includes the following valves:

• 1CS-165, VCT OUTLET LCV-115C

• 1CS-166, VCT OUTLET LCV-115E

• 1CS-321, Seal Water Return Isolation Valve to CSIP Suction Initial

• 1CS-793, ECCS Inner Vent Valve VCT Discharge Downstream 1CS-167 Conditions:

• 1CS-794, ECCS Outer Vent Valve VCT Discharge Downstream 1CS-167 Operator accumulated Whole Body dose for 2016 Operator 1 = 1640 mrem Operator 2 = 1600 mrem Operator 3 = 1580 mrem AND worked at the Surry Nuclear Plant this year from January through March where he has accumulated a whole body does of1540 mrem.

In accordance with PD-RP-ALL-0001, the Radiation Protection Manager has authorized Operator 3 a dose extension to the limit that his signature authority is authorized to administer.

The ALARA group has determined that additional shielding is not warranted for this work.

Using the supplied valve maps and survey map, determine the maximum allowable individual stay times for each Operator that would prevent exceeding the Duke Energy Annual Administrative dose limit while performing the clearance.

Do not consider dose received during transit. The calculated dose should be Initiating ONLY what they would receive while working at the valves for the clearance.

Cue:

Individual stay times for each Operator that would prevent exceeding the Duke Energy Annual Administrative dose limit plus the extension provided to Operator 3 while performing these activities.

Complete the information below and return to the evaluator when complete.

Name: __________________________________________

Date:

Record the maximum allowable stay time calculations below to the nearest hour and minute.

Operator 1: _______________ Operator 2: _______________ Operator 3: _______________

2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Form ES-C-1 JPM CUE SHEET 2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Form ES-C-1 JPM CUE SHEET 2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Form ES-C-1 JPM CUE SHEET 2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Form ES-C-1 JPM CUE SHEET 2016 NRC Exam Admin JPM RO/SRO A3 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Page 1 of 13 Facility: Harris Nuclear Plant Task No.: 119015H301 Task

Title:

Perform RCS Average Temperature JPM No.: 2016 NRC Exam Data Sheet and Determine Inverse Admin JPM SRO A1-1 Count Rate Ratio (1/M)

K/A

Reference:

G2.1.43 RO 4.1 SRO 4.3 Alternate Path - NO Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate, discuss or perform, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

A Reactor trip occurred 7 days ago due to a trip of the A MFW pump.

The pump has been repaired.

The unit is in Mode 2, with a plant startup in progress per GP-004, REACTOR STARTUP (MODE 3 TO MODE 2).

Initial Conditions:

The OATC has just completed the 3rd doubling.

The Reactor Startup spreadsheet cannot be used due to a unsatisfactory Benchmark Test message. This has not as yet been resolved, the startup will continue.

The CRS has assigned you to perform manual calculations and plotting of the 1/M data points per Sheet 2 of Attachment 2 and Attachment 3 to predict when the Reactor will become critical. Another operator will verify your calculations when you have completed the attachments.

Place your predicted rod height for Reactor Criticality in the space Initiating Cue:

provided.

Circle the YES / NO response for the question and provide actions (IF ANY) based on your determination.

When complete return your papers to the Examiner.

2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Page 2 of 13 Task Standard: Determines the predicted criticality will occur ABOVE the 500 pcm ECC rod height of 160 steps on CBD.

Required Materials: Calculator General

References:

GP-004, Rev 60 Handouts: JPM Cue Sheets Pages 8 and 9, GP-004 Attachment 3 page 31 Time Critical Task: No Validation Time: 20 minutes Critical Step Justification Critical to correctly calculate inverse count ratio in order to predict the Step 3 estimated critical rod position following the doubling of source range counts Critical to correctly plot the inverse count ratio and controlling rod Step 4 position in order to predict the estimated critical rod position following the doubling of source range counts Critical to correctly calculate inverse count ratio in order to predict the Step 5 estimated critical rod position following the doubling of source range counts Critical to correctly plot the inverse count ratio and controlling rod Step 6 position in order to predict the estimated critical rod position following the doubling of source range counts Critical to correctly calculate inverse count ratio in order to predict the Step 7 estimated critical rod position following the doubling of source range counts Critical to correctly plot the inverse count ratio and controlling rod Step 8 position in order to predict the estimated critical rod position following the doubling of source range counts Critical to determine the estimated critical rod height of 174 steps Step 9 (166 - 182 steps, tolerance based on curve division readability) and NOT within the required band. (ABOVE the + 500 pcm limit)

Critical to determine the requirements of GP-004 for criticality not within Step 10 the required band (ABOVE the + pcm limit) is to perform the actions of step 26.b - this is a reactivity concern 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 3 of 13 Start Time: __________

Performance Step: 1 OBTAIN PROCEDURE (GP-002 Attachment 2 and 3 will be provided to allow candidates to write on)

Standard: Obtains GP-004 and refers to Attachment 2 and 3.

Comment:

Performance Step: 2 Complete Attachment 3 Startup Information Standard: Transfers the following information from Attachment 2 to Attachment 3:

• Date

• Startup Number

• Rod Insertion Limit

• 500 pcm below ECC

• 500 pcm above ECC

• Control Operator The above information has been provided to the candidate as part of the JPM Cue sheet for Attachment 2.

The transfer of this information from Attachment 2 to Attachment 3 may be performed at any time before the Evaluator Note: completion of the JPM IT IS ONLY REQUIRED TO PLOT OUT THE SOURCE RANGE PLOT. IT IS NOT REQUIRED TO PLOT OUT THE INTERMEDIATE RANGE DETECTOR PLOT Comment:

Performance Step: 3 Calculation of 1/M data point at 1421 Standard: Divides initial source range count rate by source range count reading for 1421

• 250 cps / 490 cps = 0.51 (0.510)

Comment:

- Denotes Critical Steps 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 4 of 13 Performance Step: 4 Plot 1/M data point to determine predicted Criticality position at 1421 Standard: Manually plots 1/M data point on Attachment 3 using data from Attachment 2 and interpolates data to determine predicted critical position is > 220 steps on Control Bank D.

Comment:

Performance Step: 5 Calculation of 1/M data point at 1429 Standard: Divides initial source range count rate by source range count reading for 1429

• 250 cps / 975 cps = 0.26 (0.256)

Comment:

Performance Step: 6 Plot 1/M data point to determine predicted Criticality position at 1429 Standard: Manually plots 1/M data point on Attachment 3 using data from Attachment 2 and interpolates data to determine predicted critical position is still > 220 steps on CB D.

Comment:

Performance Step: 7 Calculation of 1/M data point at 1437 Standard: Divides initial source range count rate by source range count reading for 0545

• 250 cps / 2100 cps = 0.12 (0.119)

Comment:

- Denotes Critical Steps 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 5 of 13 Performance Step: 8 Plot 1/M data point to determine predicted Criticality position at 1437 Standard: Manually plots 1/M data point on Attachment 3 using data from Attachment 2 and interpolates data to determine predicted critical position is 174 steps on Control Bank D.

(166 - 182 steps, tolerance based on curve division readability)

Comment:

Performance Step: 9 Documents Attachment 3 results of predicted rod height for criticality Standard: Documents the predicted rod height is 174 steps on Control Bank D (166 - 182 steps, tolerance based on curve division readability)

Answers: Is the predicted Rod Height for Reactor Criticality between the required band? NO Determines the predicted criticality position is ABOVE the 500 pcm ECC value of 160 steps on Control Bank D Comment:

- Denotes Critical Steps 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 PERFORMANCE INFORMATION Page 6 of 13 Performance Step: 10 Complete the following: Assuming the Reactor achieves criticality (or would achieve criticality) at your predicted rod height. What actions are required (IF ANY)?

B. ACTIONS are required in accordance with GP-004 -

(if true then list the required step number(s) below:

Correct response is: GP-004 step 26.b GP-004, Actions required when criticality is NOT achieved within 500 pcm of the ECC are as follows:

Notes prior to step 26:

NOTE: The maximum allowable administrative difference between the ECC and the actual critical condition is 500 pcm.

NOTE: The following Step is only required to be performed if criticality is NOT achieved within 500 pcm of the ECC.

Step 26.b IF criticality is NOT achieved by the Maximum Estimated Critical Rod Height/Minimum Estimated Boron concentration (rod height or Boron concentration for 500 pcm above the ECC, as listed in Step 5.0.14)

THEN PERFORM the following:

(1) REINSERT all Control Banks.

(2) EVALUATE the conditions resulting in not achieving criticality.

(3) Before any subsequent startup, COMPLETE a new ECC and 1/M Data Plot using the same GP-004 and startup number.

Standard: Provides the requirements that are stated in GP-004 step 26.b for criticality NOT achieved by the Maximum Estimated Critical Rod Height/Minimum Estimated Boron concentration.

Completes evaluation and returns paper work to Evaluator.

Evaluator Note: End of JPM.

Completion Time: _ _______

- Denotes Critical Steps 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 KEY Page 7 of 13 My predicted Rod Height for Reactor Criticality is 174 (166 - 182 steps) steps on Bank D .

Complete the following: Assuming the Reactor achieves criticality at your predicted rod height. What actions (IF ANY) are required if the Reactor were to reach criticality at this height?

A. NO ACTIONS are required (if true then circle this response) - NOT CIRCLED B. ACTIONS are required in accordance with GP-004 - (if true then list the required performance step number below:

26.b 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 KEY Page 8 of 13 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 KEY Page 9 of 13 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 VERIFICATION OF COMPLETION Page 10 of 13 Job Performance Measure No.: 2016 NRC JPM Common RO SRO A1-1 Perform RCS Average Temperature Data Sheet and Determine Inverse Count Rate Ratio (1/M)

GP-004 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 A Reactor trip occurred 7 days ago due to a trip of the A MFW pump.

The pump has been repaired.

The unit is in Mode 2, with a plant startup in progress per GP-004, REACTOR STARTUP (MODE 3 TO MODE 2).

Initial Conditions:

The OATC has just completed the 3rd doubling.

The Reactor Startup spreadsheet cannot be used due to a unsatisfactory Benchmark Test message. This has not as yet been resolved, the startup will continue.

The CRS has assigned you to perform manual calculations and plotting of the 1/M data points per Sheet 2 of Attachment 2 and Attachment 3 to predict when the Reactor will become critical. Another operator will verify your calculations when you have completed the attachments.

Record your predicted rod height for Reactor Criticality in the space Initiating Cue:

provided below.

Record the correct response for the question and provide actions (IF ANY) based on your determination.

When complete return your papers to the Examiner.

Name:

Date:

My predicted Rod Height for Reactor Criticality is steps on Bank Complete the following: Assuming the Reactor achieves criticality (or would achieve criticality) at your predicted rod height. What actions are required (IF ANY)?

A. NO ACTIONS are required (if true then circle this response)

B. ACTIONS are required in accordance with GP-004 - (if true then list the required performance step number below:

2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Admin Exam SRO A1-1 Rev. 2

Appendix C Page 1 of 9 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 002005H101 Task

Title:

Determine the amount of RCS JPM No.: 2016 NRC Exam inventory that will be drained from Admin JPM RO A1-2 RCS during the performance of GP-008, Draining the RCS K/A

Reference:

G2.1.25 RO 3.9 SRO 4.2 ALTERNATE PATH: NO Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues.

When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• A shift turnover is underway

• The previous shift is implementing GP-008, Draining The Reactor Coolant System

• An initial drain down of the RCS is in progress in preparation for the upcoming Refueling Outage Initial Conditions:

• The drain is on hold for turnover and level is currently being maintained stable at the Reactor Vessel Flange

• After turnover, the directions are to drain the RCS to -70 in preparation for Nozzle Dam installation

• The Shift Manager has given permission to the crew to enter lower inventory conditions.

You are an Extra Licensed Operator on shift. The CRS wants to know how many gallons of water will be drained from the RCS to ensure adequate drain tank volume.

You are required to calculate the amount of additional water that will be Initiating Cue: drained from the RCS from the current level to the directed level using GP-008, Draining the Reactor Coolant System, Attachment 5.

Record your total to the nearest gallon.

2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C Page 2 of 9 Form ES-C-1 Worksheet Task Standard: Determine the calculated amount of RCS to be drained within specified limits.

Required Materials: Calculator, GP-008, Draining the RCS, Attachment 5 - Vessel Volume to Level Comparison General

References:

GP-008, RCS, Attachment 5 , Rev. 44 Time Critical Task: No Validation Time: 12 minutes Critical Step Justification An accurate total of RCS drainage is required to be provided to the Shift Manager (within tolerances). 32,900 gallons + 1000 gallons Step 6 Draining the RCS too low could cause the RHR pumps to loose suction which would be a loss of shutdown cooling.

2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C Page 3 of 9 Form ES-C-1 PERFORMANCE INFORMATION START TIME:

Step 1 Performance Step: 1 Obtain procedure Standard: Obtains GP-008, Draining The Reactor Coolant System, Attachment 5, Vessel Volume to level comparison Comment:

Step 2 Performance Step: 2 Determine current RCS volume Standard: Using GP-008, Attachment 5 Determines current RCS volume with RCS level at Vessel Flange with SGs FULL to be 60,595 gallons Comment:

Step 3 Performance Step: 3 Determine RCS volume Top of Loops with SGs FULL Standard: Using GP-008, Attachment 5 Determines RCS volume with RCS level at Top of Loops with SGs FULL to be 55,965 gallons Comment:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C Page 4 of 9 Form ES-C-1 PERFORMANCE INFORMATION Step 4 Performance Step: 4 Determine RCS volume at Top of Loops after SGs drain Standard: Using GP-008, Attachment 5 Determines RCS volume with RCS level at Top of Loops with SGs DRAINED to be 28,925 gallons (-65)

Comment:

Step 5 Performance Step: 5 Determine RCS volume at -70 with SGs DRAINED Standard: Using GP-008, Attachment 5 Determines RCS volume with RCS level at -70 with SGs DRAINED to be 28,925 gallons Mid Loop -82 (24,744 gallons)

Top of Loops -65 (28,925 gallons) 17 (4,181 gallons) between Top of Loops and Mid Loop 245.941 gallons per inch

-65 to -70 = 5 inches x 245.941 = 1,229.705 1,230 gallons Comment:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C Page 5 of 9 Form ES-C-1 PERFORMANCE INFORMATION Step 6 Performance Step: 6 Determine total RCS volume drained from Vessel Flange with SGs FULL to -70 Standard: Using GP-008, Attachment 5 Determines total volume drained to be:

32,900 gallons (+ 1000 gallons since curve could be used)

Current level Vessel Flange with SGs Full to Top of Loops with SGs FULL 60,595 - 55,965 = 4,630 gallons SGs Drained level maintained at Top of Loops 55,965 - 28,925 = 27,040 gallons Top of Loops with SGs Drained to -70 1,230 gallons Total drained volume = 4,630 + 27,040 + 1,230 = 32,900 gallons Comment:

After the total volume of RCS from Vessel Level at Flange with SGs FULL to -70 below the Flange is calculated:

Evaluator Note: Evaluation on this JPM is complete.

END OF JPM STOP TIME:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C Page 6 of 9 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Admin JPM RO A1-2 Determine the amount of RCS inventory that will be drained from RCS during the performance of GP-008, Draining the RCS.

Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• A shift turnover is underway

• The previous shift is implementing GP-008, Draining The Reactor Coolant System

• An initial drain down of the RCS is in progress in preparation for the upcoming Refueling Outage Initial Conditions:

• The drain is on hold for turnover and level is currently being maintained stable at the Reactor Vessel Flange

• After turnover, the directions are to drain the RCS to -70 in preparation for Nozzle Dam installation

• The Shift Manager has given permission to the crew to enter lower inventory conditions.

You are an Extra Licensed Operator on shift. The CRS wants to know how many gallons of water will be drained from the RCS to ensure adequate drain tank volume.

Initiating Cue: You are required to calculate the amount of additional water that will be drained from the RCS from the current level to the directed level using GP-008, Draining the Reactor Coolant System, Attachment 5.

Record your total to the nearest gallon.

Name ________________________________________

Date _____________________________

2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 Name:

Date:

Show how the calculation was performed in the space provided below:

Total number of RCS gallons to be drained:

2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM RO A1-2 Rev. 2

Appendix C Page 1 of 13 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301079H401 Task

Title:

During a Loss of Shutdown Cooling, JPM No.: 2016 NRC Exam determine the time that the RCS will Admin JPM reach Core Boiling and Boil-Off SRO A1-2 K/A

Reference:

G2.1.20 RO 4.6 SRO 4.6 Examinee: _________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

The unit was operating at 100% power for the last 15 months.

On 06/05/16 at 0000 the plant was shut down for a refueling outage.

• The Reactor cavity has been filled.

• A RHR pump tripped when it was restarted after being realigned for Shutdown Cooling mode. Maintenance has determined that motor repairs are required and are not expected to be completed until 06/25/16.

• The Refuel crew has reported the Manipulator Crane has interlock issues

• No fuel has been moved The current date and time is 06/18/16 at 1200 Initial Conditions:

• All of the fuel still remains in the vessel due to complications with the Manipulator crane.

• A RCS leak has developed with cavity level lowering at a rate of 2 feet per hour. Current level is approximately 3 feet above the flange.

• The crew is implementing AOP-020, Loss of RCS Inventory or RHR While Shutdown.

• The B RHR pump just tripped.

• The last valid RCS temperature logged just before the B RHR pump tripped was 135°F.

In Accordance with AOP-020, Section 3.4, step 13, you are directed to refer to curves H-X-8 through H-X-11 and determine:

1. The time to reach core boiling and Initiating Cue: 2. Core boil-off time Mark up your curves to indicate where you are determining these times.

Write your estimates of time to boil and time to boil-off on the lines at the bottom of this page (below).

Calculate your times in hours and minutes 2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Page 2 of 13 Form ES-C-1 Worksheet Task Standard: Candidate obtains curves and correctly identifies the time to reach core boiling and core boil-off time Required Materials: Curve Book Straight Edge General

References:

AOP-020 (Rev. 38) Curve Book curves H-X-8, 9, 10 and 11 (All Rev. 3)

Time Critical Task: No Validation Time: 15 minutes Critical Step Justification Step required in order to accurately determine time to boil using the Step 3 appropriate curve.

Step required in order to accurately determine time to boil-off using the Step 4 appropriate curve.

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Page 3 of 13 Form ES-C-1 VERIFICATION OF COMPLETION START TIME:

Performance Step: 1 OBTAIN CURVES NEEDED FOR CALCULATION (Curve Book will be provided to the candidate)

Standard: Refers to curves H-X-8 through H-X-11 Comment:

Performance Step: 2 Refers to provided data and determines that curve H-X-9 is required to calculate time to boil and curve H-X-11 is required to calculate boil-off time Standard: Reviews curves and determines which ones are appropriate to determine the time to boil and boil-off time Comment:

Performance Step: 3 Based on time since shutdown (06/05/16 - 6/18/16) 13 days 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> since shutdown and current RCS temperature of 135°F using curve H-X-9 determine time to boil.

(Interpolate 125°-150° lines)

Standard: Reviews curve H-X-9 Determines that time to boil is ~18 minutes

(+ 2 minutes, 16 - 20 min is acceptable)

Comment:

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Page 4 of 13 Form ES-C-1 VERIFICATION OF COMPLETION Performance Step: 4 Based on time since shutdown (06/05/16 - 6/18/16) 13 days 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> since shutdown and current RCS temperature of 135°F using curve H-X-11 determine time to boil-off Standard: Reviews curve H-X-11 Determines that time to boil-off is 4 hrs

(+ 15 minutes) or (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 15 minutes to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 45 minutes)

Note: The answer can be calculated in just minutes Acceptable to have 240 minutes (255 minutes to 225 minutes).

Comment:

Terminating Cue: After completing the time to boil and time to boil-off calculation, the evaluation on this JPM is complete.

END OF JPM STOP TIME:

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Page 5 of 13 Form ES-C-1 VERIFICATION OF COMPLETION KEY Initial conditions: Reactor cavity filled for refueling without fuel movement due to Manipulator Crane and Source Range problems. Core cooling is lost at 1200 and 13 days after shutdown. Core Exit Thermocouples are rising and are currently 135°F (a point between the dashed 150°F line and the solid 125°F line on the graph). Estimated time to boiling onset will be approximately 18 minutes from the time of the loss of cooling event.

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Page 6 of 13 Form ES-C-1 VERIFICATION OF COMPLETION KEY Initial conditions: Reactor cavity filled for refueling without fuel movement due to Manipulator Crane and Source Range problems. Core cooling is lost at 1200 and 13 days after shutdown. Core Exit Thermocouples are rising and are currently 135°F. Estimated time to reach boil off will be approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the time of the loss of cooling event. (A point on the curve between the dashed 150°F line and the solid 125°F line.)

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Page 7 of 13 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Admin JPM SRO A1-2 During a Loss of Shutdown Cooling, determine the time that the RCS will reach Core Boiling and Boil-Off Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 The unit was operating at 100% power for the last 15 months.

On 06/05/16 at 0000 the plant was shut down for a refueling outage.

• The Reactor cavity has been filled.

• A RHR pump tripped when it was restarted after being realigned for Shutdown Cooling mode. Maintenance has determined that motor repairs are required and are not expected to be completed until 06/25/16.

• The Refueling SRO has reported the Manipulator Crane has interlock issues

• No fuel has been moved Initial Conditions: The current date and time is 06/18/16 at 1200

• All of the fuel still remains in the vessel due to complications with the Manipulator crane.

• A RCS leak has developed with cavity level lowering at a rate of 2 feet per hour. Current level is approximately 3 feet above the flange.

• The crew is implementing AOP-020, Loss of RCS Inventory or RHR While Shutdown.

• The B RHR pump just tripped.

• The last valid RCS temperature logged just before the B RHR pump tripped was 135°F.

In Accordance with AOP-020, Section 3.4, step 13, you are directed to refer to curves H-X-8 through H-X-11 and determine:

1. The time to reach core boiling and

2. Core boil-off time Initiating Cue:

Mark up your curves to indicate where you are determining these times.

Write your estimates of time to boil and time to boil-off on the lines at the bottom of this page (below).

Calculate your times in hours and minutes Name _________________________________________________

Date __________________

Record your calculations here and return your curves to the examiner:

TIME TO BOIL (hours / minutes) _________________________

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 TIME TO BOIL-OFF (hours / minutes) _____________________

2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM SRO A1-2 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet

`Facility: Harris Nuclear Plant Task No.: 345001H602 Task

Title:

Classify an Event JPM No.: 2016 NRC Exam Admin JPM SRO A4 K/A

Reference:

G2.4.41 RO 2.9 SRO 4.6 Alternate Path - NO Examinee: _______________________ NRC Examiner: _________________

Facility Evaluator: _______________________ Date: _________________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate, discuss or perform, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

This is a TIME CRITICAL JPM Given the following plant conditions:

At 1158

• The unit is operating at 100% power

• A-SA EDG is under clearance The following occurs at 1159:

• A Loss of Offsite power

• Several Reactor First Out Annunciators are received but the Reactor remains at Power

• The OATC attempts to trip the Reactor from the MCB but neither Reactor Trip switch functions Initial Conditions:

• FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, is entered and an Operator is dispatched to trip the Reactor locally

• The Turbine is tripped from the MCB

• Reactor power is ~ 12%

At 1214:

• The Turbine Building Operator has opened the Reactor trip breakers

• RCS pressure rapidly decreased causing an automatic Safety Injection

• The Outside Operator reports that the A UAT is on fire The time is now 1233:

• The A UAT fire is extinguished Evaluate the EAL Matrix and determine the HIGHEST classification required for these plant conditions.

NOTE: DO NOT use SEC judgment.

Initiating Cue:

Write out the HIGHEST EAL classification in blank provided then return your assessment page to the Evaluator.

2016 NRC Exam Admin JPM SRO A4 Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Task Standard: Classify the highest EAL as an Site Area Emergency SS3.1.

Required Materials: None General

References:

PEP-110 EAL Matrix PEP-110 Rev. 25 EP-EAL (allowed reference) Rev. 16 Handouts:

• Attached Initial Conditions

• PEP-110 Rev. 25

• PEP-110 EAL Matrix

• EP-EAL Rev. 16 Time Critical Task: YES - 15 minutes for classification.

Validation Time: 15 minutes for classification CRITICAL STEP JUSTIFICATION Classification of the event is critical for determining State and County Step 2 notifications, public information notices, site information notices, and event reportability to the Nuclear Regulatory Commission.

2016 NRC Exam Admin JPM SRO A4 Rev. 2

Appendix C Page 3 of 6 Form ES-C-1 PERFORMANCE INFORMATION Start Time for this portion of JPM begins when the Evaluator Cue:

individual has been briefed.

START TIME:

Performance Step: 1 OBTAINS EP-EAL and EAL Matrix.

Standard : Obtains EP-EAL and EAL Matrix Comments:

Performance Step: 2 Identify EAL Classification for events in progress.

Standard : There are 3 possible classifications for these conditions:

Two lower level classifications:

HU2.1 Unusual Event Fire not extinguished within 15 min. of Control Room notification or verification of a Control Room fire alarm in any Table H-1 area HA2.1 Alert - does not apply since the UAT is NOT a safety related component SA1.1 Alert (A-EDG under clearance with LOOP)

AC power capability to 6.9 KV emergency buses 1A-SA and 1B-SB reduced to a single power source for 15 min.

AND Any additional single power source failure will result in station blackout The HIGHEST EAL classification is:

SS3.1 Site Area Emergency An automatic trip failed to shut down the reactor AND Manual actions taken at the reactor control console (actuation of MCB Reactor Trip Switch #1, #2 or MCB Turbine Trip switch) do not shut down the reactor as indicated by reactor power 5%

Collect the candidates classification page.

Examiners Cue: After the candidate returns this JPM classification.

END of JPM.

STOP TIME:

- Denotes Critical Steps 2016 NRC Exam Admin JPM SRO A4 Rev. 2

Appendix C Page 4 of 6 Form ES-C-1 PERFORMANCE INFORMATION

- Denotes Critical Steps 2016 NRC Exam Admin JPM SRO A4 Rev. 2

Appendix C Page 5 of 6 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Admin JPM SRO A4 Classify an Event PEP-110 and EP-EAL Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam Admin JPM SRO A4 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 This is a TIME CRITICAL JPM Given the following plant conditions:

At 1158

• The unit is operating at 100% power

• A-SA EDG is under clearance The following occurs at 1159:

• A Loss of Offsite power

• Several Reactor First Out Annunciators are received but the Reactor remains at Power

• The OATC attempts to trip the Reactor from the MCB Initial Conditions: but neither Reactor Trip switch functions

• FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, is entered and an Operator is dispatched to trip the Reactor locally

• The Turbine is tripped from the MCB

• Reactor power is ~ 12%

At 1214:

• The Turbine Building Operator has opened the Reactor trip breakers

• RCS pressure rapidly decreased causing an automatic Safety Injection

• The Outside Operator reports that the A UAT is on fire The time is now 1233:

• The A UAT fire is extinguished Evaluate the EAL Matrix and determine the HIGHEST classification required for these plant conditions.

Initiating Cue:

NOTE: DO NOT use SEC judgment.

Write out the HIGHEST EAL classification in blank provided then return your assessment page to the Evaluator.

Name:

Date:

Highest EAL Classification for the plant conditions:

2016 NRC Exam Admin JPM SRO A4 Rev. 2

Appendix C Page 1 of 19 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 088017H601 Task

Title:

Determine the Minimum and JPM No.: 2016 NRC Exam Maximum Allowed Indicated Flow Admin JPM for MCR Ventilation RO / SRO A2 K/A

Reference:

G2.2.44 RO 4.2 SRO 4.4 Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Control Room HVAC is being aligned for Post-Accident Operation in accordance with OP-173, Control Room Area HVAC System, Section Initial Conditions: 8.2. The crew has completed steps up through step 8.2.2.8.

The CRS has directed you to perform OP-173, section 8.2.2, steps 9.a through 9.d. Assume all verifications will be completed after you submit your completed paper work.

Initiating Cue:

Current RM-23 Outside Air Intake Radiation Monitor readings and ERFIS Met Tower data are provided on the following pages.

Record your answers in the spaces provided below:

2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 2 of 19 Form ES-C-1 Worksheet Task Standard: Determines the Minimum and Maximum Allowed Indicated Flow for MCR ventilation.

Required Materials: Marked up copy of OP-173, Control Room Ventilation, with section 8.2 completed up to step 8.2.2.8.

General

References:

OP-173, Rev. 37 Time Critical Task: No Validation Time: 15 minutes The following setup was required to achieve a request by Mike Donithan (2016 NRC Chief Examiner) to obtain photos or data screen prints to provide as reference material for the JPM.

This data also includes a wind direction thats toward the correct (South) intake to increase the cognitive ability by having the candidate consider wind direction in their decision.

Simulator Setup Reset Simulator to IC 19 Set Rad Monitors:

South Intake RC-1CZ-3505A1-SA 4.16E-8 µc/ml RC-1CZ-3505B1-SB 3.87E-7 µc/ml North Intake RC-1CZ-3505A2-SA 8.76E-6 µc/ml RC-1CZ-3505B2-SB 6.93E-5 µc/ml Set atmospheric conditions to:

• MMT1011 86°F

• MMT1002 29.54 in Hg Additionally set all other ambient temperatures on ERFIS METTOWER screen to near 86°F Set All wind directions on ERFIS METTOWER screen to near 180° (south)

NOTE: ERFIS screens will take some time to show the values Take photos of RM-23 indications for the Rad Monitors to allow the candidate to determine which intake to use.

Take photos or print screen for ERFIS METTOWER data.

2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 3 of 19 Form ES-C-1 Worksheet Critical Step Justification Must determine which intake to use to prevent potential unnecessary Step 2 contamination or inhalation of airborne radiation of the MCR personnel Step 3 Must determine the correct correction factor or calculation will be wrong.

Step 4 Must determine Minimum Allowed Indicated flow to complete the task.

Step 5 Must determine Maximum Allowed Indicated flow to complete the task.

2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 4 of 19 Form ES-C-1 VERIFICATION OF COMPLETION START TIME:

OP-173, Section 8.2.2 Note prior to Step 9 Performance Step: 1 NOTE: Opening the EMERGENCY FILTRATION OUTSIDE AIR INLET in a timely manner will minimize habitability concerns for the number of people in the MCR when it is isolated. (Reference 2.6.4)

NOTE: The decision of which Control Room Emergency Outside Air Intakes should be used to supply air to the system is based on the event in progress which has made recirculation necessary.

• In an emergency situation such as toxic gas, wind direction may be of primary consideration.

• In the event of high radiation, The RM-23 for the following radiation monitors should be used to determine which Outside Air intake has the lowest radiation reading:

o SOUTH INTAKE NORTH INTAKE o RC-1CZ-3505A1-SA RC-1CZ-3505A2-SA o RC-1CZ-3505B1-SB RC-1CZ-3505B2-SB

• If the system is being placed into recirculation for maintenance reasons, either intake may be used.

NOTE: If, during the performance of this Section, MMT1011 decreases by 20°F or more, or MMT1002 increases by 0.5 in Hg or more, the correction factor in the following step would have to be re-determined. An increase in MMT1011, or a decrease in MMT1002 results in a more conservative calculation, so would not require re-determining the correction factor.

NOTE EMERGENCY FILTRATION OUTSIDE AIR INLETS can be manually opened on loss of power.

NOTE The intent of the following Step is to obtain a flow high enough to ensure MCR DP is greater than 0.125 inwg to ALL adjacent areas. However, it may not be necessary to obtain maximum flow conditions to obtain required DP.

Care should be taken to avoid excessive DP across MCR boundary doors.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 5 of 19 Form ES-C-1 VERIFICATION OF COMPLETION OP-173 Section 8.2.2 Step 9.a Performance Step: 2 PERFORM the following to pressurize the control room to 0.125 INWG:

a. Using the NOTE above, DETERMINE which of the below Control Room Emergency Outside Air Intakes should be used to supply outside air to the system:

SOUTH / NORTH (circle one)

Standard: Reviews ERFIS screen printout and identifies South Intake RC-1CZ-3505A1-SA reading 4.32E-8 µc/ml as the lowest and circles SOUTH.

Comment:

OP-173 Section 8.2.2 Step 9.b Performance Step: 3 DETERMINE the correction factor as follows:

(17.714) ( in Hg)

MMT1002 = _________

__________ °F + 460 MMT1011 Standard: Determines correction factor to be 0.96 (17.714) ( 29.54 in Hg)

MMT1002 = 0.95837281 86°F + 460 MMT1011 (Rounded to 0.96)

Comment:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 6 of 19 Form ES-C-1 VERIFICATION OF COMPLETION OP-173 Section 8.2.2 Step 9.c Performance Step: 4 DETERMINE Minimum Allowed Indicated flow as follows:

315 cfm = cfm Correction Factor (from Step 8.2.2.9.b)

Standard: Determines Minimum Allowed indicated flow as 300 cfm 315 cfm = 328 cfm

______0.96____________

Correction Factor (from Step 8.2.2.9.b)

(Deviation + 2 = 326 - 330)

Comment:

OP-173 Section 8.2.2 Step 9.d Performance Step: 5 DETERMINE Maximum Allowed Indicated flow as follows:

400 cfm = cfm Correction Factor (from Step 8.2.2.9.b)

Standard: Determines Maximum Allowed indicated flow as 380.9 cfm 400 cfm = 417 cfm

______0.96____________

Correction Factor (from Step 8.2.2.9.b)

(Deviation + 2 = 415 - 419)

Comment:

Performance Step 6 Provide results to CRS.

Evaluator Cue: When results are provided by applicant, END OF JPM Stop Time:

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 7 of 19 Form ES-C-1 VERIFICATION OF COMPLETION KEY Step 9.a SOUTH / NORTH (circle one)

Step 9.b correction factor = 0.95837 (Could round to 0.96)

Step 9.c Minimum Allowed Indicated Flow = 328.6821128 (Deviation + 2 = 326 - 330) cfm Step 9.d Maximum Allowed Indicated Flow = 417.6666667 (Deviation + 2 = 415 - 419) cfm

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 8 of 19 Form ES-C-1 VERIFICATION OF COMPLETION KEY

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 9 of 19 Form ES-C-1 VERIFICATION OF COMPLETION KEY

- Denotes Critical Steps 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C Page 10 of 19 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Common Admin JPM RO / SRO A2 Determine the Minimum and Maximum Allowed Indicated Flow for MCR Ventilation OP-173.

Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 Control Room HVAC is being aligned for Post-Accident Operation in accordance with OP-173, Control Room Area HVAC System, Initial Conditions:

Section 8.2. The crew has completed steps up through step 8.2.2.8.

The CRS has directed you to perform OP-173, section 8.2.2, steps 9.a through 9.d. Assume all verifications will be completed after you submit your completed paper work.

Initiating Cue:

Current RM-23 Outside Air Intake Radiation Monitor readings and ERFIS Met Tower data are provided on the following pages.

Record your answers in the spaces provided below:

Name:

Date:

Step 9.a SOUTH / NORTH (circle one)

Step 9.b correction factor =

Step 9.c Minimum Allowed Indicated Flow = cfm Step 9.d Maximum Allowed Indicated Flow = cfm 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 RM-23 current conditions 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 RM-23 current conditions 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 ERFIS METTOWER current display 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 2016 NRC Exam Admin JPM RO/SRO A2 Rev. 2

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Harris Nuclear Plant Date of Examination: July 11, 2016 Examination Level: RO SRO Operating Test Number: 05000400/2016301 Administrative Topic Type Describe activity to be performed (see Note) Code*

Perform RCS Average Temperature Data Sheet and Determine Inverse Count Rate Ratio (1/M) (GP-004)

(JPM ADM-072-b)

Conduct of Operations M, R K/A G2.1.43 2016 NRC RO A1-1 Determine the amount of RCS inventory that will be drained from RCS during the performance of GP-008, Conduct of Operations D, P, R Draining the RCS (GP-008) (JPM ADM-070-a)

K/A G2.1.25 2016 NRC RO A1-2 Determine the Minimum and Maximum Allowed Indicated Flow for MCR Ventilation (OP-173)

Equipment Control (JPM ADM-052-c) Common M, R K/A G2.2.44 2016 NRC RO / SRO A2 Using Valve Maps And Survey Maps Determine Stay Times For A Clearance (PD-RP-ALL-0001)

(JPM ADM-057-a) Common Radiation Control M, R K/A G2.3.4 2016 NRC RO / SRO A3 NOT SELECTED FOR RO Emergency Procedures/Plan N/A 2016 NRC RO A4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (4)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (1)

(N)ew or (M)odified from bank ( 1) (3)

(P)revious 2 exams ( 1; randomly selected) (1) 05/05//2016 Rev. 2 1

2016 NRC RO Admin JPM Revision Summary 2016 NRC RO A1 Perform RCS Average Temperature Data Sheet and Determine Inverse Count Rate Ratio (1/M) (GP-004) (JPM ADM-072-b) MODIFIED K/A G2.1.43 - Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

(CFR: 41.10 / 43.6 / 45.6) RO 4.1 SRO 4.3 The plant data for various times during a plant startup will be provided to the candidate. The information provided will allow the candidate complete Attachment 2 and Attachment 3 of GP-004, Reactor Startup (Mode 3 To Mode 2). The candidate must review the plant data provided to plot the results on Attachment 3 and determine the Inverse Count Rate Ratio (1/M). The candidate must predict the reactor will NOT achieve criticality until AFTER exceeding the +500 pcm ECC control rod position.

NOTE: Modified by varying the plant data which results in a 1/M plot prediction that the Reactor will NOT achieve criticality until AFTER the + 500 pcm rod height limit has been exceed.

2016 NRC RO A1 Determine the amount of RCS inventory that will be drained from RCS during the performance of GP-008, Draining the RCS (GP-008) (JPM ADM-070-a)

Previous - from the 2013 SRO Retest Exam. Note - this JPM is a RO / SRO JPM (Randomly selected from the Admin JPM bank)

K/A G2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, ect.

(CFR: 41.10 / 43.5 / 45.12) RO 3.9 SRO 4.2 The applicant will be provided with initial plant conditions. A plant shutdown for refueling is in progress with the RCS drain down required. This is the initial drain so the SGs will be full. They will be required to calculate the amount of RCS volume in gallons to drain down with filled SGs to -70 below the flange using GP-008 Attachment 5.

2016 NRC RO A2 - (Common RO / SRO) - Determine the Minimum and Maximum Allowed Indicated Flow for MCR Ventilation (OP-173) (JPM ADM-052-c) MODIFIED K/A G2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45/12) RO 4.2 SRO 4.4 Initial conditions are that a LOCA has occurred and the Control Room HVAC is being aligned for Post-Accident Operation IAW OP-173. They are directed by the CRS to determine which Control Room emergency Outside Air intake to use, the correction factor, and the minimum and the maximum allowed indicated flow for MCR ventilation based on current radiation levels.

NOTE: Modified by changing the radiation levels and outside air temps which alters the MCR ventilation times and selection of which Control Room emergency Outside Air Intakes that should be used to supply outside air to the system. These changes make the modified JPM significantly different from the answer that is in the HNP Admin JPM bank.

05/05//2016 Rev. 2 2

2016 NRC RO Admin JPM Revision Summary 2016 NRC RO A3 - (Common RO / SRO) - Using Valve Maps And Survey Maps Determine Stay Times For A Clearance (PD-RP-ALL-0001) (JPM-ADM-057-a) - MODIFIED K/A G2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10) RO 3.2 SRO 3.7 The candidate will be provided with accumulated TEDE doses received for the year for three operators tasked with a clearance for a Letdown line leak. The will also be supplied valve maps and survey maps. They must determine the dose rates in the areas that the clearance will be hung to ensure that the maximum allowable individual stay times for each Operator will not exceed the Duke Annual Administrative dose limit while performing the clearance.

NOTE: Two modifications were performed. Increased the number of operators from two to three and changed the dose limit from 4.0 rem/year to 3.4 rem/year with an extension for a worker that had a previous dose history from another nuclear facility outside of Duke Energy where dose monitoring occurred.

The bank JPM was written to a superseded Progress Energy procedure:

DOS-NGGC-0004 Progress Energy Annual Administrative Dose Limits Section 9.3 (Rev.12)

LIMIT = 2 rem Progress Energy dose not to exceed 4 rem total dose if non- Progress Energy dose for the current year has been determined.

Current Duke Energy procedure that replaced the DOS-NGGC-0004 procedure that will be used for the 2016 JPM:

PD-RP-ALL-0001 Radiation Worker Responsibilities Section 5.2.2 Occupational Annual Dose Limits (Rev. 3) TEDE limit is 2.0 rem/year with up to 5.0 rem/year with extension.

The Radiation Protection Manager (RPM) or designee has the authority to extend an individuals limit up to 3.4 rem.

Since the procedural limits have changed the JPM had to be modified to reflect the new limit extension of 3.4 rem that can be granted by the RPM or designee. One worker in the JPM will still have a dose history containing non-Duke Energy dose that will be granted this extension.

2016 NRC RO A4 - Not selected 05/05//2016 Rev. 2 3

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Harris Nuclear Plant Date of Examination: July 11, 2016 Examination Level: RO SRO Operating Test Number: 05000400/2016301 Administrative Topic (see Note) Type Describe activity to be performed Code*

Perform RCS Average Temperature Data Sheet and Determine Inverse Count Rate Ratio (1/M) (GP-004)

(JPM ADM-072-c)

Conduct of Operations M, R K/A G2.1.43 2016 NRC SRO A1-1 During a loss of shutdown cooling, determine the time that the RCS will reach core boiling and core boil-off conditions (AOP-020, Curve Book) (JPM ADM-005-c)

Conduct of Operations D, P, R K/A G2.1.20 2016 NRC SRO A1-2 Determine the Minimum and Maximum Allowed Indicated Flow for MCR Ventilation (OP-173) (JPM ADM-052-Equipment Control c) Common M, R K/A G2.2.44 2016 NRC RO / SRO A2 Using Valve Maps And Survey Maps Determine Stay Times For A Clearance (PD-RP-ALL-0001) (JPM ADM-057-a) Common Radiation Control M, R K/A G2.3.4 2016 NRC RO / SRO A3 Classify an Event (EP-EAL) (JPM ADM-073-a)

Emergency Procedures/Plan N, R K/A G2.4.41 2016 NRC SRO A4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (5)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (1)

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

(P)revious 2 exams ( 1; randomly selected) (1) 1 05/05/2016 Rev. 2

2016 NRC SRO Admin JPM Summary 2016 NRC SRO A1 Perform RCS Average Temperature Data Sheet and Determine Inverse Count Rate Ratio (1/M) (GP-004)

(JPM ADM-072-c) - MODIFIED K/A G2.1.43 - Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

(CFR: 41.10 / 43.6 / 45.6) RO 4.1 SRO 4.3 The plant data for various times during a plant startup will be provided to the candidate. The information provided will allow the candidate complete Attachment 2 and Attachment 3 of GP-004, Reactor Startup (Mode 3 To Mode 2). The candidate must review the plant data provided to plot the results on Attachment 3 and determine the Inverse Count Rate Ratio (1/M). The candidate must predict the reactor will NOT achieve criticality until AFTER exceeding the +500 pcm ECC control rod position.

NOTE: Modified by varying the plant data which results in a 1/M plot prediction that the Reactor will NOT achieve criticality until AFTER the + 500 pcm rod height limit has been exceed.

The first part to determine the 1/M Count Rate Ration will be the same for the ROs and SROs but after the SROs determine that criticality will be achieved ABOVE the +500 pcm limit they will be asked provide the direction given to the crew.

2016 NRC SRO A1 During a loss of shutdown cooling, determine the time that the RCS will reach core boiling and core boil-off conditions. (AOP-020, Curve Book) (JPM ADM-005-c)

Previous - from the 2013 SRO Retest Exam (Randomly selected from the Admin JPM bank)

K/A G2.1.20 - Ability to interpret and execute procedure steps.

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

The SRO applicants must first determine which of the four plant curves to use (H-X-8 through H-X-11) and then calculate the time the RCS will reach core boiling and core boil-off based on the figures.

2 05/05/2016 Rev. 2

2016 NRC SRO Admin JPM Summary 2016 NRC RO SRO A2 - (Common RO / SRO) - Determine the Minimum and Maximum Allowed Indicated Flow for MCR Ventilation (OP-173) (JPM ADM-052-c) MODIFIED K/A G2.2.44 - Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45/12) RO 4.2 SRO 4.4 Initial conditions are that a LOCA has occurred and the Control Room HVAC is being aligned for Post-Accident Operation IAW OP-173. They are directed by the CRS to determine which Control Room emergency Outside Air intake to use, the correction factor, and the minimum and the maximum allowed indicated flow for MCR ventilation based on current radiation levels.

NOTE: Modified by changing the radiation levels and outside air temps which alters the MCR ventilation times and selection of which Control Room emergency Outside Air Intakes that should be used to supply outside air to the system. These changes make the modified JPM significantly different from the answer that is in the HNP Admin JPM bank.

2016 NRC RO SRO A3 - (Common RO / SRO) - Using Valve Maps And Survey Maps Determine Stay Times For A Clearance (PD-RP-ALL-0001) (JPM-ADM-057-a) - MODIFIED K/A G2.3.4 - Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10) RO 3.2 SRO 3.7 The candidate will be provided with accumulated TEDE doses received for the year for three operators tasked with a clearance for a Letdown line leak. The will also be supplied valve maps and survey maps. They must determine the dose rates in the areas that the clearance will be hung to ensure that the maximum allowable individual stay times for each Operator will not exceed the Duke Annual Administrative dose limit while performing the clearance.

NOTE: Two modifications were performed. Increased the number of operators from two to three and changed the dose limit from 4.0 rem/year to 3.4 rem/year with an extension for a worker that had a previous dose history from another nuclear facility outside of Duke Energy where dose monitoring occurred.

The bank JPM was written to a superseded Progress Energy procedure:

DOS-NGGC-0004 Progress Energy Annual Administrative Dose Limits Section 9.3 (Rev.12)

LIMIT = 2 rem Progress Energy dose not to exceed 4 rem total dose if non- Progress Energy dose for the current year has been determined.

Current Duke Energy procedure that replaced the DOS-NGGC-0004 procedure that will be used for the 2016 JPM:

PD-RP-ALL-0001 Radiation Worker Responsibilities Section 5.2.2 Occupational Annual Dose Limits (Rev. 3) TEDE limit is 2.0 rem/year with up to 5.0 rem/year with extension.

The Radiation Protection Manager (RPM) or designee has the authority to extend an individuals limit up to 3.4 rem.

Since the procedural limits have changed the JPM had to be modified to reflect the new limit extension of 3.4 rem that can be granted by the RPM or designee. One worker in the JPM will still have a dose history containing non-Duke Energy dose that will be granted this extension.

3 05/05/2016 Rev. 2

2016 NRC SRO Admin JPM Summary 2016 NRC SRO Admin JPM Summary (continued) 2016 NRC SRO A4 - Classify an Event (EP-EAL) (JPM-ADM-073-a) NEW K/A G2.4.41 - Knowledge of the emergency action level thresholds and classifications (CFR: 41.10 / 43.5 / 45.11) RO 2.9 SRO 4.6 Given a set of initial conditions and the EAL Flow Matrix, the candidate must classify the appropriate Emergency Action Level for the event in progress.

4 05/05/2016 Rev. 2

2016 NRC SRO Admin JPM Revision Summary Rev. 0 Initial Development Rev. 1 NRC D-1 Outline comments incorporated Rev. 2 Operation validation comments incorporated Rev. 3 NRC 60 day submittal comments incorporated Rev. 4 NRC Prep Week comments incorporated Rev. Final Approved for administration by NRC Region II 5 05/05/2016 Rev. 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Harris Nuclear Plant Date of Examination: July 11, 2016 Exam Level: RO SRO-I SRO-U (bolded) Operating Test No.: 05000400/2016301 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function

a. Initiate Emergency Boration following a Reactor Trip (AOP-002) (JPM CR-037-d)

A, D, L, S 1 K/A APE 024 AA1.17

b. Place Excess Letdown in Service (OP-107) (JPM-CR-211-a)

RO Only D, S 2 K/A 004 A4.06

c. Transfer to Hot Leg Recirculation (EOP ES-1.4)

(JPM-CR-066-d) A, D, EN, 3

L, S K/A EPE 011 EA1.11

d. Perform Max Rate Cooldown for a SG Tube Rupture (E-3)

(JPM-CR-283-c)

A, M, L, S 4S K/A 041 A4.08

e. Align the RHR System for ECCS Mode (OP-111)

(JPM-CR-290-a)

L, N, S 4P K/A 005 A4.01

f. Manually Align Containment Spray (EOP E-0)

(JPM CR-106-c)

A, D, EN, S 5 K/A 026 A4.01

g. Restoration of Offsite Power to Emergency Buses (EOP ECA-0.0) (JPM-CR-291-a) A, N, S 6 K/A 055 EA1.07

h. Restoring Control Room Area HVAC to Normal After a CRIS (OP-173) (JPM-CR-171-b) A, D, EN, 8

P, S K/A APE 067 AA1.05 05-6-2016 Page 1 Rev. 2

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Harris Nuclear Plant Date of Examination: July 11, 2016 Exam Level: RO SRO-I SRO-U (bolded) Operating Test No.: 05000400/2016301 In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Shift Auxiliary Feedwater Pump Suction Locally (OP-137) (JPM-IP-004-c) E, L, M, R 4S K/A 061 A1.04

j. Align UPS Instrument Bus to Bypass Power Supply (OP-156.02) (JPM-IP-254-b)

D, E 6 K/A 062 A1.03

k. Start Up A Rod Drive MG Set (OP-104) (JPM-IP-022-a)

D, L 1 K/A 001 A4.08

• All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five 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 (6 / 0 / 3)

(C)ontrol room (D)irect from bank 9/8/4 (7 / 0 / 3)

(E)mergency or abnormal in-plant 1/1/1 (2 / 0 / 2)

(EN)gineered safety feature 1 / 1 / 1 (control room system) (3 / 0 / 2)

(L)ow-Power / Shutdown 1/1/1 (6 / 0 / 3)

(N)ew or (M)odified from bank including 1(A) 2/2/1 (4 / 0 / 2)

(P)revious 2 exams 3 / 3 / 2 (randomly selected) (1 / 0 / 1)

(R)CA 1/1/1 (1 / 0 / 1)

(S)imulator The HNP 2016 License class does not have any SRO-I candidates. The count above for SRO-I has been marked 0 since there are no SRO-Is.

05-6-2016 Page 2 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary Simulator JPMs JPM a - Initiate Emergency Boration Following a Reactor Trip (AOP-002) (JPM CR-037-d)

- Alternate Path K/A APE024 AA1.17 (3.9/3.9) Ability to operate and / or monitor the following as they apply to Emergency Boration: Emergency borate control valve and indicators (CFR 41.7 / 45.5 / 45.6) RO 3.9 SRO 3.9 Evaluated position: Operator at the Controls (OATC) responsibilities.

Turnover: The plant was operating at 100% power when the A MFW pump tripped. The crew initiated a manual Reactor Trip in accordance with AOP-010, Feedwater Malfunctions.

The crew has just completed the immediate actions of EOP E-0, Reactor Trip or Safety Injection and have transitioned to ES-0.1, Reactor Trip Response Task: Perform the actions of EOP ES-0.1, Reactor Trip Response. During the performance of the procedure the candidate will determine that two rods have not fully inserted. The action for two or more rods not fully inserting on a Reactor Trip is to perform an emergency boration referring to AOP-002, Emergency Boration.

Verifiable actions: The candidate will start a Boric Acid pump and attempt to establish an emergency boration flow path by opening 1CS-278 Emergency Boric Acid Addition valve.

The valve will fail to open requiring the candidate to establish an alternate flow path by opening two other boration valves and then raise flow to >30 gpm using a FCV with the flow rate indication on a meter on the MCB.

Alternate Path - YES. 1CS-278 will fail to open requiring the candidate to utilize an alternate boration flow path and also establish a flow rate to the CSIP of > 30 gpm using FCV-122 in manual.

JPM completion: After the candidate has established an alternate boration flow path with > 30 gpm flow this JPM is complete.

05-6-2016 Page 3 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary Simulator JPMs (continued)

JPM b -Place Excess Letdown in Service (OP-107) (JPM-CR-211-a) - RO Only K/A 004 A4.06 - Ability to manually operate and/or monitor in the control room: Letdown isolation and flow control valves (CFR: 41/7 / 45.5 to 45.8) RO 3.6 / SRO 3.1 Evaluated position: Operator at the Controls (OATC) responsibilities.

Turnover: The plant is at 100%, steady state power middle of life (MOL). Normal Letdown needs to be secured for maintenance due to a problem with PCV-145. The CRS has directed the OATC to establish Excess Letdown to the VCT per OP-107, Section 8.2.

Task: Establish Excess Letdown to the VCT in accordance with OP-107, Section 8.2 Verifiable actions: The candidate will perform a valve lineup to establish a flow path from Excess Letdown to the Reactor Coolant Drain Tank. This flow path will be used to flush the lines to establish the same boron concentration as the RCS. They will then establish a valve lineup to the VCT and adjust a hand control valve to establish Excess Letdown flow at a rate that does not cause Excess Letdown temperature to exceed 174°F or pressure to exceed 150 psig. The MCB has indications and alarms for the parameters. Temperature and pressure limits prevent damage to the Excess Letdown Heat Exchanger and prevent lifting a relief in the Excess Letdown line.

Alternate Path - No - There are no failures with this JPM.

JPM completion: Excess letdown is in service and is flowing with temperature < 174°F and pressure < 150 psig in accordance with OP-107, Section 8.2.

JPM c - Transfer to Hot Leg Recirculation (EOP-ES-1.4) (JPM-CR-066-d) - SRO Upgrade -

Alternate Path K/A EPE 011 EA1.11:Ability to operate and monitor the following as they apply to a Large Break LOCA:

Long-term cooling of core.

(CFR: 41.7 / 45.6, 45.7) RO 4.2 SRO 4.2 Evaluated position: Operator at the Controls (OATC) responsibilities.

Turnover: The plant was operating at 100% power steady state middle of life (MOL) when a Large Break LOCA occurred. As a result of the LOCA an automatic Reactor Trip / Safety Injection has occurred. 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> has elapsed since the event occurred. The ESF equipment is operating and presently aligned per EOP ES-1.3, Transfer to Cold Leg Recirculation. The CRS has directed the OATC to place to transfer the Cold Leg injection line up to the Hot Leg injection line up in accordance with EOP ES-1.4, Transfer to Hot Leg Recirculation.

05-6-2016 Page 4 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary Simulator JPMs (continued)

JPM c - Transfer to Hot Leg Recirculation (continued)

Task: Align the ECCS injection lineup from Cold Leg Recirc to to Hot Leg Recirc.

Verifiable actions: The candidate will be required to perform a change valve positions to establish a flow path from the existing ECCS injection flow path lineup to the Hot Legs.

They will also have to secure the lineup due to a valve that fails to open. Additionally, this will require them to start a CSIP and open valves that had been pressure locked.

Alternate Path - YES. During the valve alignment 1SI-107, Alternate High Head SI to Hot Leg will fail to open. This failure will require the operator to use RNO actions to reopen the previously shut alternate high head SI to Cold Leg valve 1SI-52 then restart the Train A CSIP and use Attachment 1 to open 1SI-107 from a pressure locked condition.

JPM completion: When the Hot Leg lineup is completed and both the A and the B CSIPs are in operation.

JPM d - Perform Max Rate Cooldown for a SG Tube Rupture (E-3) (JPM-CR-283-c) - Modified

- Alternate Path K/A 041 A4.08 Ability to manually operate and/or monitor in the control room: Steam dump valves (CFR: 41.7 / 45.5 to 45.8) RO 3.0 SRO 3.1 Evaluated position: Balance of Plant (BOP) Operator responsibilities.

Turnover: The plant was operating at 100% power steady state Middle of Life (MOL) when a SG tube leak developed in the A SG. The tube leak has deteriorated into a tube rupture which prompted the crew to perform a manual Reactor trip / Safety Injection. The crew is presently implementing EOP-E-3, Steam Generator Tube Rupture. The CRS has directed the BOP to continue with the performance of E-3 commencing with step 28.

Task: The candidate must determine the required core exit target temperature based on the lowest ruptured SG pressure. ERFIS will NOT be available a lower conservative RCS temperature must be selected. By using RVLIS panel or active loop WR Thot. the candidate should then commence the Max Rate cooldown to the required core exit temperature utilizing the Steam Dumps.

Verifiable actions: The candidate will manipulate a switch to change Steam Dump operation from Tavg mode to Steam Pressure mode and two other switches to bypass the Steam Dump Interlock to allow continued Steam Dump operation when the RCS low temperature interlock signal closes the Steam Dumps. The Steam Dumps will fail closed in the JPM which will then require the candidate to operate the SG PORV M/A station controllers in manual to establish a Max Cooldown rate. They will then have to again use the M/A stations to secure the cooldown when the target temperature that they have determined is met.

05-6-2016 Page 5 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary Simulator JPMs (continued)

JPM d - Perform Max Rate Cooldown for a SG Tube Rupture (continued)

Alternate Path - YES - While the cooldown is in progress with the Steam Dumps the will fail shut which will stop the RCS cooldown. The candidate should be monitoring the cooldown and should diagnose that the Steam Dumps have shut by observation of status lights and RCS rate of change. The Max rate cooldown should be restarted by fully opening the two intact SG PORVs (B and C).

JPM completion: The RCS is below the target temperature and has been adequately stabilized.

MODIFICATION: Modified by changing the affected SG to A, securing ERFIS and lowering the pressures of the non-effected SGs. These changes require the candidate to use different valves to operate to accomplish the cooldown and perform a more detailed analysis to determine the required target temperature for the cooldown. The target temperature is a different value than the original version of this JPM.

JPM e - Align the RHR System for ECCS Mode (OP-111) (JPM-CR-290-a) - NEW K/A 005 A4.01 Ability to manually operate and/or monitor in the control room: Controls and indication for RHR pumps (CFR: 41.7 / 45.5 to 45.8) RO 3.6 SRO 3.4 Evaluated position: Balance of Plant (BOP) Operator responsibilities.

Turnover: The plant is in Mode 4. GP-002, Normal Plant Heatup from Cold Solid to Hot Subcritical Mode 5 to Mode 3 is in progress. Train A RHR is in cooldown mode and Train B RHR is in ECCS mode.

Task: The CRS has directed the BOP to align the A RHR train for ECCS Mode per OP-111 Section 7.2.2 starting at step 19. RHR Pump A-SA discharge temperature indication on ERFIS is < 140°F. Forced cooling of the suction line has been in progress for the last 30 minutes.

Verifiable actions: The candidate will secure a running RHR pump and then establish a valve lineup to change the flow path of the A RHR pump from a RCS cooldown lineup to the ECCS injection flow path lineup.

Alternate Path - No - There are no failures with this task JPM completion: A Train RHR ECCS Mode alignment is completed prior to RCS temperature exceeding 350°F.

05-6-2016 Page 6 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary Simulator JPMs (continued)

JPM f - Manually Align Containment Spray (EOP E-0) (JPM-CR-106-c) - Alternate Path K/A 026 A4.01 Ability to manually operate and/or monitor in the control room: CSS controls (CFR: 41.7 / 45.5 to 45.8) RO 4.5 SRO 4.3 Evaluated position: Operator at the Controls (OATC) responsibilities.

Turnover: A plant event is in progress with RCS pressure is lowering and Containment pressure is rising. An automatic Reactor Trip and SI have been occurred. EOP E-0, Reactor Trip or Safety Injection Loss Of Reactor or Secondary Coolant is being implemented. The immediate actions of E-0 have just been completed Task: Stabilize the plant following a Large Break LOCA.

Verifiable actions: Secure all RCPs after verification of SI flow > 200 gpm and RCS pressure < 1400 psig. With Containment pressure exceeding 10 psig (Phase B and Containment Spray actuation setpoints) manually actuate Containment Spray by taking both actuation switches to actuate. The switches fail therefore both Containment Spray pumps and must be manually started and the associated discharge valves must be manually opened by positioning switches for 1CT-11, 12, 50 and 88 to open.

Alternate Path - YES The automatic and manual actuation of Containment Spray will not occur requiring manual alignment of the Containment Spray system by starting the pumps and opening the discharge and Sodium Hydroxide Tank valves.

JPM completion: After the candidate has manually started both Containment Spray pumps, aligned the discharge valves to the associated pumps and flow is indicated on each train the JPM is complete.

JPM g - Restoration of Offsite Power to Emergency Buses (EOP ECA-0.0) (JPM-CR-291-a)

- SRO Upgrade - Alternate Path, NEW K/A 055 EA1.07 Ability to operate and monitor as they apply to station blackout: Restoration of power from offsite (CFR: 41.7 / 45.5 / 45.6) RO 4.3 SRO 4.5 Evaluated position: Balance of Plant (BOP) Operator responsibilities.

Turnover: The plant was operating at 100% power. A EDG is under clearance due to the generator field not flashing during OST-1013. The failure of a major line on the Duke grid resulted in the cascading trip of several units and low grid frequency. A loss of offsite power occurred. B EDG failed to start and the problem is being investigated. The crew is implementing ECA-0.0. The load dispatcher has stabilized the grid and has given permission to restore offsite power to 6.9 KV buses and to reset any tripped Start Up XFMR lockout relays (there are currently no lockout relays tripped).

05-6-2016 Page 7 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary Simulator JPMs (continued)

JPM g - Restoration of Offsite Power to Emergency Buses (continued)

Task: Restore offsite power to a (one) AC emergency bus using EOP ECA-0.0, Attachment 1.

Verifiable actions: The candidate will be manipulating electrical supply breaker switches on the MCB to restore power to the dead Emergency Bus.

Alternate Path - YES - During the lineup for power restoration on the A-SA Emergency Bus the breaker that if closed would have powered the bus fails to close.

JPM completion: Emergency Bus 1B-SB is being powered via offsite power and the 480 V breakers powering emergency equipment is energized.

JPM h - Restoring Control Room Area HVAC to Normal After a Control Room Isolation Signal (OP-173) (JPM-CR-171-b) - SRO Upgrade - Previous - 2014NRC Exam

• randomly selected from bank - Alternate Path K/A APE 067 AA1.05 - Ability to operate and / or monitor the following as they apply to the Plant Fire on Site: Plant and control room ventilation systems (CFR 41.7 / 45.5 / 45.6) RO 3.0 SRO 3.1 Evaluated position: Balance of Plant (BOP) Operator responsibilities.

Turnover: The plant is at 100% power steady state Middle of Life (MOL). A fire occurred at the Emergency Shutdown Diesel Generator during testing. The smoke from the fire caused a Control Room Ventilation Isolation signal to occur (smoke detected at the normal intake Zone 1-150). The Fire Brigade has put the fire out and the smoke has been cleared. The CRS has directed the BOP to restore the Control Room Area HVAC System to NORMAL in accordance with OP-173, Control Room Area HVAC System, Section 8.4. The initial conditions of the OP section will be satisfied and the HVAC system will be in operation per section 8.1.

Task: Restore Control Room Area HVAC to normal in accordance with OP-173, Control Room Area HVAC System, Section 8.4 Verifiable actions: The candidate must reset the Control Room Ventilation isolation signal with a switch to allow the ventilation system to be placed in its normal configuration. They will open intake valves, shut dampers and start and stop exhaust fans.

Alternate Path - YES - While shutting the Control Room Emergency Filtration Recirc dampers the running Control Room Normal Supply Fan breaker will trip open. This will result in annunciator ALB-030-6-4, Cont Room HVAC Normal Supply Fans AH-15 Low Flow

- O/L. Using the annunciator panel procedure the candidate will be required to start the standby Normal Supply Fan in accordance with OP-173 section 5.1.

JPM completion: Restoration of the Control Room Normal air supply and lineup is complete.

05-6-2016 Page 8 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary In-Plant JPMs JPM i - Shift Auxiliary Feedwater Pump Suction Locally (OP-137) (JPM-IP-004-c), Modified

- SRO Upgrade K/A 061 A1.04 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW controls including: AFW source tank level (CFR: 41.5 / 43.5) RO 3.9 SRO 3.9 NOTE: This JPM is inside the RCA.

Evaluated position: Auxiliary Operator in the RAB (AO RAB)

Turnover: The plant was operating at 100% power steady state middle of life (MOL) when a Large Break LOCA occurred. As a result of the LOCA an automatic Reactor Trip / Safety Injection has occurred. The crew is implementing EOP-ES-1.2, Post-LOCA Cooldown and Depressurization. Both Motor Driven AFW pumps started and are being used to maintain SG levels. Makeup to the Condensate Storage Tank cannot be established and CST level is decreasing (currently 9% where 10% is the minimum required level). MCC 1A35-SA and 1B35-SB are de-energized preventing operation of the Emergency Service Water valves from the MCB. The CRS has directed the operator to locally align ESW to A-SA MD AFW pumps in accordance with OP-137 Section 8.1.

Task: Locate and lineup the required ESW and AFW valves to realign the suction source of the AFW pumps from the CST to the ESW system to continue feeding the SGs which are providing a heat sink for the core.

Verifiable actions: Note- all actions will be simulated. Enter the RAB and sign onto the correct RWP. The candidate will have to locate then position valves in the ESW system that align a supply of water to the AFW pumps when the normal suction source is no longer able to be used, This task requires them to locate and change valve positions of MOVs by engaging a clutch and turning valve handwheels on both the A and B AFW trains.

Alternate Path - No - There are failures with this task but the decisions and directions to the local Operator will be provided by the Main Control Room Operators.

MODIFICATION: Modified by causing the second in series ESW valve on the A Train (1SW-123) to remain in the closed position (fail to open). The candidate will have to notify the Control Room that the lineup cannot be completed with this failure. The Control Room will then direct the candidate to stop efforts on the A Train lineup and continue with ESW suction alignment to the B AFW pump.

JPM completion: The B AFW pump suction has been aligned to the ESW suction source and the Main Control Room has been informed that the ESW lineup is completed on the 1B-SB AFW pump. This lineup will allow continued operation of the B AFW pump to provide makeup to the SGs.

05-6-2016 Page 9 Rev. 2

2016 NRC Control Room/In-Plant JPM Revision Summary In-Plant JPMs (continued)

JPM j - Align UPS Instrument Bus to Bypass Power Supply (OP-156.02) (JPM-IP-254-b)

K/A 062 A1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Effect on instrumentation and controls of switching power supplies (CFR: 41.5 / 45.5) RO 2.5 SRO 2.8 Evaluated position: Auxiliary Operator in the Turbine Building (AO TB)

Turnover: The plant is in Mode 3. Maintenance has requested that UPS Inverter S-IV be placed on bypass power source and that the Inverter be shutdown in order to inspect the high resistance contacts. The CRS has directed you to align UPS Instrument Channel IV to its Bypass power supply, and shutdown inverter S-IV, in accordance with OP-156.02, AC Electrical Distribution, Section 8.33.

Task: Place UPS Inverter SIV on bypass power source and shutdown the Inverter.

Verifiable actions: Note- all actions will be simulated. The candidate will have to locate UPS Inverter S-IV then bypass and secure a running safety inverter by manipulating the pushbuttons, switches and circuit breakers.

Alternate Path - No JPM completion: UPS Inverter S-IV is on the bypass power source and the Inverter is shutdown in accordance with OP-156.02 Section 8.33 JPM k - Start Up A Rod Drive MG Set (OP-104) (JPM-IP-022-a) - SRO Upgrade K/A 001 A4.08 Ability to manually operate and/or monitor in the control room: Mode select for CRDS; operation of rod control M/G sets and control panel (CFR 41.7 / 45.5 / 45.8) RO 3.7 / SRO 3.4 Evaluated position: Auxiliary Operator (AO) responsibilities Turnover: With the plant in Hot Standby at 550°F and PRZ pressure of 2240 psig. Both Rod Drive MG sets are secured. The CRS has directed that the 1A and 1B MG sets be placed into operation in accordance with OP-104, Rod Control System, Section 5.1.

Task: At the Rod Drive MG sets and control panels place the 1A MG set in operation in accordance with OP-104.

Verifiable actions: Note- all actions will be simulated. The candidate will reset ground protection and directional overcurrent relays, start the 1A MG set, flash the MG set generator field, adjust the voltage for 260 VAC and close the generator circuit breaker.

Alternate Path - No JPM completion: After the Motor Generator voltage is verified following Reactor trip breakers closed at the MCB which will load the motor-generator.

05-6-2016 Page 10 Rev. 2

Appendix C Page 1 of 12 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 061003H104 Task

Title:

Shift AFW Pump Suction Locally JPM No.: 2016 HNP NRC Exam In-Plant JPM i K/A

Reference:

061 K4.01 4.1 RO 4.2 SRO ALTERNATE PATH - NO Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: X Actual Performance:

Classroom Simulator Plant X READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

A LOCA has occurred and EOP-ES-1.2, Post-LOCA Cooldown and Depressurization, is being performed.

Initial Conditions:

AFW Pumps A-SA and B-SB are being used to maintain SG levels.

Makeup to the CST cannot be established and CST level is lowering.

MCCs 1A35-SA and 1B35-SB are deenergized.

• CST level has decreased to 9%.

• The CRS has directed you to locally align ESW to A-SA MD AFW pumps IAW OP-137 Section 8.1.

Initiating Cue:

• The A ESW header is in service.

• Steps 8.1.2.1.a - f have been completed by the MCR.

• You are to start at step 8.1.2.1.g and continue At this time provide the candidate with a copy of OP-137 Evaluator Note:

Section 8.1 Evaluator Note: Expect that the entry and exit from the RCA and take a minute on location of the JPM will add time for completion of this JPM.

2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 2 of 12 Form ES-C-1 Worksheet Task Standard: AFW Pump B-SB suction is aligned to ESW.

Required Materials: In-plant required PPE General

References:

OP-137, Auxiliary Feedwater System, Section 8.1 Rev. 42 Handout: OP-137, Rev. 42, Prerequisites, P&Ls, and Section 8.1, Using Emergency Service Water System as a Backup Source of Water to Auxiliary Feedwater System Time Critical Task: No Validation Time: 15 minutes SIMULATOR SETUP N/A This is an In-Plant JPM Critical Task Justification Required to be repositioned to isolate flow path and prevent water from Step 5 the ESW header from entering the RAB floor drain system.

Required to be repositioned to establish a flow path from the ESW Step 6 header in order to maintain the AFW system a functional heatsink Required to be repositioned to establish a flow path from the ESW Step 7 header in order to maintain the AFW system a functional heatsink 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 3 of 12 Form ES-C-1 Worksheet Simplified AFW suction layout 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 4 of 12 Form ES-C-1 PERFORMANCE INFORMATION BEFORE YOU START THIS JPM INPLANT JPM SAFETY CONSIDERATIONS:

CAUTION: EQUIPMENT MAY AUTO START OR MAY BE ENERGIZED

- SIMULATE ONLY - DO NOT OPERATE ANY ACTUAL PLANT EQUIPMENT!!!

Before entering the performance location of this JPM, ensure you AND the candidate have the proper PPE for the area you are going to go to or will travel through to get there.

Avoid contacting any plant equipment.

Maintain 6 from touching any equipment during the JPM.

Follow ALARA practices in the RCA.

Do NOT remove ladders from their storage locations. Simulate obtaining and using a ladder if one would be needed during the actual performance of this task.

NOTE: Add one minute for Take a Minute Core 4 checks.

START TIME:

OP-137 Section 8.1.1 Initial Conditions Performance Step: 1 1. Condensate storage tank unavailable or level < 10%

2. Service Water System in operation per OP-139 Standard: Reviews initial conditions and initials completion (Initial conditions were met with the initiating cues)

Comment:

A MDAFW Pump OP-137 Section 8.1 Step 1.g Performance Step: 2 SHUT 1SW-122, AFWP 1A SW Drain Isol.

Standard: Locates 1SW-122, and demonstrates shutting valve.

1SW-122 handwheel has rotated in the clockwise direction Evaluator Cue:

and the valve stem has come to a hard stop.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 5 of 12 Form ES-C-1 PERFORMANCE INFORMATION To engage the handwheels for the Service Water MOVs, the operator must turn the small handle (UNDER the body of 1SW-121SA and on the side of 1SW-123SA). The valve handwheel will NOT be engaged and the valve will NOT open Evaluator Note:

unless the small handle is turned to engage. Valve position indication is located on the top of the valve next to the wall. The operator should check this indication to ensure the valve has OPENED.

OP-137 Section 8.1 Step 1.h Performance Step: 3 OPEN 1SW-121SA, SW HEADER A TO AUX FW MOTOR PMP A-SA.

Standard: Locates 1SW-121SA, engages handwheel, and opens valve.

(NOTE: You can cue them by the position arrow IF they are checking the position with the arrow.)

CUE: (If handwheel is properly engaged) 1SW-121SA handwheel has rotated in the counter clockwise direction and has come to a hard stop.

Evaluator Cue:

If the device to engage the handwheel is NOT turned then CUE: 1SW-121SA handwheel has rotated in the counter clockwise direction and the handwheel spins freely.

(provide the cue for when the handwheel is properly engaged when the candidate engages the handwheel first)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 6 of 12 Form ES-C-1 PERFORMANCE INFORMATION OP-137 Section 8.1 Step 1.i Performance Step: 4 OPEN 1SW-123SA, SW HEADER A TO AUX FW MOTOR PMP A-SA.

Standard: Locates 1SW-123SA, engages handwheel, and attempts to open the valve.

CUE: The handwheel will not turn, the valve is mechanically bound.

Evaluator Cue: (NOTE: You can cue them by the position arrow IF they are checking the position with the arrow.)

The arrow is still pointing to closed - shut.

Standard: Contacts MCR and informs them 1SW-123 SA will not open.

MCR acknowledges 1SW-123SA will not open. The MCR is directing you to return to 1SW-121SA and shut 1SW-121SA.

Evaluator Cue:

Contact the MCR after 1SW-121SA is shut for additional directions.

Standard: Returns to 1SW-121SA, turns the handwheel and shuts 1SW-121SA.

(NOTE: They may engage the handwheel by using the clutch. This is not necessary but there is no harm in doing so. You can also cue them by the position arrow IF they are checking the position with the arrow.)

Evaluator Cue:

CUE: (After the handwheel is positioned) 1SW-121SA handwheel has rotated in the clockwise direction and has come to a hard stop.

Standard: Contacts MCR and informs them 1SW-121SA has been shut.

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 7 of 12 Form ES-C-1 PERFORMANCE INFORMATION Acknowledge report of 1SW-121SA closure.

State: Using Time Compression The MCR operators have now reopened the A Train Containment Fan Cooler Service Water isolation valves and Evaluator Cue: isolated the B Train Containment Fan Cooler Service Water Isolation valves.

Section 8.1.2 steps 2.a - f have been completed. You are being directed to continue with section 8.1.2.2 steps g, h and i to align Service Water to supply AFW pump 1B-SB from ESW header B. Call the MCR when these steps are completed.

Standard: Repeats back communications Complete third leg of communication:

IF commination is repeated back correct then cue:

Evaluator Cue: Thats correct OR re-read directions until communications are correct.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 8 of 12 Form ES-C-1 PERFORMANCE INFORMATION B MDAFW Pump alignment OP-137 Section 8.1 Step 2.g Performance Step: 5 SHUT 1SW-131, AFWP 1B SW Drain Isol.

Standard: Locates 1SW-131, and demonstrates shutting valve.

1SW-131 handwheel has rotated in the clockwise direction Evaluator Cue:

and the valve stem has come to a hard stop.

Comment:

To engage the handwheels for the Service Water MOVs the operator must turn the small handle (UNDER the body of 1SW-130 and on the side of 1SW-132). The valve handwheel Examiners Note: will NOT be engaged and the valve will NOT open unless the small handle is turned to engage. Valve position indication is located on the top of the valve next to the wall. The operator should check this indication to ensure the valve has OPENED.

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 9 of 12 Form ES-C-1 PERFORMANCE INFORMATION OP-137 Section 8.1 Step 2.h Performance Step: 6 OPEN 1SW-130SB, SW HEADER B TO AUX FW MOTOR PMP B-SB.

Standard: Locates 1SW-130SB, engages handwheel, and opens valve.

(NOTE: You can cue them by the position arrow IF they are checking the position with the arrow.)

CUE: (If handwheel is properly engaged) 1SW-130SB handwheel has rotated in the counter clockwise direction and has come to a hard stop.

Evaluator Cue: If the device to engage the handwheel is NOT turned then CUE: 1SW-130SB handwheel has rotated in the counter clockwise direction and the handwheel spins freely.

(provide the cue for when the handwheel is properly engaged when the candidate engages the handwheel first)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 10 of 12 Form ES-C-1 PERFORMANCE INFORMATION OP-137 Section 8.1 Step 2.i Performance Step: 7 OPEN 1SW-132SB, SW HEADER B TO AUX FW MOTOR PMP B-SB.

Standard: Locates 1SW-132SB, engages handwheel, and opens valve.

(NOTE: You can cue them by the position arrow IF they are checking the position with the arrow.)

CUE: (If handwheel is properly engaged) 1SW-132SB handwheel has rotated in the counter clockwise direction and has come to a hard stop.

Evaluator Cue: If the device to engage the handwheel is NOT turned then CUE: 1SW-132SB handwheel has rotated in the counter clockwise direction and the handwheel spins freely.

(provide the cue for when the handwheel is properly engaged when the candidate engages the handwheel first)

Comment:

OP-137 Section 8.1 Step 2.j Performance Step: 8 MONITOR AFW system parameters to ensure proper operation.

Standard: Notifies Control Room ESW is aligned to both MDAFW Pumps 1A-SA and 1B-SB and to monitor for proper operation.

Control Room acknowledges that ESW is aligned to AFW Pump 1B-SB.

Evaluator Cue:

Evaluation on this JPM is complete.

END OF JPM Comment:

STOP TIME:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 11 of 12 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam In-Plant JPM i Shift AFW Pump Suction Locally IAW OP-137 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 BEFORE YOU START THIS JPM INPLANT JPM SAFETY CONSIDERATIONS:

CAUTION: EQUIPMENT MAY AUTO START OR MAYBE ENERGIZED

- SIMULATE ONLY - DO NOT OPERATE ANY ACTUAL PLANT EQUIPMENT!!!

Before entering the performance location of this JPM, ensure you AND the examiner have the proper PPE for the area you are going to go to or will travel through to get there.

Avoid contacting any plant equipment.

Maintain 6 from touching any equipment during the JPM.

Follow ALARA practices in the RCA.

Do NOT remove ladders from their storage locations. Simulate obtaining and using a ladder if one would be needed during the actual performance of this task.

A LOCA has occurred and EOP-ES-1.2, Post-LOCA Cooldown and Depressurization, is being performed.

AFW Pumps A-SA and B-SB are being used to maintain SG levels.

Initial Conditions:

Makeup to the CST cannot be established and CST level is lowering.

MCCs 1A35-SA and 1B35-SB are deenergized.

• CST level has decreased to 9%.

• The MCR has directed you to locally align ESW to A-SA MD AFW pumps IAW OP-137 Section 8.1.

Initiating Cue:

• The A ESW header is in service.

• Steps 8.1.2.1.a - f have been completed by the MCR.

• You are to start at step 8.1.2.1.g and continue 2016 HNP NRC Exam In-Plant JPM i Rev. 2

Appendix C Page 1 of 10 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301089H401 Task

Title:

Place an UPS Instrument Bus On JPM No.: 2016 HNP NRC Exam Bypass Power and Shutdown the In-Plant JPM j Associated Inverter K/A

Reference:

062 A1.03 RO 2.5 SRO 2.8 ALTERNATE PATH - NO Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: X Actual Performance:

Classroom Simulator Plant X READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

The plant is in Mode 3. Maintenance has requested that UPS Inverter Initial Conditions: SIV be placed on bypass power source and that the Inverter be shutdown in order to inspect the high resistance contacts.

The CRS has directed you to align UPS Instrument Channel IV to its Bypass power supply, and shutdown inverter SIV, in accordance with OP-156.02, AC Electrical Distribution, Section 8.33.

Initiating Cue:

Contact the MCR when you have the Inverter on the Bypass power source and the Inverter is shutdown.

DO NOT READ Provide the candidate with a copy of OP-156.02 Section 8.33. Allow TO CANDIDATE them a few minutes for a procedure / pre-job review. The start time will Examiner: begin when the candidate is at the Inverter.

2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 2 of 10 Form ES-C-1 Worksheet Task Standard: Instrument Bus SIV is powered from the Bypass power supply and the inverter is shutdown.

Required Materials: In-plant required PPE General

References:

AOP-024, Loss of Uninterruptible Power Supply, Rev. 56 OP-156.02, AC Electrical Distribution, Section 8.33, Rev. 141 Handout: OP-156.02, Rev. 141, Prerequisites, P&Ls, and Section 8.33, Bypass Source Operation for Safety 7.5 KVA Inverters Time Critical Task: No Validation Time: 5 minutes SIMULATOR SETUP N/A This is an In-Plant JPM Critical Task Justification Critical to depress the bypass to load pushbutton otherwise you would be Step 3 unable to place the place the UPS Instrument Bus on bypass power and shutdown the associated inverter in accordance with plant procedures.

Critical to place the manual bypass switch to the bypass to load positions otherwise you would be unable to place the place the UPS Instrument Bus Step 4 on bypass power and shutdown the associated inverter in accordance with plant procedures.

Critical to open the inverter output circuit breaker switch otherwise you would be unable to place the place the UPS Instrument Bus on bypass Step 5 power and shutdown the associated inverter in accordance with plant procedures..

Critical to open the battery input circuit breaker otherwise you would be Step 6 unable to place the place the UPS Instrument Bus on bypass power and shutdown the associated inverter in accordance with plant procedures.

Critical to open the rectifier AC input circuit breaker otherwise you would be Step 7 unable to place the place the UPS Instrument Bus on bypass power and shutdown the associated inverter in accordance with plant procedures..

Critical to open the bypass source AC input circuit breaker otherwise you would be unable to place the place the UPS Instrument Bus on bypass Step 8 power and shutdown the associated inverter in accordance with plant procedures.

2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 3 of 10 Form ES-C-1 PERFORMANCE INFORMATION BEFORE YOU START THIS JPM INPLANT JPM SAFETY CONSIDERATIONS:

CAUTION: EQUIPMENT MAY AUTO START OR MAY BE ENERGIZED

- SIMULATE ONLY - DO NOT OPERATE ANY ACTUAL PLANT EQUIPMENT!!!

Before entering the performance location of this JPM, ensure you AND the candidate have the proper PPE for the area you are going to go to or will travel through to get there.

Avoid contacting any plant equipment.

Maintain 6 from touching any equipment during the JPM.

Follow ALARA practices in the RCA.

Do NOT remove ladders from their storage locations. Simulate obtaining and using a ladder if one would be needed during the actual performance of this task.

NOTE: Add one minute for Take a Minute Core 4 checks.

START TIME:

Performance Step: 1 OP-156.02 Section 8.33, NOTE Prior to step 1 NOTE: When aligned to the Manual Bypass:

• Tech Spec 3.8.3.1 ACTION c applies. (Modes 1, 2, 3, and 4)

• Tech Spec 3.8.3.2 ACTION applies. (Modes 5 an 6)

Initial Conditions:

1. It is desired to place an inverter on its bypass source.

OR

2. An inverter is on its bypass source and normal operation is desired.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Initials condition 1 and N/As condition 2 IF the candidate calls the MCR CRS about the Tech Spec implication found in the NOTE prior to the Initial Conditions Evaluator Note: then acknowledge the information.

Initial conditions: The initiating conditions were to place the inverter on its bypass source.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 4 of 10 Form ES-C-1 PERFORMANCE INFORMATION OP-156.02 Section 8.33, Step 1 Performance Step: 2 To place the inverter on the bypass source, PERFORM the following:

a. VERIFY the IN SYNC light is lit.

Standard: Identifies the IN SYNC light and verifies it is lit Evaluator Cue: The IN SYNC light is lit Comment:

Note: The procedure step below reads Bypass to Load and the label for the switch reads Inverter bypass Source To Load. The critical step is to depress the pushbutton. A non -critical action is to STOP and review the mismatch with by contacting the CRS prior to proceeding. There will be a PRR written against this discrepancy at the conclusion Evaluator Note: of the exam. The step is considered a soft match.

See page 8 for more details.

The candidate may stop and contact MCR for the label mismatch. IF so then cue that this has been identified and although the labeling is a soft match it has been verified as correct and you may continue.

OP-156.02 Section 8.33, Step 1 continued Performance Step: 3 b. DEPRESS the Bypass to Load pushbutton.

Standard: Locates the Bypass to Load pushbutton and depresses the pushbutton (may simulate lifting plastic button cover).

Evaluator Cue: The Bypass to Load pushbutton has been depressed.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 5 of 10 Form ES-C-1 PERFORMANCE INFORMATION OP-156.02 Section 8.33, Step 1 continued Performance Step: 4 c. TRANSFER Manual Bypass Switch to BYPASS TO LOAD position Standard: Locates the transfer switch and places the manual Bypass switch to the BYPASS TO LOAD position (turns switch to right). May simulate lifting plexiglass cover for these steps.

The manual Bypass switch is in the Bypass TO LOAD Evaluator Cue:

position.

Comment:

OP-156.02 Section 8.33, Step 1 continued Performance Step: 5 d. IF shutdown of the inverter is desired, THEN PERFORM the following:

1. OPEN Inverter Output circuit breaker.

Standard: Locates the Inverter Output circuit breaker switch and takes switch to OPEN.

Evaluator Cue: The Inverter Output circuit breaker is OPEN.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 6 of 10 Form ES-C-1 PERFORMANCE INFORMATION OP-156.02 Section 8.33, Step 1 continued Performance Step: 6 2. OPEN Battery Input circuit breaker.

Standard: Locates the Battery Input circuit breaker and takes switch to the OPEN position.

Evaluator Cue: The Battery Input circuit breaker is OPEN.

Comment:

OP-156.02 Section 8.33, Step 1 continued Performance Step: 7 3. OPEN Rectifier AC Input circuit breaker.

Standard: Locates the Rectifier AC Input circuit breaker and takes switch to the OPEN position.

Evaluator Cue: The Rectifier AC Input circuit breaker is OPEN.

Comment:

OP-156.02 Section 8.33, Step 1 continued Performance Step: 8 4. OPEN Bypass Source AC Input circuit breaker.

Standard: Locates the Bypass Source AC Input circuit breaker and takes switch to the OPEN position.

Evaluator Cue: The Bypass Source AC Input circuit breaker is OPEN.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 7 of 10 Form ES-C-1 PERFORMANCE INFORMATION OP-156.02 Section 8.33, Step 1 continued Performance Step: 9 Communicate task completion to CRS.

Standard: Contacts MCR and informs the CRS that the SIV Inverter is on Bypass power and shutdown.

The CRS acknowledges communication that Inverter SIV is Evaluator Cue: on Bypass power and is shutdown.

END OF JPM Comment:

After the Bypass Source AC input circuit breaker is open Terminating Cue: and communications are completed. Evaluation of this JPM is complete.

STOP TIME:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 8 of 10 Form ES-C-1 PERFORMANCE INFORMATION In accordance with AD-HU-ALL-0004, Section 5.2.22

22. For plant and equipment labels:

a. When reading equipment component identification labels, there are typically two pieces of information provided. The first is the Equipment ID Number (e.g., 1CS-12 is the equipment ID for a valve). The second is the Equipment Description (e.g., Unit 1 Upper Surge Tank Dome Inlet Isol).

b. Concerning Equipment ID Numbers, during procedure performance a 'hard match' (which is an exact character for character match) is required when reading Equipment ID Numbers in a procedure step or safety tag and comparing to the component identification label in the plant. If a hard match does not exist, STOP and contact supervision.

c. Concerning Equipment Description, during procedure performance, a 'soft match' can be used when exact wording is not the same between a procedure or safety tag and the component label Equipment Description. Correct Component Verification must be ensured. Slight variations (e.g., abbreviation, sequencing of wording, spacing, and dashes) are permitted as long as no ambiguity exist in proper identification of the component and the Equipment ID (if provided - always provided for a safety tag but may not be provided in a procedure step) is a hard match. A few examples are:

(1) A Procedure step states: "Open the breaker for U-1 Seal Oil Vacuum Pump Motor." An acceptable 'soft match' on the component label would read "U-1 SO Vac Pmp Mtr Bkr".

(2) A Red Safety Tag has been printed for the following tag out: "Rack out 2A HDP Bkr." An acceptable 'soft match' on the component label would read "2A Heater Drain Pump Bkr" or "Heater Drain Pump 2A Bkr".

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 9 of 10 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam In-Plant JPM j Place an UPS Instrument Bus On Bypass Power and Shutdown the Associated Inverter IAW OP-156.02 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1 BEFORE YOU START THIS JPM INPLANT JPM SAFETY CONSIDERATIONS:

CAUTION: EQUIPMENT MAY AUTO START OR MAYBE ENERGIZED

- SIMULATE ONLY - DO NOT OPERATE ANY ACTUAL PLANT EQUIPMENT!!!

Before entering the performance location of this JPM, ensure you AND the examiner have the proper PPE for the area you are going to go to or will travel through to get there.

Avoid contacting any plant equipment.

Maintain 6 from touching any equipment during the JPM.

Follow ALARA practices in the RCA.

Do NOT remove ladders from their storage locations. Simulate obtaining and using a ladder if one would be needed during the actual performance of this task.

The plant is in Mode 3. Maintenance has requested that UPS Inverter Initial Conditions: SIV be placed on its bypass power source, and that the Inverter be shutdown in order to inspect the high resistance contacts.

The CRS has directed you to align UPS Instrument Channel IV to its Bypass power supply, and shutdown inverter SIV, in accordance with OP-156.02, AC Electrical Distribution, Initiating Cue: Section 8.33.

Contact the MCR when you have the Inverter on the Bypass power source and the Inverter is shutdown.

2016 HNP NRC Exam In-Plant JPM j Rev. 2

Appendix C Page 1 of 15 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 001001H104 Task

Title:

Start Up A Rod Drive MG Set JPM No.: 2016 HNP NRC Exam In-Plant JPM k K/A

Reference:

001 A4.08 RO 3.7 SRO 3.4 ALTERNATE PATH - No Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: X Actual Performance:

Classroom Simulator Plant X READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The Reactor is shut down with Tavg at 550°F and PRZ pressure Initial Conditions: of 2240 psig.

• Both Rod Drive MG sets are secured.

• The CRS is directing you to place the 1A Rod Drive MG set in Initiating Cue: service per OP-104 section 5.1

• The initial conditions are completed DO NOT READ Provide the candidate with a copy of OP-104 Section 5.1. Allow them a TO CANDIDATE few minutes for a procedure / pre-job review. The start time will begin Examiner: when the candidate is at the Rod Drive MG Set.

2016 NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 2 of 15 Form ES-C-1 Worksheet Task Standard: 1A and 1B Rod Drive MG set operating and supplying power to the Rod Control System.

Required Materials: In-plant required PPE General

References:

OP-104, Rod Control System, Section 5.1, Rev. 39 Handout: OP-104, Rev. 39, Prerequisites, P&Ls, and Section 5.1, Single Rod Drive MG Startup with step 5.1.1 initial conditions complete Time Critical Task: No Validation Time: 15 minutes SIMULATOR SETUP N/A This is an In-Plant JPM Critical Step Justification Step 13 Must close the motor circuit breaker to start the MG Set motor.

Step 15 Must flash the field to obtain generator voltage then adjust the Generator line voltage to 240 VAC to verify that the generator is operational.

Step17 Must adjust generator voltage to 260 VAC to obtain the normal operational voltage in preparations to load the MG set.

Step 18 Must close the generator circuit breaker to allow the Reactor Trip breakers to have power to be closed.

2016 NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 3 of 15 Form ES-C-1 PERFORMANCE INFORMATION BEFORE YOU START THIS JPM INPLANT JPM SAFETY CONSIDERATIONS:

CAUTION: EQUIPMENT MAY AUTO START OR MAY BE ENERGIZED

- SIMULATE ONLY - DO NOT OPERATE ANY ACTUAL PLANT EQUIPMENT!!!

Before entering the performance location of this JPM, ensure you AND the candidate have the proper PPE for the area you are going to go to or will travel through to get there.

Avoid contacting any plant equipment.

Maintain 6 from touching any equipment during the JPM.

Follow ALARA practices in the RCA.

Do NOT remove ladders from their storage locations. Simulate obtaining and using a ladder if one would be needed during the actual performance of this task.

NOTE: Add one minute for Take a Minute Core 4 checks.

START TIME:

The Rod Drive MG Cabinets are oriented such that the 1B Evaluator NOTE: MG is on the left side AND the 1A MG is on the right side when facing the cabinets.

Obtains OP-104 section 5.1.1 Initial Conditions Performance Step: 1 NOTE: Both Rod Drive MG Set output breakers must be racked in before either output breaker will close.

1. Rod Control System lined up per Rod Control System Electrical Lineup Prestart Checklist, Attachment 1.

2. Shunt Trip Test per MST-I0072 or MST-I0001 (MST-I0073 or MST-I0320) has been performed on the Reactor Trip Breakers or is to be performed prior to making the Rod Control System capable of withdrawing rods.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Evaluator Cue: Initial conditions have been satisfied.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 4 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 NOTE prior to step 1 Performance Step: 2 NOTE: The Rod Drive MG Set SYNCHRONIZING switch handle is normally kept in the Operations Key Locker Standard: Operator reads and place-keeps at any procedure note/caution (initials, checks or circle/slash)

Comment:

OP-104 section 5.1.2 step 1 Performance Step: 3 OBTAIN the handle for the Rod Drive MG Set SYNCHRONIZING switch Standard: Obtains handle or describes the method for obtaining the synchronizing switch handle.

• Get key from CRS for key locker.

• Obtain handle from key locker.

Comment:

OP-104 section 5.1.2 NOTE prior to step 2 Performance Step: 4 B component equipment nomenclature is in parentheses Standard: Operator reads and place-keeps at any procedure note/caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 5 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 step 2 Performance Step: 5 At the ROD PWR SUPPLY CNTL CABINETs for GENERATOR NO. 1A (1B) CUB. 2 (1) DEPRESS the RELAY FLAG RESET button Standard: Locates and depresses the RELAY FLAG RESET button Evaluator Cue: RELAYS FLAG RESET P/B DEPRESSED.

Comment:

OP-104 section 5.1.2 NOTE prior to step 3 Performance Step: 6 Failure of the push rods to reset the flag would indicate that the corresponding relay may not have reset.

Standard: Operator reads and place-keeps at any procedure note/caution (initials, checks or circle/slash)

Comment:

OP-104 section 5.1.2 step 3 Performance Step: 7 On the front of GENERATOR NO. 1B CUB. 1, LIFT the mechanical push rod to reset the red flag on the GROUND PROTECTIVE RELAY Standard: Locates and lifts the mechanical push rods to reset the RED FLAG on the GROUND PROTECTIVE RELAY Evaluator Cue: GROUND PROTECTIVE RELAY red flag is reset.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 6 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 step 4 Performance Step: 8 On the front of the GENERATOR NO. 1A CUB. 2 LIFT BOTH mechanical push rods to reset the red flags on the DIRECTIONAL OVERCURRENT A and C relays.

Standard: Locates and lifts the mechanical push rod to reset A and C DIRECTIONAL OVERCURRENT RELAYS A and C DIRECTIONAL OVERCURRENT RELAYS BOTH Evaluator Cue:

lifted and red flags are reset.

Comment:

OP-104 section 5.1.2 NOTE prior to step 5 Performance Step: 9 The potentiometer pointer is the thin white line at the upper left corner of the potentiometer.

Standard: Operator reads and place-keeps at any procedure note/caution (initials, checks or circle/slash)

The potentiometer pointer is at the 11 oclock position on Evaluator NOTE: the potentiometer meter face. This indication may be hard to see without a flashlight.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 7 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 step 5 Performance Step: 10 POSITION Generator No. 1A VOLTAGE ADJUST potentiometer to the mid position.

Standard: Locates and adjusts VOLTAGE ADJUST potentiometer to mid position Evaluator Cue: VOLTAGE ADJUST potentiometer is in mid position Comment:

OP-104 section 5.1.2 NOTE prior to step 6 Performance Step: 11 Only one handle is used between the two synchronizing selector switches.

Standard: Operator reads and place-keeps at any procedure note/caution (initials, checks or circle/slash)

Comment:

OP-104 section 5.1.2 step 6 Performance Step: 12 VERIFY Generator No. 1A SYNCHRONIZING Selector Switch is in OFF.

Standard: Verifies Generator No. 1A SYNCHRONIZING selector switch is in the OFF position.

SYNCHRONIZING selector switch is inserted and in the OFF Evaluator Cue:

position.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 8 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 step 7 Performance Step: 13 PLACE CS-495 MOTOR CIRCUIT BREAKER CONTROL SWITCH A in CLOSE Standard: Locates and places the motor circuit breaker control switch No.

1A to the CLOSE position.

Motor circuit breaker No. 1A control switch is in the CLOSE Evaluator Cue: position AND the MG set is operating.

(IF checked): Red light is on, green light is off.

Comment:

OP-104 section 5.1.2 step 8 Performance Step: 14 ALLOW 15 seconds for the Rod Drive MG to obtain full rated speed Standard: Waits 15 seconds If desired using time compression report 15 seconds has Evaluator Cue:

elapsed.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 9 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 step 9 Performance Step: 15 WHILE OBSERVING the GENERATOR LINE VOLTS meter, DEPRESS AND HOLD the GEN. FIELD FLASH pushbutton for GENERATOR NO. 1A CUB. 2. The Rod Drive MG voltage should rise to 230 to 260 VAC. After the voltage has stabilized, release the GEN. FIELD FLASH pushbutton Standard: Locates and depresses the GEN. FIELD FLASH pushbutton until voltage has stopped increasing and then releases GEN. FIELD FLASH pushbutton Generator line voltage is steady at 240 VAC on the Evaluator Cue:

Generator Line Voltmeter.

Comment:

OP-104 section 5.1.2 step 10 Performance Step: 16 CHECK the range of voltage control, 230 to 300 VAC, by POSITIONING to minimum then maximum Generator No. 1A VOLTAGE ADJUST potentiometer.

Standard: Locates and positions GENERATOR NO. 1A VOLTAGE ADJUST potentiometer from minimum to maximum to verify 230 to 300 VAC( turns potentiometer dial left to right)

Evaluator Cue: 230 to 300 VAC range verified on voltage control.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 10 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 step 11 Performance Step: 17 ADJUST the Generator No. 1A VOLTAGE ADJUST potentiometer for 260 VAC.

Standard: Adjusts the generator VOLTAGE ADJUST potentiometer for 260 VAC. ( turns potentiometer dial left or right until 260 is reached)

Evaluator Cue: Generator VOLTAGE ADJUST potentiometer set for 260 VAC Comment:

OP-104 section 5.1.2 step 12 Performance Step: 18 PLACE 1IC-E171:017 GENERATOR CIRCUIT BREAKER CONTROL SWITCH A in CLOSE.

Standard: Locates and places GENERATOR CIRCUIT BREAKER NO. 1A control switch in the CLOSE position and obverses status light indication GENERATOR CIRCUIT BREAKER NO. 1A control switch is Evaluator Cue: in the CLOSE position Red light is lit, green light is off.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 11 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 NOTE prior to step 13 Performance Step: 19 NOTE:

- When the Reactor Trip Breaker closes, Rod Drive MG voltage will drop slightly as load is increased.

Standard: Operator reads and place-keeps at any procedure note (initials, checks or circle/slash)

Comment:

OP-104 section 5.1.2 Caution prior to step 13 Performance Step: 20 CAUTION:

- Before closing the Reactor Trip Breakers, the Turbine must be manually tripped to prevent the possibility of the Turbine latching automatically.

- Before closing the Reactor Trip Breakers, the Main Feed Regulating Valve Controllers must be verified to be in Manual and at 0% demand.

Standard: Operator reads and place-keeps at any procedure caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 12 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-104 section 5.1.2 steps 13-15 (MCR)

Performance Step: 21 13. PERFORM the following:

a. At the MCB, INITIATE a manual Turbine trip

b. VERIFY the following Main FW Regulating Valve Controllers are in Manual and at 0% demand (minimum):

• 1FW-133, MAIN FW A REGULATOR FK-478

• 1FW-249, MAIN FW B REGULATOR FK-488

• 1FW-191, MAIN FW C REGULATOR FK-498

14. VERIFY OST-1054, Permissives P-6 and P-10 Verification, is within periodicity prior to closing Reactor Trip Breakers.

15. At the MCB, CLOSE the Reactor Trip Breakers to load the Rod Drive MG.

Standard: Contacts and request the MCR to perform OP-104 section 5.1.2 steps 13-15.

Candidate might state hearing Reactor Trip Breakers Evaluator NOTE:

operating.

MCR to [Candidate], the Reactor Trip Breakers have been closed. OP-104 section 5.1.2 steps 13-15 are complete.

Evaluator Cue:

Proceed with the next step to verify generator output voltage.

Comment:

OP-104 section 5.1.2 step 16 Performance Step: 22 VERIFY the Generator No. 1A GENERATOR LINE VOLTS is 260 VAC.

Standard: Verifies Generator line voltage is 260 VAC on the Generator Line Voltmeter Evaluator Cue: Generator Line Voltmeter reads 260 VAC.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 13 of 15 Form ES-C-1 PERFORMANCE INFORMATION After the Motor Generator voltage is verified following Reactor trip breakers closed at the MCB which will load the motor-generator.

Evaluator Cue:

Evaluation on this JPM is complete.

END OF JPM STOP TIME:

- Denotes Critical Steps 2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 14 of 15 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam JPM k Start Up A Rod Drive MG Set OP-104 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam In-Plant JPM k Rev. 2

Appendix C Page 15 of 15 Form ES-C-1 JPM CUE SHEET BEFORE YOU START THIS JPM INPLANT JPM SAFETY CONSIDERATIONS:

CAUTION: EQUIPMENT MAY AUTO START OR MAYBE ENERGIZED

- SIMULATE ONLY - DO NOT OPERATE ANY ACTUAL PLANT EQUIPMENT!!!

Before entering the performance location of this JPM, ensure you AND the examiner have the proper PPE for the area you are going to go to or will travel through to get there.

Avoid contacting any plant equipment.

Maintain 6 from touching any equipment during the JPM.

Follow ALARA practices in the RCA.

Do NOT remove ladders from their storage locations. Simulate obtaining and using a ladder if one would be needed during the actual performance of this task.

• The Reactor is shut down with Tavg at 550°F and PRZ pressure Initial Conditions: of 2240 psig.

• Both Rod Drive MG sets are secured.

• The CRS is directing you to place the 1A Rod Drive MG set in Initiating Cue: service per OP-104 section 5.1

• The initial conditions are completed.

2016 HNP NRC Exam In-Plant JPM k Rev. 2

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 001/BANK/C/A//EOP-ES-0.1/NONE//007EK3.01/

Given the following plant conditions:

- Offsite Power has been lost

- The crew is performing EOP-ES-0.1, Reactor Trip Response In accordance with EOP-ES-0.1 which ONE of the following identifies (1) the temperature indications required to be used per Table 1, RCS Temperature Control Guidelines to control and stabilize temperature AND (2) the reason why?

A. (1) Tavg (2) To ensure adequate RCS heat removal is occurring.

B. (1) Tavg (2) To check for natural circulation established.

C. (1) Tcold (2) To ensure adequate RCS heat removal is occurring.

D. (1) Tcold (2) To check for natural circulation established.

Thursday, May 19, 2016 5:04:38 PM 1

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct:Tcold is the correct indication to use, per EOP-ES-0.1, and because there are no RCPs in service, Tcold is the most accurate indication. Basis is in accordance with WOG Background Document for ES-0.1.

A. Incorrect. The first part is plausible since Tavg is a commonly used indication for many aspects of transients, but in this case, with a loss of offsite power, there is no power to the RCPs, and therefore Tavg is not a reliable indication. The second part is the correct.

B. Incorrect. The first part is plausible since Tavg is a commonly used indication for many aspects of transients, but in this case, with a loss of offsite power, there is no power to the RCPs, and therefore Tavg is not a reliable indication. The second part is plausible since checking for natural circulation is plausible since this is a goal of the procedure, but only towards the end, and is not the specific reason.

C. Correct.

D. Incorrect. The first part is correct, since Tcold is the correct indication to use, since there are no RCPs in service. The second part is plausible since checking for natural circulation is plausible since this is a goal of the procedure, but only towards the end, and is not the specific reason.

Thursday, May 19, 2016 5:04:38 PM 2

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000007 (BW/E02&E10; CE/E02) Reactor Trip - Stabilization - Recovery / 1 007EK3.01; Knowledge of the reasons for the following as the apply to a reactor trip:

Actions contained in EOP for reactor trip (CFR 41.5 /41.10 / 45.6 / 45.13)

Importance Rating: 4.0 4.6 Technical

Reference:

EOP-ES-0.1, step 4 page 6 WOG Background Document for ES-0.1, pp 10 References to be provided: None Learning Objective: EOP-LP-3.1 Objective 3.e Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 3

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 002/BANK/C/A//FSAR 15.6.1/NONE//008AK2.01/

Given the following plant conditions:

- A Reactor Trip and Safety Injection have occurred

- Containment pressure is 2.5 psig and rising

- RCS pressure is 900 psig and lowering

- Tavg is 550°F and lowering slowly

- PRZ level is 85% and rising Which ONE of the following identifies the cause of this event?

A. Small break on an RCS hot leg B. Large break on an RCS cold leg C. A stuck open PRZ PORV D. A stuck open PRZ Spray Valve Thursday, May 19, 2016 5:04:38 PM 4

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: A steam space (or vapor space) LOCA is occurring as evident by PRZ level rising with RCS pressure lowering and Containment pressure rising. An open PRZ PORV or PRZ Safety will depressurize the RCS and an opening on the PRZ will allow the vapor space to exit the PRZ. Both the PORV and Safety valves relieve to the PRT. The PRT will eventually rupture (100 psig with 2 rupture discs) at which time the relief valves will essetially be relieving directly into the Containment which will cause Containment pressure to rise. Safety Injection flow will cause RCS temperature (Tavg) to lower.

A. Incorrect. Plausible since a break in the RCS to Containment would result in the conditions provided in the stem (rising Containment pressure, lowering RCS pressure, Reactor Trip and Safety Injection and eventually rising PRZ level as SI flow exceeds break flow) except RCS pressure is below saturation pressure so a Small Break LOCA cannot be in progress.

B. Incorrect. Plausible since a break in the RCS to Containment would result in the conditions provided in the stem (rising Containment pressure, lowering RCS pressure, Reactor Trip and Safety Injection and eventually rising PRZ level as SI flow exceeds break flow) except RCS pressure would be rapidly lowering and PRZ level would be lowering to zero or at zero and NOT rising.

C. Correct.

D. Incorrect. Plausible since a stuck open PRZ spray valve failure would result in lowering RCS pressure and rising PRZ level; however this is incorrect because this condition will not result in Containment pressure rising unless there was a break in the RCS or a open release path such as a open PORV or PRZ safety valve.

Thursday, May 19, 2016 5:04:38 PM 5

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000008 Pressurizer Vapor Space Accident / 3 008AK2.01; Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Valves (CFR 41.7 / 45.7)

Importance Rating: 2.7 2.7 Technical

Reference:

HNP FSAR Chapter 15, Section 15.6, page 1 References to be provided: None Learning Objective: BD-LP-3-3, Objective 1.f Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 6

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 003/NEW/C/A//STEAM TABLES, ES-1.2/STEAM TABLES//009EK1.02/

Given the following plant conditions:

- The crew is implementing EOP-ES-1.2, Post LOCA Cooldown And Depressurization Subquently the following plant conditions exist:

- Containment pressure is 3.4 psig and lowering

- PRZ level is 35% and lowering slowly

- RCS pressure is 1325 psig and stable

- RCS Loop THOT is 555°F in all 3 loops and lowering

- Highest CET indicates 568°F and lowering slowly

- ERFIS is NOT available Which ONE of the following completes the statements below?

For the current plant conditions, SI Re-Initiation is (1) , AND RCS cooldown will be maintained by the (2) .

A. (1) required (2) SG PORVs B. (1) required (2) condenser steam dumps C. (1) NOT required (2) SG PORVs D. (1) NOT required (2) condenser steam dumps Thursday, May 19, 2016 5:04:38 PM 7

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The conditions of the plant are indicative of a small break LOCA. Based on the parameters monitored RCS saturation temperature at 1340 psia is 581.5°F and in accordance with EOP-ES-1.2, to determine if SI re-initiation is required RCS subcooling is required to be greater than 40°F for adverse containment conditions. With CET at 568°F the subcooling margin is 13.5°F. Because containment pressure is above 3 psig a MSLI signal has occurred and the condenser steam dumps are not available therefore the SG PORVs will be used as the method to cooldown the RCS.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the Steam Dumps are the normal method of heat removal post trip; however this is incorrect since contaiment pressure has generated a MSLI the main condenser is not available.

C. Incorrect. The first part is plausible since the RCS subcooling is greater than the 10°F requirement for non-adverse conditions in containment; however this is incorrect because containment is greater than 3 psig and adverse values apply therefore the RCS subcooling is required to be greater than 40°F. The second part is correct.

D. Incorrect. The first part is plausible see B(2). The second part is plausible see C(1).

Thursday, May 19, 2016 5:04:38 PM 8

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000009 Small Break LOCA / 3 009EK1.02; Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Use of steam tables (CFR 41.8 / 41.10 / 45.3)

Importance Rating: 3.5 4.2 Technical

Reference:

EOP-ES-1.2, Foldout Page, Rev 1, Page 11 EOP-ES-1.2, Step 10.f RNO, Rev 1, Page 12 References to be provided: Steam Tables Learning Objective: EOP-LP-3.5, Objective 4 Question Origin: New Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 9

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 004/NEW/C/A//ALB-004, EOP-E-1 BGD/NONE//011EG2.4.46/

Given the following plant conditions:

- ALB-001-4-1, Containment Spray Actuation has alarmed

- 1A-SA Containment Spray pump has tripped on over current

- 1B-SB Containment Spray pump is operating 35 minutes later the following alarm annunciates:

- ALB-004-2-4, Refueling Water Storage Tank 2/4 Low Low Level, alarms Which ONE of the following completes the statement below?

Based on the conditions above, (1) Containment Sump Recirculation valve(s) automatically open(s) due to a (2) event occurring A. (1) ONLY the 1B CT pump (2) Main Steamline Break B. (1) ONLY the 1B CT pump (2) Large Break LOCA C. (1) BOTH 1A and 1B CT pumps (2) Main Steamline Break D. (1) BOTH 1A and 1B CT pumps (2) Large Break LOCA Thursday, May 19, 2016 5:04:38 PM 10

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: With a CNMT Spray pump running and 2/4 low-low levels present, the following valves reposition for the running pump:

1CT-105 or 1CT-102, CNMT Sump to CT Pump, opens.

1CT-26 or 1CT-71, RWST to CT Pump, shuts.

During a a large break LOCA after successful initial operation of the ECCS, the reactor core is once again covered with borated water. This water has enough boron concentration to maintain the core in a shutdown condition. Decay heat is removed by a continuous supply of water from the ECCS. This supply initially comes from the refueling water storage tank (RWST). When the RWST level reaches the switchover setpoint the ECCS pumps are transferred into the recirculation mode (using ES-1.3, TRANSFER TO COLD LEG RECIRCULATION) wherein water is drawn from the containment sump and is cooled in the residual heat removal heat exchangers.

A. Incorrect. The first part is correct. The second part is plausible since ALB-001-3-1, Containment High 3 Press Alert, is in alarm the pressure inside CNMT has exceeded 10 psig and the design pressure of CNMT is based on a Main Steamline break inside CNMT; however this is incorrect because ALB-004-2-4, Refueling Water Storage Tank 2/4 Low Low Level, indicates the RWST inventory has been discharged into the CNMT sump which can only occur due to a loss of coolant accident vice a loss of secondary inventory, which reduces the volume of the RCS due to contraction of the RCS volume as a result of cooling down. This change in volume is correctable once the contents of the faulted SG are isolated from the intact SG's requiring only AFW to remove decay heat.

B. Correct.

C. Incorrect. The first part is plausible if the candidate misapplies the RHR pump CNMT Sump suction valves logic to the CNMT Spray pump CNMT suction valve logic since the RHR pump CNMT Sump suction valves SI-300, 301, 310 and 311 automatically open due to the presence of an SI signal and low low RWST level; however this is incorrect for the CNMT Spray pump CNMT suctions valve because they require the breaker for its associated pump to be closed in addition to the low low RWST level.

The second part is plausible see A(2).

D. Incorrect. The first part is plausible see C(1). The second part is correct.

Thursday, May 19, 2016 5:04:38 PM 11

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000011 Large Break LOCA / 3 011EG2.4.46; Ability to verify that the alarms are consistent with the plant conditions.

(CFR: 41.10 / 43.5 / 45.3 / 45.12)

Importance Rating: 4.2 4.2 Technical

Reference:

APP-ALB-004, Window 2-4, Rev 18, Page 9 ERG-BKGRD-E-1, Rev 2, Page 30 References to be provided: None Learning Objective: EOP-LP-3.1, Objective 1.a Question Origin: New Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 12

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 005/NEW/C/A//T.S. 3.1.2.3/NONE//022AK3.05/

Given the following plant conditions:

- The unit is in Mode 6 with refueling in progress

- RCS temperature is 114°F

- 'A' CSIP is in service

- 'B' and 'C' CSIPs are under clearance per the outage schedule.

Subsequently the 'A' CSIP trips on overcurrent Concerning Tech Spec 3.1.2.3, Reactivity Control Systems: Charging Pump -

Shutdown, which ONE of the following identifies:

(1) The impact, IF any, on refueling operations AND (2) the reason for the decision?

A. (1) Suspend all operations involving CORE ALTERATIONS or positive reactivity changes (2) There is now ONLY one method to borate the RCS when two are required B. (1) Suspend all operations involving CORE ALTERATIONS or positive reactivity changes (2) Both flow paths for boration from either the BA tank or RWST require a CSIP C. (1) There is no impact on refueling operations, refueling can continue (2) ONLY one flow path to borate the RCS is required and one is still available D. (1) There is no impact on refueling operations, refueling can continue (2) There is sufficient shutdown margin with the boron concentration in the RCS during refueling as long as Reactor Coolant temperature is maintained < 200°F Thursday, May 19, 2016 5:04:38 PM 13

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: ONLY one charging/safety injection pump (CSIP) is in service during Mode 6 the other CSIP's are under clearance. The other 2 (of 3) are under clearance below 325°F to provide assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

When the 'A' CSIP tripped Tech Spec 3.1.2.3 could not be met since it requires a CSIP injection pump with the flow path and NO other CSIP is availabe due to being under clearance. The action statement for the loss of boron injection flow paths becoming inoperable is to suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

A. Incorrect. The first part is correct. The second part is plausible if the candidate has a misconception that LCO 3.1.2.1 requires the flow path from the BAT or the RWST to be operable the RWST gravity drain flowpath remains available and the BAT does not have a gravity drain flowpath; however this is incorrect because a CSIP is required to be operable in both 3.1.2.1 AND 3.1.2.3.

B. Correct.

C. Incorrect. The first part is plausible if the candidate has a misconception that the RWST gravity drain flowpath meets the requirements for both LCO 3.1.2.1 and 3.1.2.3; however this is incorrect because a CSIP is required to be operable in both 3.1.2.1 AND 3.1.2.3. The second part is plausible see A(2).

D. Incorrect. The first part is plausible see C(1). The second part is plausible since during refueling the boron concentration in the RCS would be more than adequate to keep the Reactor shutdown with a Keff of more than 0.95 but Tech Spec 3.1.2.1 deals with RCS boration flow paths and not shutdown margins.

Thursday, May 19, 2016 5:04:38 PM 14

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000022 Loss of Rx Coolant Makeup / 2 022AK3.05; Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: Need to avoid plant transients (CFR 41.5, 41.10 / 45.6 / 45.13)

Importance Rating: 3.2 3.4 Technical

Reference:

Tech Spec 3.1.2.3 References to be provided: None Learning Objective: PMS Objective 12.a Question Origin: New Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 15

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 006/BANK/C/A//EOP-ES-1.3/NONE//025AK2.05/

Given the following plant conditions:

- Large-break LOCA occurred

- RWST level 18%

- CNMT Wide Range Sump level 138.2 inches Which ONE of the following describes the significance of the indicated CNMT wide range sump level as operators take action to transfer to cold leg recirculation?

A. Sump Boron may be inadequate to maintain the Reactor shutdown.

B. RHR pump NPSH may be inadequate to maintain recirculation.

C. Sump pH may be higher than required for post accident limits.

D. Safety related equipment in the Containment may be flooded.

Thursday, May 19, 2016 5:04:38 PM 16

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Because RWST level is less than 23.4% the RHR system has automatically aligned to the Containment Recirc sump. A loss of RHR would occur in the conditions due to a degraded Contianment sump. Degraded sump performance could only occur if water level fell below this level OR the strainer modules experienced excessive clogging. Inadequate sump inventory can be diagnosed by observation of CNMT Wide Range Sump Level (LI-7162A SA/LI-7162B SB) and Recirculation Sump level (LI-7160A SA/LI-7160B SB). A minimum of 142 INCHES CNMT wide range level ensures the recirculation sump strainers are completely submerged and assures a long term recirculation suction source.

A. Incorrect. Plausible because a CNMT Wide Range Sump high level of 196 inches is associated with RB flooding and could be indicative of a leak of service water, CCW or other non-borated water source into containment, which could lower sump boron concentration to less than expected. However this is incorrect because 138.2 inches. of CNMT Wide Range Sump level is low, not high.

B. Correct.

C. Incorrect. Plausible because a low sump level would indicate that an abnormal Sodium Hydroxide level may be present in the sump, which would affect sump pH. However this is incorrect because in accordance with EOP-ES-1.3 Attachment 1 post-accident sump pH should remain within acceptable limits provided that the RWST inventory is not refilled and subsequently added to the CNMT Wide Range Sump level.

D. Incorrect. Plausible because a CNMT Wide Range Sump high level of 196 inches is associated with RB flooding and would indicate a potential for flooding of vital equipment. However this is incorrect because 138.2 inches. of CNMT Wide Range Sump level is low, not high.

Thursday, May 19, 2016 5:04:38 PM 17

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000025 Loss of RHR System / 4 025AK2.05; Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: Reactor building sump (CFR 41.7 / 45.7)

Importance Rating: 2.6 2.6 Technical

Reference:

EOP-ES-1.3, Note prior to Step 1, page 4 and Attachment 1 page 32, Rev 2 References to be provided: None Learning Objective: EOP-LP-2.3/3.3, Objective 5.c Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 18

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 007/BANK/C/A//AOP-019/NONE//027AA1.01/

Given the following plant conditions:

- The plant is operating at 100% power Subsequently PK-444A, PRZ Pressure Master Controller malfunctions

- The crew enters AOP-019, Malfunction of RCS Pressure Control

- PK-444A, PRZ Pressure Master Controller, is placed in MANUAL

- PRZ Pressure is 2050 psig and stable Which ONE of the following describes the action required to return pressure to 2235 psig using PK-444A?

A. Lower the output B. Lower the setpoint C. Raise the output D. Raise the setpoint Plausibility and Answer Analysis Reason answer is correct: From the PRZPC Student Text, with the controller in MANUAL, the output of the controller is controlled by two manual pushbuttons. While in MANUAL, pressing the raise pushbutton will affect the system by simulating a pressure that is above reference pressure, therefore, it is sensed as a demand to decrease plant pressure. Conversely,pressing the Lower pushbutton is sensed as a demand by the operator to increase plant pressure. Lowering controller output will energize heaters and raise pressure.

A. Correct.

B. Incorrect. Once in manual adjusting the setpoint will have no effect. Plausible if applicant believes setpoint is still in the control circuitry while in manual.

C. Incorrect. Raising controller output will de-energize heaters/open spray valves lowering pressure. Plausible if candidate doesn't understand the inverse relationship between controller output and actual pressure, ie., believes raising controller output raises pressure.

D. Incorrect. Once in manual adjusting the setpoint will have no effect. Plausible if applicant believes setpoint is still in the control circuitry while in manual.

Thursday, May 19, 2016 5:04:38 PM 19

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000027 Pressurizer Pressure Control System Malfunction / 3 027AA1.01; Ability to operate and / or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: PZR heaters, sprays, and PORVs (CFR 41.7 / 45.5 / 45.6)

Importance Rating: 4.0 3.9 Technical

Reference:

AOP-019, Attachment 2, Rev. 25, Page 20 AOP-019-BD, Rev 27, Page 4 PZRPC Student Text References to be provided: None Learning Objective: LP-AOP-3.19, Objective 5 Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 20

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 008/BANK/FUNDAMENTAL//EOP-FR-S.1/NONE//029EA1.13/

Given the following plant conditions:

- The Reactor was operating at 100% power when an ATWS occurred

- All Turbine throttle valves are NOT shut Which ONE of the following is the preferred method to trip the Main Turbine in accordance with EOP-FR-S.1, Response to Abnormal Nuclear Power Generation?

A. Trip the Turbine from the MCB.

B. Shut all MSIVs and bypass valves.

C. Trip the Turbine locally at the front standard.

D. Manually runback the Turbine using fast action.

Plausibility and Answer Analysis Reason answer is correct: EOP-FR-S.1 checks the status of the Turbine Throttle Valve to determine if the Turbine has tripped. In the event the Turbine did not trip automatically the first action directed is to trip the Turbine from the MCB.

A. Correct.

B. Incorrect. Plausible since this is an RNO action that is procedurally directed in the event that the turbine is not successfully tripped from the main control board. However this is not the most preferred method to trip the Turbine.

C. Incorrect. Plausible since local action is required to Trip the Reactor the candidate may misapply the Reactor Trip breaker local action requirements to the requirements for the Turbine. Additionally the user's guide states that "if the operator cannot satisfy a condition with actions at the MCB, the RNO column should be consulted for instructions regarding local actions. If no local actions are specified in the RNO, the operator should continue attempts to satisfy the condition using all means at his disposal. However this is not the most preferred method to trip the Turbine.

D. Incorrect. Plausible since this is an RNO action that is procedurally directed in the event that the turbine is not successfully tripped from the main control board. However this is not the most preferred method to trip the Turbine.

Thursday, May 19, 2016 5:04:38 PM 21

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000029 ATWS / 1 029EA1.13; Ability to operate and monitor the following as they apply to a ATWS:

Manual trip of main turbine (CFR 41.7 / 45.5 / 45.6)

Importance Rating: 4.1 3.9 Technical

Reference:

EOP-FR-S.1, Step 2 RNO, Rev 2, Page 4 References to be provided: None Learning Objective: EOP-LP-3.15, Objective 2 Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 22

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 009/BANK/FUNDAMENTAL//ERG-BKGD-E-3/NONE//038EK3.09/

Which ONE of the following identifies the REASON why it is desirable to terminate SI flow in EOP-E-3, Steam Generator Tube Rupture, after a rapid cooldown and depressurization of the RCS has been completed?

(Assume SI Termination Critieria is satisfied)

A. To prevent SG overfill.

B. To prevent RWST depletion.

C. To prevent cycling the PRZ PORVs.

D. To prevent a excessive RCS cooldown.

Thursday, May 19, 2016 5:04:38 PM 23

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: If SI flow is not terminated, leakage into the secondary will eventually fill the steam generator with water and lift the atmospheric relief valves.

This could damage the relief valve and main steamline which would complicate subsequent recovery and aggravate the radiological consequences. Hence, SI must be terminated when the criteria in subsequent steps are satisfied to prevent steam generator overfill.

A. Correct.

B. Incorrect. Plausible since the RWST is the suction source the ECCS pumps are aligned to following an actuation signal; however for this event since the RCS inventory is being lost into the S/G, the inventory is recoverable once the RCS pressure is reduced below ruptured S/G pressure.

C. Incorrect. Plausible since operation of the SI system results in reflood of the RCS and once break flow is reduced to less than SI flow the PRZ will refill and repressurize which could result in PRZ PORVs cycling once the PRZ becomes solid; however this is incorrect for this event since the Secondary PORV and Safety setpoints are lower than the PRZ PORV and Safety setpoints which results in the re-establishing of break flow as the RCS inventory is being lost in to the S/G prior to the PRZ PORVs reaching their actuation setpoint.

D. Incorrect. Plausible since the RWST water temperature is normally maintained less than 125°F, injection of this water source when mixed with the RCS coolant will lower the RCS temperature; however this is incorrect for this event since the inventory is being lost into the S/G the effect of lowering the bulk RCS temperture is mimimal compared to the inventory transfer from the RCS.

Thursday, May 19, 2016 5:04:38 PM 24

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000038 Steam Generator Tube Rupture / 3 038EK3.09; Knowledge of the reasons for the following responses as the apply to the SGTR: Criteria for securing/throttling ECCS (CFR 41.5 / 41.10 / 45.6 / 45.13)

Importance Rating: 4.1 4.5 Technical

Reference:

WOG Background Doc EOP E-3 References to be provided: None Learning Objective: EOP-LP-3.2, Objective 4.i Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 25

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 010/BANK/FUNDAMENTAL//EOP-FR-H.1/NONE//054AK1.02/

Given the following plant conditions:

- EOP-FR-H.1, Response to a Loss of Secondary Heat Sink, is being implemented

- RCS bleed and feed has been initiated Subsequently the following conditions exists:

- All SGs are completely dry and depressurized

- Auxiliary Feedwater capability is restored Which ONE of the following describes the STRATEGY used to re-establish Feedwater AND why?

A. Feed ONLY one (1) SG to ensure RCS cooldown rates are established within Technical Specification limits.

B. Feed ONLY one (1) SG to ensure a failure due to excessive thermal stresses is limited to one SG.

C. Feed ALL SGs to establish subcooled conditions in the RCS as soon as possible.

D. Feed ALL SGs to allow termination of RCS bleed and feed as soon as possible.

Plausibility and Answer Analysis Reason answer is correct: One SG is fed at minimal rate to minimize thermal shock and potential damage to the SG tubesheet when SGs are hot and dry. If a failure in an SG occurs due to excessive thermal stresses, the failure is isolated to one steam generator.

A. Incorrect. Plausible as operator is cautioned to control feedwater rates to prevent excessive cooldown for enhanced plant control, not to comply with tech spec requirements.

B. Correct.

C. Incorrect. Plausible as the operator is allowed to depressurize multiple steam generators to allow condensate flow to be used for recovery of heat sink .

D. Incorrect. Plausible as the operator is allowed to depressurize multiple steam generators to allow condensate flow to be used for recovery of heat sink and may be confused with actions to use max rate cooldown during SGTR events Thursday, May 19, 2016 5:04:38 PM 26

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000054 (CE/E06) Loss of Main Feedwater / 4 054AK1.02; Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G (CFR 41.8 / 41.10 / 45.3)

Importance Rating: 3.6 4.2 Technical

Reference:

EOP-FR-H.1 Pg 62 Rev. 1 WOG background document FR-H.1 pg 51 References to be provided: None Learning Objective: Heat Sink Status Tree, EOP3.11 Obj. 4.c Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 27

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 011/BANK/C/A//ECA-0.0, CSFST/NONE//055EG2.4.21/

Given the following plant conditions:

- The crew has just finished the immediate actions of EOP-ECA-0.0, Loss Of All AC Power

- Narrow Range S/G levels are ALL 20%

- Total FW Flow to the S/G's is 350 KPPH Which ONE of the following completes the statements below?

CSFST's (1) being monitored for INFORMATION ONLY.

A RED path (2) exist on CSF-3, HEAT SINK.

A. (1) are (2) does B. (1) are (2) does NOT C. (1) are NOT (2) does D. (1) are NOT (2) does NOT Thursday, May 19, 2016 5:04:38 PM 28

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: EOP-ECA-0.0 requires the CSFST's to be monitored for information only until directed by the procedure. A red path does not exist on CSF-3.

Even though S/G level is <25%, Total FW flow is >210 KPPH.

A. Incorrect. The first part is correct. The second part is plausible since the 25%

minimum SG level required for Heat Sink is not satisfied the candidate may misapply the CSFST terminus; however this is incorrect since the minimum AFW flow of 210 KPPH is satisfied the CSFST terminus is YELLOW.

B. Correct.

C. Incorrect. The first part is plausible since the normal protocol when transitioning from E-0 is to implement monitoring of the CSFSTs; however this is incorrect because the EOP network actions are based on power being available to one train of ECCS equipment and therefore once power is restored the CSFST's are monitored and the procedure is exited. The second part is plausible since the 25% minimum SG level required for Heat Sink is not satisfied the candidate may misapply the CSFST terminus; however this is incorrect since the minimum AFW flow of 210 KPPH is satisfied the CSFST terminus is YELLOW.

D. Incorrect. The first part is plausible since the normal protocol when transitioning from E-0 is to implement monitoring of the CSFSTs; however this is incorrect because the EOP network actions are based on power being available to one train of ECCS equipment and therefore once power is restored the CSFST's are monitored and the procedure is exited. The second part is correct.

Thursday, May 19, 2016 5:04:38 PM 29

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000055 Station Blackout / 6 055EG2.4.21; Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

(CFR: 41.7 / 43.5 / 45.12)

Importance Rating: 4.0 4.6 Technical

Reference:

EOP-CSFST Heat Sink CSF-3, Rev 11 EOP-ECA-0.0, Step 1 Note, Page 3, Rev 3 References to be provided: None Learning Objective: EOP-LP-3.7, Objective 6 Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 30

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 012/NEW/C/A//EOP-ES-0.1, OMM-004/NONE//056AA1.18/

Given the following plant conditions:

- A Reactor Trip occurred due to a Loss of Offsite Power

- The crew is performing actions of EOP-ES-0.1, Reactor Trip Response Which ONE of the following identifies the status of (1) 1AH-15A SA, Control Room Cooling Unit Normal Supply Fan AND (2) the selected AH-2 A-SA, Containment Fan Cooler?

A. (1) Running (2) LO-SPD B. (1) Running (2) HI-SPD C. (1) NOT Running (2) LO-SPD D. (1) NOT Running (2) HI-SPD Thursday, May 19, 2016 5:04:38 PM 31

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The loss of offsite power will start both EDG's and resequence the loads on the A-SA and B-SB safety busses using program A. With the A sequencer running program A it will start 1AH-15A SA, Control Room Cooling Unit Normal Supply Fan. The start of 1AH-15A SA will automatically open 1CZ-1 SA, Control Room Normal Supply Intake Valve.

A. Incorrect. The first part is correct. The second part is plausible since the loss of power to the MCR OAI damper radiation monitors results in a Control Room Isolation Signal (CRIS). The CRIS signal will shut both Control Room Normal Supply Intake Valves, 1CZ-1 SA and 1CZ-2 SB. This however is incorrect because the CRIS has a time delay to allow the EDG to restore power to the Safety Bus before the signal will generate an actuation.

B. Correct.

C. Incorrect. The first part is plausible since the loss of power to the MCR OAI damper radiation monitors results in a Control Room Isolation Signal (CRIS). The CRIS signal will trip 1E-9A, Control Room Cooling Unit Normal Exhaust Fan and the candidate may misapply this knowledge to the 1AH-15 SA, Control Room Cooling Unit Normal Supply Fan. This however is incorrect because the CRIS generates a start signal for 1AH-15 SA. The second part is plausible since the 1CZ-1 SA, Control Room Normal Supply Intake Valve repositions based on the status of the 1AH-15A SA and will automatically shut with the air handler not running. This however is incorrect because the CRIS generates a start signal for 1AH-15 SA.

D. Incorrect. The first part is plausible since the loss of power to the MCR OAI damper radiation monitors results in a Control Room Isolation Signal (CRIS). The CRIS signal will trip 1E-9A, Control Room Cooling Unit Normal Exhaust Fan and the candidate may misapply this knowledge to the 1AH-15 SA, Control Room Cooling Unit Normal Supply Fan. This however is incorrect because the CRIS generates a start signal for 1AH-15 SA. The second part is plausible since a CRIS will open 1CZ-D66 SA, Emergency Filtration Recirc valve and the candidate may misapply this knowledge to 1CZ-1 SA, Control Room Normal Supply Intake Valve. This however is incorrect because 1CZ-1 SA repositions based on the status of the 1AH-15A SA and will automatically shut with the air handler not running.

Thursday, May 19, 2016 5:04:38 PM 32

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000056 Loss of Off-site Power / 6 056AA1.18; Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power: Control room normal ventilation supply fan (CFR 41.7 / 45.5 / 45.6)

Importance Rating: 3.2 3.2 Technical

Reference:

EOP-ES-0.1, Step 15.e, Rev 2, Page 22 EOP-ES-0.1, Attachment 1, Step 4, Rev 2, Page 31 OMM-004, Attachment 12, Rev 38, Page 62 References to be provided: None Learning Objective: EOP-LP-3.22, Objective 5 Question Origin: New Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 33

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 013/BANK/C/A//AOP-024, APP-013/NONE/EARLY/057AA2.20/SAT Given the following plant conditions:

- The plant is operating at 100% power

- Instrument Bus SI has de-energized

- The crew is implementing AOP-024, Loss Of Uninterruptible Power Supply Which ONE of the following completes the statements below?

Placing the ROD STOP BYPASS switch on the Miscellaneous Control and Indication Panel to the "Bypass PR 41" position will bypass the (1) overpower rod stop signal from N-41.

This action will change the coincidence for the overpower rod stop to (2) remaining channels.

(1) (2)

A. 103% 1 of 3 B. 103% 2 of 3 C. 108% 1 of 3 D. 108% 2 of 3 Plausibility and Answer Analysis Reason answer is correct: Overpower Rod Stop (103%) logic is 1 of 4 NI's. After bypassing N-41, the coincidence will be 1 of 3 remaining NI channels A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the logic for the Power Range High Neutron Flux Reactor trip is normally 2 of 4 NI's.

When one of the instruments is bypassed the logic is reduced to 2 of 3 NI's.

C. Incorrect. The first part is plausible since the Power Range High Neutron Flux Reactor trip is 108% of rated thermal power; however this is incorrect because the Rod Stop Bypass setpoint is 103%. The second part is correct D. Incorrect. The first part is plausible see C(1). The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:38 PM 34

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000057 Loss of Vital AC Inst. Bus / 6 057AA2.20; Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: Interlocks in effect on loss of ac vital electrical instrument bus that must be bypassed to restore normal equipment operation (CFR: 43.5 / 45.13)

Importance Rating: 3.6 3.9 Technical

Reference:

AOP-024, Step 1, Rev 56, Page 6 AOP-024, Attachment 1, Rev 56, Page 28 APP-ALB-013, Window 5-1, Rev 34, Page 21 References to be provided: None Learning Objective: AOP-LP-3.24, Objective 4 Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 35

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 014/BANK/C/A//FSAR 8.3.2/NONE//058AK1.01/

With the DSDG is under clearance the following sequence of events occur:

- At 1455 a Loss of the DP-1A-SA occurs

- At 1500 a Loss of All AC Power occurs

- At 1700 power is restored to buses 1A-SA and 1B-SB

- At 1730 Battery chargers 1A-SA and 1B-SB are restored and are charging their respective batteries Which ONE of the following completes the statement below?

The battery chargers will be .

A. unable to carry steady state normal or emergency loads until its associated battery has been fully charged B. unable to carry steady state normal or emergency loads until its associated battery is charged for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. immediately able to carry steady state normal or emergency loads while its associated battery is being charged D. immediately able to carry emergency loads but unable to carry steady state normal loads until its associated battery has been fully charged Thursday, May 19, 2016 5:04:38 PM 36

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: each individual safety-related charger is capable of maintaining the connected battery in a fully charged condition by supplying a float charge at 133.5V or an equalizing charge at 138.6V, and has the capability to restore sufficient battery capacity to successfully perform the design basis duty cycle in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an emergency discharge while supplying 100 percent of the continuous load on the D.C. bus.

A. Incorrect. Plausible since each battery charger is designed to provide adequate capacity to restore its associated battery to full charge in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the battery has been fully discharged, while carrying steady state normal or emergency loads.

B. Incorrect. Plausible since each battery charger is designed to provide adequate capacity to restore its associated battery to full charge in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the battery has been fully discharged, while carrying steady state normal or emergency loads. A Battery is designed to power all emergency loads up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but it does not have to be charged for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after discharge prior to carrying all loads.

C. Correct.

D. Incorrect. Plausible since each battery charger is designed to provide adequate capacity to restore its associated battery to full charge in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the battery has been fully discharged, while carrying steady state normal or emergency loads Thursday, May 19, 2016 5:04:38 PM 37

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000058 Loss of DC Power / 6 058AK1.01; Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation (CFR 41.8 / 41.10 / 45.3)

Importance Rating: 2.8 3.1 Technical

Reference:

FSAR Chapter 8, Section 3.2.1.2, amendment 57, page 8.3.2-2 References to be provided: None Learning Objective: DCP Student Text, Objective 2 Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 38

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 015/BANK/C/A//EOP-ECA-1.2/NONE//WE04EK2.2/

Given the following plant conditions:

- Reactor Trip and Safety Injection have occurred from 100% power

- PRZ level is off scale low

- The crew is implementing EOP-ECA-1.2, LOCA Outside Containment

- The leak has been isolated by shutting 1SI-340, Low Head SI Train 'A' to Cold Leg Valve Which ONE of the following describes (1) the parameter used to determine that the break was isolated AND (2) which RHR pump(s) must now be secured in accordance with EOP-ECA-1.2?

A. (1) RCS Pressure rising (2) ONLY 'A' RHR Pump B. (1) RCS Pressure rising (2) 'A' and 'B' RHR Pumps C. (1) RCS Subcooling rising (2) ONLY 'A' RHR Pump D. (1) RCS Subcooling rising (2) 'A' and 'B' RHR Pumps Thursday, May 19, 2016 5:04:38 PM 39

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: EOP-ECA-1.2 step directs use of RCS Pressure. Stopping

'A' RHR Pump is correct as its flow path is isolated to support break isolation.

A. Correct.

B. Incorrect. EOP-ECA-1.2 directs use of RCS Pressure. Stopping 'A' and 'B' RHR Pump is incorrect. Only 'A' RHR Pump is secured because its flow path is isolated to support break isolation. Other procedures, such as EOP-E-1 direct stopping 'A' and 'B' RHR Pump if RCS Pressure is greater than 230 psig, but the candidate is not given this information and must evaluate why EOP-ECA-1.2 directs stopping 'A' RHR.

C. Incorrect. EOP-ECA-1.2 directs use of RCS Pressure. RCS Subcooling could be an indication of RCS Pressure rising if RCS Temperature is stable but the candidate is not given a temperature trend and EOP-ECA-1.2 directs use of RCS Pressure. Stopping 'A' RHR Pump is correct as its flow path is isolated to support break isolation.

D. Incorrect. EOP-ECA-1.2 directs use of RCS Pressure. RCS Subcooling could be an indication of RCS Pressure rising if RCS Temperature is stable but the candidate is not given a temperature trend and EOP-ECA-1.2 directs use of RCS Pressure. Stopping 'A' and 'B' RHR Pump is incorrect. Only 'A' RHR Pump is secured because its flow path is isolated to support break isolation. Other procedures such as EOP-E-1 direct stopping 'A' and 'B' RHR Pump if RCS Pressure is greater than 230 psig, but the candidate is not given this information and must evaluate why EOP-ECA-1.2 directs stopping 'A' RHR.

Thursday, May 19, 2016 5:04:38 PM 40

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal W/E04 LOCA Outside Containment / 3 WE04EK2.2; Knowledge of the interrelations between the (LOCA Outside Containment) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7)

Importance Rating: 3.8 4.0 Technical

Reference:

EOP-ECA-1.2, Step 4.d, Rev 0, page 4 References to be provided: None Learning Objective: LP-EOP-3.3, Obj. 2d Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 41

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 016/NEW/C/A//TS 3.7.1 EOP-FR-H.1/NONE//WE05EG2.2.36/

Given the following plant conditions:

- The plant is operating at 100%

- The TDAFW pump is under clearance for electrical work on the Trip and Throttle Valve solenoid Subsequently the following occurs:

- The crew manually trips the Reactor due to a loss of the 'B' CBP

- The 'B' SUT Lockout trips during the fast bus transfer

- 'B' Emergency Diesel Generator fails to start

- ALB-017-5-4, Aux Feedwater Pump A Trip Or Close Ckt Trouble, alarms

- Narrow Range SG level indications are as follows:

- 'A' SG 13% and lowering

- 'B' SG 15% and lowering

- 'C' SG 11% and lowering Which ONE of the following identifies (1) the required number of OPERABLE Auxiliary Feedwater pumps in accordance with the LCO for Technical Specification 3.7.1.2 Plant Systems - Auxiliary Feedwater AND (2) the preferred AVAILABLE source of feedwater for restoration of heat sink in accordance with EOP-FR-H.1, Response To Loss Of Heat Sink?

A. (1) Two (2) 'A' Main Feedwater Pump B. (1) Two (2) 'A' MDAFW Pump C. (1) Three (2) 'A' Main Feedwater Pump D. (1) Three (2) 'A' MDAFW Pump Thursday, May 19, 2016 5:04:38 PM 42

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Technical Specification 3.7.1.2 states at least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps each capable of being powered from separate emergency buses. and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

EOP-FR-H.1 attempts to restore feedwater flow in the following sequence:

Step 5, Establish AFW Flow To At Least One SG Step 10, Establish Main FW Flow To At Least One SG Step 13, Depressurize One SG To Less Than 500 PSIG AND Establish Condensate Flow Based on the conditions in the stem of the question both MDAFW pumps ( 'B' MDAFW pump has no power and 'A' MDAFW pump has a breaker trip) and the TDAFW pump are not available, and the use of the first preferred source to restore heat sink capabilities with AFW flow is not available. The use of the Main Feedwater system is available via the 'A' train equipment. The 'A' Condensate pump, Condensate Booster pump and Main Feedwater pump all have power available and do not require the SG pressure to be reduced in order to restore feedwater flow.

A. Incorrect. The first part is plausible since both the CCW pumps and the CSIPs only require two of the three available pumps to be operable; however this is incorrect because the Technical Specifications require all three of the AFW pumps to be operable for the LCO. The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since it is normal for ALB-017-5-3, Aux Feedwater Pump A Trouble, to alarm when the 'A' MDAFW pump is started via the sequencer and the pump is available; however this is incorrect because ALB-017-5-4, Aux Feedwater Pump A Trip Or Close Ckt Trouble, is an indication of a breaker trip or breaker control power issue that must be investigated locally.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:38 PM 43

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal W/E05 Inadequate Heat Transfer - Loss of Secondary Heat Sink / 4 WE05EG2.2.36; Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

(CFR: 41.10 / 43.2 / 45.13)

Importance Rating: 3.1 4.2 Technical

Reference:

EOP-FR-H.1, Step 10, Rev 1, Page 18 References to be provided: None Learning Objective: EOP-LP-3.11, Objective 4.c Question Origin: New Comments: Ask Michael: JR thinks that we don't have the right K/A for the words for this K/A...which K/A # or words are correct? 3-23-2016 discussed with Michael and he has concurred that K/A should be WE05EG2.2.36 Phonecon 3/23: HNP concurs.

Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 44

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 017/BANK/FUNDAMENTAL//AOP-014, BD/NONE//026AA2.06/

Given the following plant conditions:

- The unit is operating at 100% power

- A loss of CCW has occurred

- The crew is attempting to restore CCW flow in accordance with AOP-014, Loss of Component Cooling Water

- BOTH trains of CCW flow indicate 0 gpm

- All RCP temperatures are currently below their alarm setpoints and slowly rising

- RCP Seal Injection flow to each RCP is approximately 9 gpm Which ONE of the following identifies (1) the MAXIMUM time allowed to trip the RCPs in accordance with AOP-014 AND (2) the components that may be damaged if the RCPs are not tripped?

A. (1) 5 minutes (2) RCP motor bearings B. (1) 5 minutes (2) RCP pump bearings C. (1) 10 minutes (2) RCP motor bearings D. (1) 10 minutes (2) RCP pump bearings Thursday, May 19, 2016 5:04:38 PM 45

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Operation of RCPs for greater than 10 minutes without CCW cooling to the motor oil coolers may result in RCP bearing damage.

A. Incorrect. This is plausible since AOP-018 has a 3 to 5 minute time window to isolate seal water to a failed # 1 seal. 9 gpm seal leak off flow is indication of a failed # 1 seal; however this is incorrect because the candidate is provided seal injection flow of 9 gpm and this 3 to 5 minute time window is not applicable to the conditions provided. The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since the themal barrier heat exchanger is cooled by CCW flow and it minimizes the temperature gradiant of the RCS to minimize the heat up of the pump radial bearing; however this is incorrect because the requirement to stop the RCP is based on no CCW flow to the Motor cooler not the thermal barrier heat exchanger.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible see B(2).

000026 Loss of Component Cooling Water / 8 026 AA2.06; Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The length of time after the loss of CCW flow to a component before that component may be damaged (CFR: 43.5 / 45.13)

Importance Rating: 2.8 3.1 Technical

Reference:

AOP-014, Section 3.2 step 12 Caution, Rev 37, Page 19 AOP-014-BD, Section 3.2 step 12 Caution, Rev 19, Page 20 References to be provided: None Learning Objective: AOP-LP-3.14, Objective 3 Question Origin: Bank Comments: Ask Michael about the K/A WE011EA2.1. This K/A appears to be at the SRO level according to 1021 Attachment 2 for SRO guidance. Requested K/A be replaced with a Thursday, May 19, 2016 5:04:38 PM 46

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal K/A that wasn't as close to SRO job task.

Phonecon 5/03: HNP states that they are concerned about the job link to RO level for this K/A, so I committed to providing a new K/A. I needed to stay in A.2 so as not to unbalance the counts in RO T1G1, and the only other A.2 in WE011 was much like EA2.1 I thought, more of an SRO Q, so I couldnt stay in WE011. Five of the topics in T1G1 were not sampled already, so I randomly chose from those and got Loss of CCW. It has 5 A.2s with RO IR of 2.5 or more, so I randomly chose this K/A.

New K/A APE 026AA2.06: Loss of CCW - The length of time after the loss of CCW flow to a component before that component may be damaged.

Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 47

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 018/BANK/FUNDAMENTAL//EOP-ECA-2.1/NONE//WE12EA2.2/

In accordance with EOP-ECA-2.1, Uncontrolled Depressurization of All Steam Generators, which ONE of the following is the required AFW flow rate if RCS cooldown rate is 120°F per hour?

A. 210 KPPH to EACH SG B. 210 KPPH total to ALL SGs C. 12.5 KPPH to EACH SG D. 12.5 KPPH total to ALL SGs Plausibility and Answer Analysis Reason answer is correct: EOP-ECA-2.1 requires feed water flow to be lowered to 12.5 KPPH to each SG when the RCS cooldown rate is greater than 100°F/HR.

A. Incorrect. Plausible since this value is the minimum total feed flow required to maintain the CSF-3, Heat Sink, yellow and prevent transition into EOP-FR-H.1 and feed flow is required to each SG during EOP-ECA-2.1 in order to prevent dryout conditions from occuring on a S/G; however this is incorrect since transition into EOP-FR-H.1 is precluded at this time due the CSFST not being satisfied because of operator actions.

B. Incorrect. Plausible since this value is the minimum total feed flow required to maintain the CSF-3, Heat Sink, yellow and prevent transition into EOP-FR-H.1; however this is incorrect since transition into EOP-FR-H.1 is precluded at this time due the CSFST not being satisfied because of operator actions.

C. Correct.

D. Incorrect. Plausible since the value is correct and total feed flow is the parameter required to be maintained in order to satisfy the CSF-3, Heat Sink; however this is incorrect since a minimum feed flow of 12.5 KPPH is required to each SG during EOP-ECA-2.1 in order to prevent dryout conditions from occuring on a S/G Thursday, May 19, 2016 5:04:38 PM 48

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal W/E12 Steam Line Rupture - Excessive Heat Transfer / 4 WE12EA2.2; Ability to determine and interpret the following as they apply to the (Uncontrolled Depressurization of all Steam Generators): Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

(CFR: 43.5 / 45.13)

Importance Rating: 3.4 3.9 Technical

Reference:

EOP-ECA-2.1, Step 3.a RNO, Rev 1 Page 8 References to be provided: None Learning Objective: EOP-LP-3.9, Objective 5 Question Origin: Bank Comments: None Tier/Group: T1/G1 Thursday, May 19, 2016 5:04:38 PM 49

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 019/NEW/FUNDAMENTAL//AOP-001, ATT 1/NONE//005AA1.01/

In accordance with AOP-001, Malfunction of Rod Control and Indication System, which ONE of the following is indicative of an inoperable/stuck control rod that is misaligned?

A. A QPTR calculation indicates a QPTR of 1.075 B. PR instruments differ by 1.75% between the highest and lowest indicator C. Delta Flux (AFD) indicators differ by 1.75% between the highest and lowest indicator D. Symmetric core outlet thermocouples (TCs) indicate a 9°F difference from the affected core outlet TC.

Plausibility and Answer Analysis Reason answer is correct: In accordance with AOP-001, Malfunction Of Rod Control And Indication System, Attachment 1 a QPTR value greater than 1.02 is indicative of Control Rod misalignment.

A. Correct.

B. Incorrect. Plausible since the Rod Control system band for automatic rod motion is 1.5 - 3.0°F the candidate may misinterpret NI's being greater than 1.5%

as indication that a control rod is misaligned; however this is incorrect since the determination for NI's is a greater than 2% difference C. Incorrect. Plausible since the Rod Control system band for automatic rod motion is 1.5 - 3.0°F the candidate may misinterpret Delta-I being greater than 1.5% as indication that a control rod is misaligned; however this is incorrect since the determination for Delta-I is a greater than 2%

difference D. Incorrect. Plausible since the normal RCS Temperature control band is <2°F and the Transient RCS Temperature control band is <5°F; however this is incorrect since the determination for Core Outlet Temperature is greater than 10°F.

Thursday, May 19, 2016 5:04:38 PM 50

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000005 Inoperable/Stuck Control Rod / 1 005AA1.01; Ability to operate and / or monitor the following as they apply to the Inoperable / Stuck Control Rod: CRDS (CFR 41.7 / 45.5 / 45.6)

Importance Rating: 3.6 3.4 Technical

Reference:

AOP-001, Attachment 1, page 45 References to be provided: None Learning Objective: AOP-LP-3.1, Objective 3 Question Origin: New Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:38 PM 51

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 020/NEW/C/A//OP-105, TS 2.2.1 BAS/NONE//032AG2.2.25/

Given the following plant conditions:

- A start up is in progress in accordance with GP-004, Reactor Startup

- Rx Power is being raised from 8.6 x 103 cps to achieve the required Source to Intermediate Range overlap

- Intermediate Range NI's currently indicate 3.9 x 10-11 amps Subsequently NI-31, Source Range NI, control power fuses blow Which ONE of the following completes the statements below?

Based on the conditions above, a trip of the Reactor (1) occur.

The Technical Specification 3.3.1, RPS Instrumentation bases of the Source Range Reactor Trip Function is to provide protection during Reactor startup to mitigate the consequences of (2) .

A. (1) will (2) a single or multiple control rod drop accident B. (1) will (2) an uncontrolled rod cluster control assembly bank withdrawal C. (1) will NOT (2) a single or multiple control rod drop accident D. (1) will NOT (2) an uncontrolled rod cluster control assembly bank withdrawal Thursday, May 19, 2016 5:04:38 PM 52

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Removal of control power fuses will disable the level trip bypass function of the affected Source Range channel. If power is reduced below P-6 with the control power fuses removed, a Reactor trip signal will be generated by the affected Source Range channel.

The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.

A. Incorrect. The first part is correct. The second part is plausible since the Reactor is crtitical all Shutdown banks are fully withdrawn and 2 of the 4 Control banks are fully withdrawn. However this is incorrect because this is the bases for the power range neutron flux rate trip.

B. Correct.

C. Inorrect. The first part is plausible since the removal of the control power fuses with the associated SRNI blocked and Reactor power is above P-6 will NOT generate a Reactor trip signal. However this is not correct because the Reactor power level is below P-6. The second part is plausible see A(2).

D. Incorrect. The first part is plausible see C(1). The second part is correct.

Thursday, May 19, 2016 5:04:38 PM 53

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000032 Loss of Source Range NI / 7 032AG2.2.25; Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2)

Importance Rating: 3.2 4.2 Technical

Reference:

OP-105, Rev 28, Page 17 T.S. 2.2.1 Bases, Page B 2-2 References to be provided: None Learning Objective: Student Text NIS, Objective 9.a Question Origin: New Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:38 PM 54

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 021/BANK/C/A//OP-105/NONE//033AK1.01/

Given the following plant conditions:

- The plant was operating at 100% power when the Reactor tripped

- IR NI-35 did not respond as expected due to a total loss of compensating voltage

- IR NI-36 responded normally Which ONE of the following describes Source Range instrument operational response to these conditions?

A. ONLY SR channel NI-31 will automatically energize B. ONLY SR channel NI-32 will automatically energize C. NEITHER SR NI will automatically energize D. BOTH SR NIs will automatically energize Plausibility and Answer Analysis Reason answer is correct: Both IR channels must be below the reset for P-6 for the SR NIs to automatically energize.

A. Incorrect. Plausible since one IR channel is below P-6 and SR NIs reset automatically when IR below P-6, but must have 2/2 IR channels <P-6 and SR resets are not train-related B. Incorrect. Plausible since one IR channel is below P-6 and SR NIs reset automatically when IR below P-6, but must have 2/2 IR channels <P-6 and SR resets are not train-related C. Correct.

D. Incorrect. Plausible since one IR channel is below P-6 and SR NIs reset automatically when IR below P-6, but must have 2/2 IR channels <P-6 Thursday, May 19, 2016 5:04:39 PM 55

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000033 Loss of Intermediate Range NI / 7 033AK1.01; Knowledge of the operational implications of the following concepts as they apply to Loss of Intermediate Range Nuclear Instrumentation: Effects of voltage changes on performance (CFR 41.8 / 41.10 / 45.3)

Importance Rating: 2.7 3.0 Technical

Reference:

OP-105 section 7.1.2, Rev 28, Page 11 References to be provided: None Learning Objective: Student Text NIs, Objective 8 Question Origin: Bank Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 56

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 022/BANK/FUNDAMENTAL//AOP-005/NONE/EARLY/061AK3.02/SAT Given the following plant conditions:

- The Control Room South Intake outside air intake (OAI) high airborne radiation monitors RC-1CZ-3505A1-SA and RC-1CZ-3505B1-SB are in ALARM

- The crew is implementing AOP-005, Radiation Monitoring System

- The Shift Manager has determined the crew will remain in the MCR Which ONE of the following identifies the reason(s) why the alarm setting must be re-adjusted in accordance with AOP-005, if an emergency OAI must be opened with the associated monitor in alarm?

A. Allows opening the associated dampers.

B. Clears the alarm in order to minimize MCR distractions.

C. Ensures the dampers will remain open on rising radiation levels.

D. Ensures alarm and auto-closure occur again on rising radiation levels.

Thursday, May 19, 2016 5:04:39 PM 57

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Alarm response for a radiation monitor in alarm will direct the operator to enter AOP-005. In accordance with note prior to AOP-005, Attachment 4 step 4, reason the alarm setting is adjusted is so that it will alarm on subsequent higher radiation levels and to regain auto-closure capability on high radiation alarm at the emergency OAIs.

A. Incorrect. Plausible since the high radiation alarm generates a Control Room Isolation Signal and the associated (AH-15) dampers reposition; However this is incorrect because adjustment of the alarm setpoint is not required since the associated supply fan dampers are opened via a 42X contact based on the status of the associated fan.

B. Incorrect. Plausible since AD-OP-ALL-1000 defines a nuisance alarm as an alarm although valid, repeated actuation of the alarm distracts the plant operators. AD-OP-ALL-1000 provides guidance to consider disabling alarms that will remain in for an extended period of time. It is possible for the CRS/SM to determine the high radiation alarm is a distraction and direct the setpoint be adjusted in response to a nuisance alarm since the crew will remain in the MCR; However this is incorrect with respect to the requirements of AOP-005 and is not the reason the alarm setpoint is adjusted.

C. Incorrect. Plausible since the Control Room Isolation Signal is a single shot actuation. This can be reset with an alarm condtion present which will result in the defeat of the auto-closure capability during subsequent high radiation conditions; However this is incorrect since the procedure sequence requires the adjustment of the high radiation setpoint above the current value prior to opening the emergency OAI's which will reset the single shot actuation and restore the auto-closure capability.

D. Correct.

Thursday, May 19, 2016 5:04:39 PM 58

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000061 ARM System Alarm / 7 061AK3.02; Knowledge of the reasons for the following responses as they apply to the Area Radiation Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM system (CFR 41.5,41.10 / 45.6 / 45.13)

Importance Rating: 3.4 3.6 Technical

Reference:

AOP-005, Attachment 4 Step 4 Note, Rev 30, page 18 AD-OP-ALL-1000, Step 5.5.2.c, Rev 5, Page 20 References to be provided: None Learning Objective: AOP-LP-3.5, Objedtive 3.c Question Origin: Bank Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 59

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 023/NEW/C/A//EOP-FR-C.1/NONE//074EK3.06/

Given the following plant conditions:

- The crew is implementing EOP-FR-C.1, Response To Inadequate Core Cooling

- IA and N2 have been restored to Containment

- The crew is checking for RCS vent paths

- RCS pressure rises to 2345 psig Which ONE of the following identifies (1) the expected PRZ PORV response to the above conditions AND (2) the reason why this response is desired?

A. (1) PRZ PORVs are OPEN.

(2) Preclude the use of the PRZ Safety valves.

B. (1) PRZ PORVs are OPEN.

(2) Preclude the use of the Reactor Vessel vent valves.

C. (1) PRZ PORVs are SHUT.

(2) Prevent primary plant depressurization.

D. (1) PRZ PORVs are SHUT.

(2) To maintain RCP seal P for continued RCP operation.

Thursday, May 19, 2016 5:04:39 PM 60

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: RCS over pressure protection is provided by PRZ PORV's (lift pressure setting of 2335 psig) and PRZ Safety Valves (lift pressure setting of 2485 psig). While checking for a RCS vent path in EOP-FR-C.1 the operator checks that the PRZ PORVs are SHUT. The expected condition of the PRZ PORV's when pressure is

> 2335 psig is OPEN. The procedure instructions when pressure exceeds 2335 psig is to ensure at least ONE PRZ Block valve is OPEN. With a lifting PORV and an OPEN Block Valve RCS pressure can be relieved to the PRT preventing an overpressure event prior to the lifting of the PRZ Safety Valves.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the Reactor Vessel vent valves may be OPEN to relieve RCS pressure; however this is incorrect because they are not designed to prevent an RCS overpressure event. The PRZ PORV's and Safety valves are designed for the prevention of RCS overpressurization.

C. Incorrect. The first part is plausible since the pressure is below the High PRZ Presseure Reactor trip setpoint of 2385 psig the candidate may believe PRZ PORVs have not reach the opening setpoint and are shut; however this is incorrect because the PRZ PORVs lift setpoint is 2335 psig. The second part is plausible since when coupled with the first part the answer "to prevent primary plant depressurization." would be a correct result for the action.

D. Incorrect. The first part is plausible see C(1). The second part is plausible since operation of the RCPs is desired during the implementation of EOP-FR-C.1 with the PRZ PORVs SHUT the RCP seal d/p will be maintained. However this is incorrect because the PRZ PORVs are checked to ensure they prevent an RCS overpressurization event.

Thursday, May 19, 2016 5:04:39 PM 61

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000074 (W/E06&E07) Inadequate Core Cooling / 4 074EK3.06; Knowledge of the reasons for the following responses as the apply to the Inadequate Core Cooling: Confirming that the PORV cycles open at the specified setpoint (CFR 41.5 / 41.10 / 45.6 / 45.13)

Importance Rating: 3.9 4.2 Technical

Reference:

EOP-FR-C.1 Step 10, Page 16, Rev. 1 SDD-FR-C.1 Step 10, Page 8, Rev. 0 References to be provided: None Learning Objective: EOP-LP-3.10 Objective 2 Question Origin: New Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 62

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 024/NEW/FUNDAMENTAL//EOP-ES-0.0/NONE//WE02EG2.4.11/

Which ONE of the following identifies a Major Action category for EOP-ES-0.0, Rediagnosis?

A. Check if there is a SGTR.

B. Check if a Heat Sink is required.

C. Check if a Small Break LOCA is in progress.

D. Check if a LOCA has occurred outside Containment.

Plausibility and Answer Analysis Reason answer is correct: In accordance with EOP E-0.0, Rediagnosis the Major Action categories are:

Check If There is A SGTR Check If Any SGs Are Not Faulted Check If Any SG Is Faulted and If It Was Isolated A. Correct.

B. Incorrect. Plausible since the purpose of EOP-ES-0.0 is to determine or confirm the most appropriate post accident recovery procedure; however this is incorrect because heat sink is a CSF which is determined by monitoring the CSFST.

C. Incorrect. Plausible since the purpose of EOP-ES-0.0 is to determine or confirm the most appropriate post accident recovery procedure; however this is incorrect because the Major action categories evaluate the status of the SG's being faulted or ruptured and if neither condition exsit the then evaluates the appropriate LOCA series procedure.

D. Incorrect. Plausible since the purpose of EOP-ES-0.0 is to determine or confirm the most appropriate post accident recovery procedure; however this is incorrect because the Major action categories evaluate the status of the SG's being faulted or ruptured and if neither condition exist then the evaluates the appropriate LOCA series procedure Thursday, May 19, 2016 5:04:39 PM 63

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal WE02 Rediagnosis & SI Termination / 3 WE02EG2.4.11; Knowledge of abnormal condition procedures.

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: 4.0 4.2 Technical

Reference:

EOP ES-0.0 major action categories, Rev. 0, Page 2 References to be provided: None Learning Objective: EOP-LP-3.19, Objective 1.c Question Origin: New Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 64

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 025/NEW/C/A//EOP-ES-1.2/NONE//WE03EA2.2/

Given the following plant conditions:

- A LOCA has occurred

- RCS pressure is 1350 psig and stable

- Containment pressure is 2.5 psig and slowly rising In accordance with EOP-ES-1.2, Post LOCA Cooldown and Depressurization, which ONE of the following completes the statement below?

Perform the cooldown of the RCS using (1) at (2) .

A. (1) S/G PORVs (2) less than 100°F per hour B. (1) S/G PORVs (2) the maximum achievable rate C. (1) Condenser Steam Dumps (2) less than 100°F per hour D. (1) Condenser Steam Dumps (2) the maximum achievable rate Thursday, May 19, 2016 5:04:39 PM 65

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Steam Dumps will be used for the cooldown because the Condenser is available until CNMT pressure rises above 3 psig and EOP-ES-1.2 limits cooldown to 100°F/hour.

A. Incorrect. Plausible since the SG PORVs will be used for the cooldown because the condenser is not available. Rate is correct.

B. Incorrect. Plausible since the SG PORVs will be used for the cooldown because the condenser is not available. The Cooldown Rate is incorrect. Other EOPs perform a max rate cooldown but EOP-ES-1.2 limits cooldown to 100°F/hour.

C. Correct.

D. Incorrect. Condenser steam dumps are not available because at 3 psig in Containment a MSLI actuated to shut all MSIVs. Credible because it is the normal method of cooldown. The Cooldown Rate is incorrect. Other EOPs perform a max rate cooldown but EOP-ES-1.2 limits cooldown to 100°F/hour.

Thursday, May 19, 2016 5:04:39 PM 66

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal WE03 LOCA Cooldown - Depressurization / 4 WE03EA2.2; Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization): Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

(CFR: 43.5 / 45.13)

Importance Rating: 3.5 4.1 Technical

Reference:

EOP-ES-1.2, step 10, Rev 1, Page 12 References to be provided: None Learning Objective: LP-EOP-3.5, Obj. 5c Question Origin: New Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 67

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 026/BANK/FUNDAMENTAL//SDD-FR-Z.1/NONE//WE14EA1.2/

During implementation of EOP-FR-Z.1, Response to Containment High Pressure, which ONE of the following is the major concern if ESW Booster pumps are not running and Containment pressure is 35 psig?

A. Potential damage to the ESW pumps.

B. Radioactivity release to the environment.

C. Reduced Containment cooling capability.

D. Reduced Margin to Containment design limits.

Plausibility and Answer Analysis Reason answer is correct: In accordance with the EOP-FR-Z.1 step deviation document plant specific Steps 7 through 10 were added to ensure there is no unmonitored release of radioactivity from CNMT through the ESW system. The steps also ensure design heat removal capability is maintained. The ESW system configuration following SI actuation has no remote radiation monitoring capability. A leakage path could exist from CNMT atmosphere if fan cooler tube leaks were present with indicated CNMT pressure greater than 11.6 PSIG, and either the ESW booster pumps or the associated orifice bypass isolation valve failed (ref ESR 951022). These EOP steps isolate the potential ESW release path.

A. Incorrect. Plausible since the higher the CNMT pressure the closer the ESW pump is to its design operating margin of 225 ft of head and therefore the potenial to damage the ESW pump is higher; however this is incorrect because the ESW booster pump and orifice bypass isolations are designed to prevent the ESW Pump from exceeding its design margin.

B. Correct.

C. Incorrect. Plausible since the ESW Booster pump has a higher discharge pressure than the ESW pump therefore the potential for higher flow through the CNMT Fan Coolers is possible; however this is incorrect because the ESW pumps are able to provide cooling for the CNMT Fan Coolers and prevent exceeding the FSAR design limits.

D. Incorrect. Plausible since the higher the CNMT pressure the closer the plant is to the design margin of 45 psig; however this is incorrect because the ESW pumps are able to provide cooling for the CNMT Fan Coolers and prevent exceeding the FSAR design limits.

Thursday, May 19, 2016 5:04:39 PM 68

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000069 (W/E14) Loss of CTMT Integrity / 5 WE14EA1.2; Ability to operate and / or monitor the following as they apply to the (High Containment Pressure): Operating behavior characteristics of the facility.

(CFR: 41.7 / 45.5 / 45.6)

Importance Rating: 3.3 3.4 Technical

Reference:

SDD-FR-Z.1, Rev 0, Page 3 References to be provided: None Learning Objective: EOP-LP-3.13, Objective 4 Question Origin: Bank Comments: None Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 69

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 027/NEW/C/A//EOP-FR-Z.2/NONE//WE15EK2.2/

Given the following plant conditions:

- 'A' SG is ruptured and faulted inside Containment

- SI termination criteria are NOT met

- The crew is currently implementing EOP-ECA-3.1, SGTR with Loss of Reactor Coolant: Subcooled Recovery Desired Subsequently the following conditions exists:

- RWST level is 24.9% and slowly lowering

- Containment Flooding is a valid ORANGE path

- There are no other RED or ORANGE paths Based on the current plant conditions, which ONE of the following completes the statements below?

The crew will transition to (1)

The reason Chemistry samples of the Containment sump are collected for the selected procedure is to determine the (2) .

Procedure Titles:

EOP-ES-1.3, Transfer to Cold Leg Recirculation EOP-FR-Z.2, Response to Containment Flooding A. (1) EOP-ES-1.3 (2) activty level in the water B. (1) EOP-ES-1.3 (2) pH level of the water C. (1) EOP-FR-Z.2 (2) activty level in the water D. (1) EOP-FR-Z.2 (2) pH level of the water Thursday, May 19, 2016 5:04:39 PM 70

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Because the crew is implementing EOP-ECA-3.1 and the EOP-Users Guide rules of usage dictate any time EOP-E-0 has been exited to transition to another EOP, FRs are to be implemented based on the status of the CSFST. With a valid Orange condition on the CSFST terminus for Containment Flooding the crew will transition to EOP-FR-Z.2 and Step 2 of EOP-FR-Z.2 will have the crew check Containment Sump activity level.

A. Incorrect. The first part is plausible since RWST level is approaching the foldout critieria of 23.4% for Cold Leg Recirculation Switchover at 24.9%;

however this is incorrect because the criteria is not satisfied until RWST level is below 23.4%. The second part is plausible since EOP-ES-1.3 will sample the Primary and Secondary for activity; however this is incorrect because the Containment sump water is only sampled to determine boron concentration and pH.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since it is the correct reason fo the selected procedure.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since it correct reason during other EOP procedures, see A(2); however this is the incorrect reason for the selected procedure.

Thursday, May 19, 2016 5:04:39 PM 71

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal W/E15 Containment Flooding / 5 WE15EK2.2; Knowledge of the interrelations between the (Containment Flooding) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

(CFR: 41.7 / 45.7)

Importance Rating: 2.7 2.9 Technical

Reference:

EOP-Users Guide, Section 5.2.3, Rev 46, Page 23 EOP-FR-Z.2, Step 2, Rev 0, Page 4 References to be provided: None Learning Objective: EOP-LP-3.13, Objective 4.e Question Origin: New Comments: This K/A is met by identifying the reason the containment recirc sump portion of the emergency coolant system is sampled during implementation of EOP-FR-Z.2 confirming the system has or has not properly operated.

Tier/Group: T1/G2 Thursday, May 19, 2016 5:04:39 PM 72

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 028/BANK/C/A//AOP-018/AOP-018 ATTACHMENT 2//003K6.02/

Given the following plant conditions:

- The unit is operating at 100% power

- Total #1 seal flow for the 'B' RCP is 7.4 gpm

'B' RCP Seal water inlet and radial bearing temperatures are rising as follows:

Time Seal water inlet temps RCP radial bearing temps 0100 160 °F 165 °F 0115 165 °F 170 °F 0130 169 °F 174 °F 0145 173 °F 178 °F 0200 176 °F 181 °F 0215 178 °F 183 °F Which ONE of the following describes the condition of the 'B' RCP #1 seal?

(Reference provided)

A. Failed B. Degraded C. Blocked D. Repsonding to a #2 seal failure Thursday, May 19, 2016 5:04:39 PM 73

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: AOP-018 Attachment 2 Specific Symptoms of Seal Malfunctions #1 Seal degraded. Total #1 seal flow > 6.0 gpm but < 8 gpm (indicated is 7.4 gpm) with both RCP seal water inlet and radial bearing temperature STABLE (as defined in Note 2, "Stable" temperature). A slow rise in temperature or a rise in temperature but at a lowering rate and well below 230°F. Under these conditions, additional time is available to evaluate the trend and contract Engineering. In the absence of additional guidance, if temperature has risen to > 190°F and is still rising, it should be considered STEADILY RISING. Since the temperature is rising at a lowering rate and no temperature has exceeded 190°F these conditions meet STABLE.

A. Incorrect. Plausible since this would be correct answer if the RCP seal water inlet or radial bearing temperature was evaluated as STEADILY RISING.

B. Correct.

C. Incorrect. Plausible since this would be correct answer if the RCP seal flow was lower since both RCP seal water inlet and redial bearing temperatures are STABLE.

D. Incorrect. Plausible since in accordance with AOP-018-Basis Document, discussion item #15 a failure of the RCP #2 seal will result in an rise in the RCP #2 seal leakage flow with a proportionate reduction in the RCP #1 seal leakoff flow. A #2 RCP seal failure does have an effect on the #1 seal flows but will NOT cause a temperature rise of either the seal water inlet or RCP radial bearing temps. Since a #2 RCP seal failure can affect the

1. 1 seal a candidate could have a misconception that a #2 seal failure would not only cause a #1 seal flow rate change but also cause both Seal water inlet temps and RCP radial bearing temps to rise.

Thursday, May 19, 2016 5:04:39 PM 74

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 003 Reactor Coolant Pump / 4 003K6.02; Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: RCP seals and seal water supply (CFR: 41.7 / 45.5)

Importance Rating: 2.7 3.1 Technical

Reference:

AOP-018 Attachment 2, pages 24 and 25, Rev. 49 AOP-018-BD discussion item 15, page 3, Rev. 25 References to be provided: AOP-018 Attachment 2, pages 24 and 25, Rev. 49 Learning Objective: LP-AOP-3.18 Objective 3.b Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 75

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 029/BANK/C/A//OP-107.02/NONE//004K3.06/

Given the following plant conditions:

- The unit is operating at 100% power

- RCS boron concentration is 1192 ppm Subsequently the following occurs:

- A new CVCS Cation Bed Demineralizer is to be placed in service and is to be flushed to the Recycle Holdup Tank for sampling

- 1CS-120, Letdown To VCT/ Holdup Tank LCV-115A is placed in the RHT position AND the valve FAILS to reposition

- The Cation Bed Demineralizer flush is initiated Which ONE of the following will occur?

A. RCS Tavg will rise B. RCS pressure lowers C. Letdown flow rises above 60 gpm D. RCS lithium and cesium concentrations rise Thursday, May 19, 2016 5:04:39 PM 76

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: 1CS-120 failing to divert will cause a dilution of the RCS with an unsaturated CVCS cation demin placed in service. RCS Tavg will rise due to an increase in Reactor power. In accordance with OP-107.02, CVCS Demineralizer Operations, a caution when placing a Cation Bed in service states that "until a new bed is borated, the effluent boron concentration is less than RCS boron concentrations.

Therefore, if a new cation bed was inadvertantly placed in service RCS boron concentrations would be reduced, Reactor power would rise and Tavg would rise.

A. Correct.

B. Incorrect. Plausible since RCS pressure would tend to rise on a power increase versus lower. RCS pressure lowering is plausible if the candidate were to think the cation bed is borating the RCS while in service or inverts the effects.

C. Incorrect. Plausible since Letdown flow is limited to 60 gpm in accordance with OP-107.2. Limiting Letdown flow to 60 gpm prior to flushing the demin will minimize any RCS inventory losses as voids in the Cation Bed that cannot be removed during the fill and vent are collapsed and pressure surges if the RC filter clogs. Additionally, 1CS-94, Mixed Bed Demins to VCT Isol Vlv is locally throttled with a flow rate NOT to exceed 60 gpm.

This choice is plausible if the candidate thinks letdown flow would increase due to placing the cation bed in service by having an additional flow path for Letdown to travel through.

D. Incorrect. Plausible if the candidate has a misconception (thinks opposite effect) on how the cation demineralizers effect lithium and cesium concentrations in the RCS. Lithium and cesium concentrations would actually lower with the CVCS cation bed placed in service.

Thursday, May 19, 2016 5:04:39 PM 77

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 004 Chemical and Volume Control / 2 004K3.06; Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: RCS temperature and pressure (CFR: 41.7/ 45.6)

Importance Rating: 3.4 3.6 Technical

Reference:

OP-107.02, CVCS Demineralizer Operations, Caution prior to step 11, Page 39 References to be provided: None Learning Objective: CVCS Objective 6.a Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 78

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 030/NEW/C/A//AOP-002/NONE//004K5.31/

Given the following plant conditions:

- 'A' Boric Acid Pump is under clearance

- An Emergency Boration per AOP-002, Emergency Boration must be performed

- 'B' Boric Acid Pump fails to start

- VCT level is 21% and lowering In accordance with AOP-002, which ONE of the following identifies (1) the valve alignment that would be attempted AND (2) the purpose of this flowpath?

Valve Noun Name:

1CS-283, Boric Acid to Boric Acid Blender FCV-113A 1CS-155, Make Up to VCT FCV-114A 1CS-165, VCT Oulet LCV-115C 1CS-166, VCT Oulet LCV-115E 1CS-291, Suction from RWST LCV-115B 1CS-292, Suction from RWST LCV-115D A. (1) OPEN 1CS-283 and 1CS-155 (2) To prevent gas binding of the CSIPs.

B. (1) OPEN 1CS-283 and 1CS-155 (2) To provide an alternate source of borated water to the CSIPs.

C. (1) OPEN 1CS-291 and 1CS-292 THEN SHUT 1CS-165 and 1CS-166 (2) To prevent gas binding of the CSIPs.

D. (1) OPEN 1CS-291 and 1CS-292 THEN SHUT 1CS-165 and 1CS-166 (2) To provide an alternate source of borated water to the CSIPs.

Thursday, May 19, 2016 5:04:39 PM 79

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: AOP-002 directs the candidate to first start a boric acid pump in order to establish a Boration flow path. If a boric acid pump is not available the operator is directed to open 1CS-291(LCV-115B) and 1CS-292(LCV-115D) RWST supply to the CSIPs, then shut 1CS-165(LCV-115C) and 1CS-166(LCV-115E) to provide a Boration flow path from the RWST to the CSIP suction.

A. Incorrect. The first part is plausible since AOP-002 Section 3.0 step 7 identifies this as part of the actions required to align the alternate boric acid flow path; however this is incorrect because no boric acid pump is running. The second part is plausible since VCT level is at the value which an auto make up will occur and this action will raise VCT level to eliminate the potential for gas binding due to low VCT level; however this is incorrect because gas binding is not a concern until the VCT level is below 5%.

B. Incorrect. The first part is plausible see A(1). The second part is correct.

C. Incorrect. The first part is correct. The second part is plausible see A(2).

D. Correct.

Thursday, May 19, 2016 5:04:39 PM 80

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 004 Chemical and Volume Control / 2 004K5.31; Knowledge of the operational implications of the following concepts as they apply to the CVCS: Purpose of flow path around boric acid storage tank (CFR: 41.5/ 45.7)

Importance Rating: 3.0 3.4 Technical

Reference:

AOP-002, Step 6, Rev. 24, Page 5 References to be provided: None Learning Objective: AOP-LP-3.2 Objective 2.a Question Origin: New Comments: HNP does not have a flow path around the boric acid storage tank. Attempted to match the K/A by having the student determine the operational implications of the failure of 1CS-278, Emergency Boric Acid Addition which bypasses the boric acid blender(mixing tee). The student is given the purpose of the flowpath in the stem of the question which is to deliver boric acid from the BAT to the CSIP suction during emergency boration.

See drawing on reference page.

Phonecon 5/03: HNP states that they are concerned about a possible K/A match because the plant design lacks a bypass around the BAST. I dont think this one matches the K/A. Mark Bates didnt think so either.

Mark and I agree that the line from the RWST through 291 & 292 is effectively a flowpath around the BAST, so if you wanted to do something with that you could.

Whether by massaging this question or finding/building another one.

Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 81

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 031/BANK/C/A//GP-002/NONE//005K5.05/

Given the following plant conditions:

- The RCS is in solid plant operation

- 'A' CSIP is in service

- 'A' train RHR is in service providing both core cooling and low pressure letdown

- Letdown Line Pressure Control valve PCV-145, (1CS-38) is in AUTO

- Charging Flow Control valve FCV-122 is being operated with its controller in MANUAL with demand set at 20%

Which ONE of the following will raise RCS pressure?

A. 'A' RHR Pump trips B. Loss of Instrument Air to Letdown Pressure Control Valve, (1CS-38)

C. FK-122.1, Charging Flow Controller (1CS-231), is adjusted towards 0% demand D. 1CC-146, RHR HX Outlet Throttle valve is opened raising flow to the 'A' RHR Heat Exchanger Thursday, May 19, 2016 5:04:39 PM 82

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: RHR Pump trip will take away the motive force for Letdown in this circumstance. With Charging in manual, inflow will be greater than outflow, and pressure will rise.

A. Correct.

B. Incorrect. Plausible since the loss of air to 1CS-28, RHR Letdown, will shut this valve raising RCS pressure; however this is incorrect because 1CS-38 is a 'fail open' valve, so pressure would be reduced.

C. Incorrect. Plausible since reducing the setpoint on controllers such as the PRZ Master Pressure controller, Steam Dump controller or SG PORV controller will result in the assosciated valve opening; however this is incorrect because reducing demand on the charging flow control valve will cause the valve to close, lowering charging flow and reduce RCS pressure while the plant is solid.

D. Incorrect. Plausible since the CCW to the RHR heat exchanger will rise; the candidate may misapply this concept and determine the RHR system flow has risen which would raise RCS pressure; however this is incorrect because the heat exchanger outlet temperature would lower reducing RCS temperature and reduce RCS pressure while the plant is solid.

005 Residual Heat Removal / 4 005K5.05; Knowledge of the operational implications of the following concepts as they apply the RHRS: Plant response during "solid plant": Pressure change due to the relative incompressibility of water (CFR: 41.5 / 45.7)

Thursday, May 19, 2016 5:04:39 PM 83

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Importance Rating: 2.7 3.1 Technical

Reference:

GP-002, P&L 38, Rev 63, Page 9 References to be provided: None Learning Objective: GP-LP-3.2. Objective 1 Question Origin: Bank Comments: Ask Michael about the K/A...we don't have any HNP specific ties with Nil ductility transition temperature and we are not able to create a valid HNP specific question.

Phonecon 4/15: HNP states that they are unable to generate a question with plausible distractors on the topic item of Nil ductility transition temperature (brittle fracture), so I committed to providing a new K/A. T2G1 K5 is at minimum, so need to stay in K5. In 005 K5, there are 4 others with > 2.5: .02, .03.

05, .09. Assigned them 1-4, randomly chose: 3.

New K/A 005K5.05: Knowledge of the operational implications of the following concepts as they apply the RHRS: Plant response during "solid plant": Pressure change due to the relative incompressibility of water.

Fleet review of question developed for K/A 005K5.05 resulted in disagreement among the exam writer's that the question may not fully meet the K/A. 2 of the 3 exam writer's felt the question met the K/A based on the candidate having to analyze the conditions in the stem, determine that the plant is solid, then evaluate the 4 operational implications to determine which one would result in the plant response identified in the stem of the question. The exam writer originally in disagreement agreed with K/A match after the explanation of how the question meets the K/A but recommended the question be discussed with the NRC to determine if the question meets the K/A as currently developed.

Phonecon 5/03: HNP states that they are concerned about a possible K/A match based on feedback during their exam review by fleet exam writer's. I think it meets the K/A. I thought distractor A, A CSIP trips was weak (my note to myself was, Why plausible?"). Reading the D/A didnt change my mind. Im not sure thats possible to write, so we might need a different distractor. Distractor D I thought was weak, because its very similar to a pump tripping. Not as sudden, of course, but you get to the same end-state. After reading the D/A I thought, Okay, maybe.. Still weak in my opinion, but plausible enough.

Thursday, May 19, 2016 5:04:39 PM 84

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 85

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 032/BANK/FUNDAMENTAL//ALB-004/NONE//006K1.02/

Which ONE of the following windows on ALB-004 annunciation will coincide with the automatic operation of 1SI-301, Containment Sump to RHR Pump Suction Valves?

A. Window 2-2, Refueling Water Storage Tank LOW Level.

B. Window 2-3, Refueling Water Storage Tank LOW-LOW Level ALERT.

C. Window 2-4, Refueling Water Storage Tank 2/4 LOW-LOW Level.

D. Window 2-5, Refueling Water Storage Tank EMPTY.

Plausibility and Answer Analysis Reason answer is correct: In accordance with APP-ALB-004, Window 2-4 the following automatic actions occur upon receiving this alarm.

Automatic Functions:

a. With an SI signal and 2/4 low-low levels present, CNMT Sump to RHR Pump suction valves 1SI-300, 1SI-301, 1SI-310, and 1SI-311 open.

A. Incorrect. Plausible since 1SI-301 will automatically open with a lowering RWST level of 23.4%; however this is incorrect because this annunciator alarms at 94.3% which is above the automatic swap over level of 23.4%.

B. Incorrect. Plausible since 1SI-301 will automatically open with a lowering RWST level of 23.4%; however this is incorrect because this annunciator alarms when 23.4% is sensed on any ONE of the 4 RWST level transmitters reaches 23.4% and automatic swap over requires 2/4 level transmitters to reach a level of 23.4%.

C. Correct.

D. Incorrect. Plausible since 1SI-301 will automatically open with a lowering RWST level of 23.4%; however this is incorrect because this annunciator alarms at 3% which is below the automatic swap over level of 23.4%.

Thursday, May 19, 2016 5:04:39 PM 86

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 006 Emergency Core Cooling / 2/3 006K1.02; Knowledge of the physical connections and/or cause effect relationships between the ECCS and the following systems: ESFAS (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Importance Rating: 4.3 4.6 Technical

Reference:

APP-ALB-004, Window 2-4, Rev 18, Page 9 References to be provided: None Learning Objective: Student Text SIS, Objective 7.c Question Origin: Bank Comments: 006K1.02 - System 006 is in both Safety Function 2, Inventory Control, and Safety Function 3, Reactor Pressure Control.

I note that Form ES-401-2 does not list Safety Functions for T2G1, so it doesnt matter which SF the question is written to, as long as it matches the K/A.

Phonecon 3/23: HNP concurs.

Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 87

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 033/PREVIOUS/C/A//APP-ALB-009/NONE//007A1.03/

Given the following plant conditions:

- The unit is operating at 100% power

- The crew is responding to a leaking PRZ Safety valve Time PRT Temp Safety Tailpipe Temp 1000 95°F 145°F 1005 115°F 255°F 1010 122°F 275°F 1015 146°F 403°F Which ONE of the following is the first time that annunciator ALB-009-8-1, PRT High-Low Level Press or Temp, will alarm?

A. 1000 B. 1005 C. 1010 D. 1015 Thursday, May 19, 2016 5:04:39 PM 88

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Annunciator ALB-009-8-1 has multiple inputs that causes the annunciator to alarm. One of which is high temperature which has a setpoint of 120°F.

At 1010 the PRT temperature is above the temperature at which the annunciator will go into alarm.

A. Incorrect. Plausible since at this PRT input temperature annunciator ALB-009-8-2, Pressurizer Relief Discharge High Temp goes into alarm. The alarm comes on when the PRZ PORV discharge line temperature exceeds 140°F as sensed by TI-463. To confirm the alarm the operator checks PRT level, pressure, and temperature for corresponding changes.

B. Incorrect. Plausible since at this PRT input temperature annunciator ALB-009-8-3, Pressurizer Safety Relief Discharge High Temp goes into alarm. The alarm comes on when the PRZ Safety valve discharge line temperature exceeds 250°F as sensed by TI-465, TI-467, or TI-469. To confirm the alarm the operator checks Safety valve discharge line temperatures and PRT level, pressure, and temperature for corresponding changes.

C. Correct.

D. Incorrect. Plausible since an input temperature of > 400°F would require that the Safety valve be declared inoperable. It is an indication that there has been a loss of loop seal. The loss of loop seal may cause the associated safety lift setpoint to shift down to normal operating pressure. PRT input temperature annunciator ALB-009-8-3, Pressurizer Safety Relief Discharge High Temp would have already been in alarm when the PRZ Safety valve discharge line temperature exceeded 250°F as sensed by TI-465, TI-467, or TI-469. The alarm would have been confirmed by they operator checking Safety valve discharge line temperatures and PRT level, pressure, and temperature for corresponding changes.

Thursday, May 19, 2016 5:04:39 PM 89

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 007 Pressurizer Relief/Quench Tank / 5 007A1.03; Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

Monitoring quench tank temperature (CFR: 41.5 / 45.5)

Importance Rating: 2.6 2.7 Technical

Reference:

ALB-009-8-1, Page 29, Rev. 17 References to be provided: None Learning Objective: PRZ Objective 5.d Question Origin: Previous 2014 NRC RO 35 radomly selected Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 90

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 034/BANK/C/A//EOP-E-0, ATT 3/NONE//064K3.01/

Given the following plant conditions:

- The plant is operating at 100%

- The TDAFW pump is under clearance for a bearing replacement Subsequently the following occurs:

- Off-site Power is lost

- The 1B-SB sequencer starts, however the 'B' MDAFW pump sequencer start relay fails, resulting in failure of the 'B' MDAFW pump to start

- The 1A-SA Diesel fails to start Which ONE of the following describes the method for restoration of feedwater?

The 'B' MDAFW pump .

A. must be started by the operator B. cannot be started until the sequencer is reset C. will start when at least 2 SG levels are less than 25%

D. will start in Load Block 9 due to loss of both Main Feed pumps Plausibility and Answer Analysis Reason answer is correct: The failure of the 'B' MDAFW pump sequencer start relay will prevent the automatic start of the 'B' MDAFW pump. Both EOP-E-0 and the EOP-Users Guide direct the operator to attempt to manually start equipment that should have automatically started once the sequencer has reached Load Block 9.

A. Correct.

B. Incorrect. Plausible since the actuation of the sequencer disable redundant automatic start signals during its operation; however this is incorrect because the manual operation of the MCB switch is not defeated during sequencer operation.

C. Incorrect. Plausible since this is an available start signal under normal conditions for the AFW system pumps; however this is incorrect because during sequencer operations this start signal is defeated and will not be restored until the sequencer is reset.

D. Incorrect. Plausible since this is an available start signal under normal conditions; however this is incorrect because during sequencer operations this start signal is defeated and will not be restored until the sequencer is reset.

Thursday, May 19, 2016 5:04:39 PM 91

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 064 Emergency Diesel Generator / 6 064K3.01; Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: Systems controlled by automatic loader (CFR: 41.7 / 45.6)

Importance Rating: 3.8 4.1 Technical

Reference:

EOP-E-0, Attachment 3, Step 12, Rev 4, Page 57 EOP-Users-Guide, Step 6.4, Rev 46, Page 36 References to be provided: None Learning Objective: Student Text Sequencer, Objective 4 Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 92

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 035/NEW/FUNDAMENTAL//5-S-1322, CVCS ST/NONE/EARLY/008A2.05/SAT Given the following plant conditions:

- The unit is operating at 100% power

- The Instrument Air line on 1CC-337, TK-144 LTDN Temperature, breaks Which ONE of the following completes the statements below?

The failed position of 1CC-337 is (1) .

In accordance with AD-OP-ALL-0203, Reactivity Management the crew will (2) once control of 1CC-337 is re-established.

A. (1) SHUT (2) reduce Reactor power below 100%

B. (1) SHUT (2) maintain current Reactor power C. (1) OPEN (2) reduce Reactor power below 100%

D. (1) OPEN (2) maintain current Reactor power Thursday, May 19, 2016 5:04:39 PM 93

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: On a loss of IA to 1CC-337, TCV-144 fails full open, raising CCW flow through the letdown heat exchanger, this causes temperature to lower. Lower letdown temperature results in absorption of BA in the demineralizers, resulting in a dilution of the RCS, raising Reactor power to >100%

requiring the crew to reduce power in accordance with AD-OP-ALL-0203.

A. Incorrect. The first part is plausible since multiple CCW isolation valves fail in the shut position.

1CC-304 and 305, Gross Failed Fuel Detector sample isolation valves, 1DW-15, Makeup to the CCW Surge Tank, and both RHR sample cooler heat exchangers are examples; however this is incorrect since TCV-144 opens to ensure the LD Heat Exchanger effluent temperature is reduced in order to protect the downstream piping from over pressurization. The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since letdown system temperature control valve 1CS-50, TCV-143 is a three way valve that realigns to bypass around or align process water flow through the demineralizer beds to protect the demineralizer resin material from damage due to high temperature. If the candidate misapplies the letdown system TCV design and determines TCV-144 operates the same as TCV-143 they would assume that valve would shut to align CCW flow around the heat exchanger. Since the letdown temperature would rise and reduce the absorption of boric acid in the demineralizers. This would result in an uncontrolled boration of the RCS and reduce Reactor power requiring the crew to stabilize the plant at the current reactor power once control of 1CC-337 has been re-established. In accordance with AD-OP-ALL-0203 the crew should stablize the unit at a power level at or below the pre-transient level. However this is incorrect since TCV-144 opens to align more flow through the LD Heat Exchanger.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since letdown system temperature control valve 1CS-50, TCV-143 is a three way valve that realigns to bypass around or align process water flow through the demineralizer beds to protect the demineralizer resin material from damage due to high temperature. If the candidate misapplies the letdown system TCV design and determines TCV-144 operates the same as TCV-143 they would assume that valve would open to align CCW flow around the heat exchanger. Since the letdown temperature would rise and reduce the absorption of boric acid in the demineralizers. This would result in an uncontrolled boration of the RCS and reduce Reactor power requiring the crew to stabilize the plant at the current reactor power once control of 1CC-337 has been re-established. In accordance with AD-OP-ALL-0203 the crew should stablize the unit at a power level at or below the pre-transient level. However this is incorrect because the will stabilize at at power level above 100% and power must be reduce to a power level at or below the pre-transient level.

Thursday, May 19, 2016 5:04:39 PM 94

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 008 Component Cooling Water / 8 008A2.05; Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effect of loss of instrument and control air on the position of the CCW valves that are air operated (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Importance Rating: 3.3 3.5 Technical

Reference:

SFD 2165-S-1322 Student Text CVCS, Page 89 AD-OP-ALL-0203, Section 5.2.1, Step 2.b, Rev 2, Page 34 References to be provided: None Learning Objective: Student Text CVCS, Objective 3.a, 11.f Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 95

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 036/BANK/FUNDAMENTAL///NONE//010K1.06/

While establishing a bubble in the PRZ per GP-002, Normal Plant Heatup From Cold Solid To Hot Subcritical Mode 5 To Mode 3, letdown pressure control valve 1CS-38, PK-145.1 LTDN Pressure, modulates open.

Which ONE of the following describes why 1CS-38 modulates open?

A. CCW heat load lowers B. Thermal expansion of liquid in the PRZ C. PRZ spray valves are shut while drawing a bubble D. Switchover of letdown to orifices from RHR-CVCS cross-connect Plausibility and Answer Analysis Reason answer is correct: Thermal expansion of the liquid due to the heaters being energized results in a pressure rise in the RCS. PK-145.1 opens to maintain letdown pressure, resulting in rising letdown flow.

A. Incorrect. Plausible since the letdown heat exchanger is cooled by CCW; however this is incorrect since temperature has little effect on the response of PK-145.1 with a bubble in the PRZ.

B. Correct.

C. Incorrect. Plausible since the spray valves are shut while a bubble is being drawn; however this is incorrect since PK-145.1 opens to maintain letdown line pressure, not RCS pressure.

D. Incorrect. Plausible since RHR letdown may be placed in service at low temperature and pressure conditions; however this is incorrect since RHR is not in service while drawing a bubble.

Thursday, May 19, 2016 5:04:39 PM 96

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 010 Pressurizer Pressure Control / 3 010K1.06; Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: CVCS (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Importance Rating: 2.9 3.1 Technical

Reference:

GP-002 References to be provided: None Learning Objective: GP-LP-3.2, Objective 2.b Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 97

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 037/NEW/C/A//ST PRZLC/5-S-1301//010K1.08/

Given the following plant conditions:

- The unit is operating at 100% power

- PT-444, PRZ Pressure, develops a leak from its sensing line Which ONE of the following completes the statement below?

PRZ PT-444 (1) inputs to the protective functions of RPS. The associated PRZ LT-461 indicated level will be (2) as a result of this leak.

(Reference provided)

A. (1) provides (2) higher B. (1) provides (2) lower C. (1) does NOT provide (2) higher D. (1) does NOT provide (2) lower Thursday, May 19, 2016 5:04:39 PM 98

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Two PRZ pressure detectors provide inputs for the PRZPCS (PT-444 and 445) and three other PRZ pressure detectors provide RPS and ESFAS protective (PT-455, 456 and 457) functions. PRZ LT-461 shares a common reference leg location with PRZ PT-444 and because the sensing line for PT-444 has developed a leak a level error can result from a reference leg leak. PRZ level indication would be erroneously high because the pressure sensed by the reference leg would lower, and the resulting differential pressure would lower and results in the indicated level being higher. This failure can be confirmed by comparison of the other level indications to identify the incorrect channel.

A. Incorrect. The first part is plausible since the PRZ level transmitter provide inputs to both the control and protection functions the candidate may conclude that the PRZ pressure transmitters function in the same fashion. However this is incorrect because PT-444 and 445 only provide input into the PRZ pressure control function and not the RPS or ESFAS protection functions.

PT-455, 456 and 457 provide the inputs for RPS. The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since the candidate has to recall the location of the level transmitter tap in reference to PT-444. The candidate could conclude the connection is on the variable side of the level transmitter and the indicated level will be the same as actual level and lower as inventory is lost through the leak.

However this is incorrect because the connection between LT-461 and PT-444 is via the reference leg and a leak from this location will reduce the transmitter D/P and LT-461 will indicate higher.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:39 PM 99

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 010 Pressurizer Pressure Control / 3 010K1.08; Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: PZR LCS (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Importance Rating: 3.2 3.5 Technical

Reference:

Student Text PRZLC References to be provided: 2165-S-1301, Sheet 2 Learning Objective: Student Text PRZLC, Objective 9.b Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 100

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 038/BANK/C/A//APP-ALB-011/NONE//012A3.04/

Given the following plant conditions:

- The unit is operating at 38% power

- Breaker 109, 6.9KV Aux Bus 1C trips open Which ONE of the following completes the statement below?

Based on the conditions above, the Reactor Trip Breaker's (1) light will be illuminated. Additionally, the P-8, Single Loop Low Flow Trip Blocked light (2) be illuminated on the Bypass Light Permissive Panel.

A. (1) red (2) will B. (1) red (2) will NOT C. (1) green (2) will D. (1) green (2) will NOT Thursday, May 19, 2016 5:04:39 PM 101

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Reactor Trip Breakers A and B remain closed (red light illuminated) because the P-8 logic is not satisfied. P-8 issatisfied when 2 of 4 Power Ranges exceed 49% RTP. The condition given in the stem is 38%. When power is between 10% and 49% two RCPs would have to trip to generate an automatic Reactor Trip signal.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since another protective function (AMSAC) is armed at 35%. However this is incorrect because P-8 is not satisfied until 2 of 4 PRNIs are above 49%.

C. Incorrect. The first part is plausible because the 1C 6.9KV Aux Bus is normally supplied by the 1A 6.9KV Aux Bus. The candidate may conclude both the A and the C RCP which will generate a Reactor Trip signal. The second part is plausible see A(2).

D. Incorrect. The first part is plausible se C(1). The second part is correct.

Thursday, May 19, 2016 5:04:39 PM 102

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 012 Reactor Protection / 7 012A3.04; Ability to monitor automatic operation of the RPS, including: Circuit breaker (CFR: 41.7 / 45.5)

Importance Rating: 2.8 2.9 Technical

Reference:

APP-ALB-011, Window 2-4, Rev 8, Page 6 References to be provided: None Learning Objective: Student Text RPS, Objective 8 Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 103

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 039/PREVIOUS/FUNDAMENTAL//AOP-024/NONE/PREVIOUS/013K2.01/

Which ONE of the following completes the statement below?

Instrument Buses (1) AND (2) provide power to the ESFAS Slave Relays.

A. (1) SI (2) SII B. (1) SII (2) SIII C. (1) SI (2) SIV D. (1) SIII (2) SIV Thursday, May 19, 2016 5:04:39 PM 104

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Train ESFAS slave relays are powered from Instrument Bus SI (SIV). A loss of SI or SIV will result in a loss of ESFAS functions driven by slave relays for that train.

A. Incorrect Plausible since the ESFAS relays are powered from the safety instrument buses. If SI or SIII is lost, the MCB controller for MDAFW pump flow control valves will not be operable; flow control valves will not shut on an AFW isolation signal and will not open on an auto open signal. Train ESFAS slave relays are powered from Instrument Bus SI (SIV). A loss of SI or SIV will result in a loss of ESFAS functions driven by slave relays for that train. A loss of SI will cause a loss of 'A' Train ONLY the question is asking for BOTH 'A' and 'B' Train.

B. Incorrect Plausible since the ESFAS relays are powered from the safety instrument buses. If power is lost to Instrument Bus SII (B Train and TDAFW) or SIII (A Train) the associated AFW pump suction pressure instrument will read low. If the AFW pump is running, it will not trip on Lo-Lo suction pressure nor will it be prevented from being started. Additionally, if power is lost to Instrument Bus SII (B Train) or SIII (A Train), the associated CNMT Spray Additive Tank level indicators will read empty but their associated CNMT Spray Chemical Addition Valve will not automatically shut. If necessary, the valve(s) may be manually operated.

C. Correct D. Incorrect Plausible since the ESFAS slave relays are powered from Instrument Bus SI (SIV). To answer this question it would take BOTH SI and SIV and only one of the two (SIV) are listed. If power is lost to Instrument Bus SII (B Train and TDAFW) or SIII (A Train) the associated AFW pump suction pressure instrument will read low. If the AFW pump is running, it will not trip on Lo-Lo suction pressure nor will it be prevented from being started.

Train ESFAS slave relays are powered from Instrument Bus SI (SIV). A loss of SI or SIV will result in a loss of ESFAS functions driven by slave relays for that train.

Thursday, May 19, 2016 5:04:39 PM 105

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 013 Engineered Safety Features Actuation / 2 013K2.01; Knowledge of bus power supplies to the following: ESFAS/safeguards equipment control (CFR: 41.7)

Importance Rating: 3.6 3.8 Technical

Reference:

AOP-024-BD Rev. 20, page 2 Discussion #5 References to be provided: None Learning Objective: ESFAS Obj. 2 Question Origin: Previous 2013 NRC RO question 40 Comments: None Tier/Group: T2G1 Thursday, May 19, 2016 5:04:39 PM 106

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 040/BANK/C/A//1364-000871/NONE//013K6.01/

Given the following plant conditions:

- The plant is operating at 100% power

- Instrument Bus SIII, is de-energized and actions are being taken in accordance with AOP-024, Loss of Uninterruptible Power Supply

- PT-953, Containment Pressure Channel IV, then fails high Which ONE of the following describes the effect on the Safety Injection (SI) AND Containment Spray Actuation Signal (CSAS) systems?

SI CSAS A. Not actuated Not actuated B. Actuated Not actuated C. Not actuated Actuated D. Actuated Actuated Plausibility and Answer Analysis Reason answer is correct: An SI actuation (deenergized to actuate) will occur, but a CSAS (energized to actuate) will not occur unless another energized channel senses a high pressure condition.

A. Incorrect. Plausible since CSAS is energized to actuate and 1 channel is in a deenergized condition so CSAS will not occur, but the 2 failed channels will cause an SI actuation.

B. Correct.

C. Incorrect. Plausible since one of the two signals is energized to actuate and the other is deenergized to actuate, but SI is deenergize to actuate and CSAS is energized to actuate.

D. Incorrect. Plausible since the 2 failed channels will cause an SI actuation, but CSAS is energized to actuate and 1 channel is in a deenergized condition so CSAS will not occur.

Thursday, May 19, 2016 5:04:39 PM 107

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 013 Engineered Safety Features Actuation / 2 013K6.01; Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS: Sensors and detectors (CFR: 41.7 / 45.5 to 45.8)

Importance Rating: 2.7 3.1 Technical

Reference:

Plant Drawing 1364-000871 References to be provided: None Learning Objective: Student Text ESFAS, Objective 8 Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 108

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 041/BANK/FUNDAMENTAL//OP-169/NONE//022K2.01/

Which ONE of the following identifies the power supply to the fan motors for Containment Fan Cooler AH-1?

A. Aux Bus 1D1 B. Aux Bus 1E1 C. MCC 1A34-SA D. MCC 1B22-SB Plausibility and Answer Analysis Reason answer is correct: The power supply to the 'B' Train Containment Fan Coolers, AH-1 and AH-4, is safety related MCC 1B22-SB.

A. Incorrect. Plausible since the Aux Busses supply the Containment Fan Coil Units; however this is incorrect since the Containment Fan Coolers are powered from the safety bus.

B. Incorrect. Plausible since the Aux Busses supply the Containment Fan Coil Units; however this is incorrect since the Containment Fan Coolers are powered from the safety bus.

C. Incorrect. Plausible since this is the correct power supply for the 'A' Train Containment Fan Coolers; however this is incorrect since AH-1 is a 'B' Train component.

D. Correct.

Thursday, May 19, 2016 5:04:39 PM 109

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 022 Containment Cooling / 5 022K2.01; Knowledge of power supplies to the following: Containment cooling fans (CFR: 41.7)

Importance Rating: 3.0 3.1 Technical

Reference:

OP-169, Attachment 1, Rev 26, Page 49 References to be provided: None Learning Objective: Student Text CCS, Objective 1.a Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 110

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 042/BANK/FUNDAMENTAL//1364-000871/NONE//026A4.01/

Which ONE of the following identifies the MINIMUM required logic for MANUAL actuation of the Containment Spray System using the MCB Containment Spray activation switches?

A. ANY one of the four switches.

B. ANY two of the four switches.

C. EITHER the LEFT two switches OR the RIGHT two switches.

D. EITHER the INSIDE two switches OR the OUTSIDE two switches.

Plausibility and Answer Analysis Reason answer is correct: There are 4 channels for Containment Spray actuation logic on MCB. (2 channels per train) To manually initiate Containment Spray the operator must turn 2 switches on either the left panel or 2 on the right panel (BOTH switches from either train). In addition the Containment Spray logic is an energize to actuate circuit.

A. Incorrect. Plausible since each switch is labeled as CSAS, the operator may misinterpret the labeling and incorrectly believe 1 switch would be sufficient to actuate spray.

B. Incorrect. Plausible because it is partially correct. The reset does require operation of 2 of 4 switches, however they must be the 2 right or the 2 left switches in combination.

C. Correct.

D. Incorrect. Plausible because it is partially correct. The reset does require operation of 2 of 4 switches, however they must be the 2 right or the 2 left switches in combination.

Thursday, May 19, 2016 5:04:39 PM 111

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 026 Containment Spray / 5 026A4.01; Ability to manually operate and/or monitor in the control room: CSS controls (CFR: 41.7 / 45.5 to 45.8)

Importance Rating: 4.5 4.3 Technical

Reference:

Logic Drawing EMDRAC 1364-000871, CSS Actuation logic References to be provided: None Learning Objective: Student Text CSS, Objective 4 Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:39 PM 112

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 043/BANK/FUNDAMENTAL//ERG-BKGD-FR-Z.1/NONE//026G2.4.20/

Which ONE of the following is the REASON why Containment Spray is operated in accordance with the guidance from EOP-ECA-1.1, Loss of Emergency Coolant Recirculation, while implementing EOP-FR-Z.1, Response to High Containment Pressure?

A. Actions required by ECA's normally have priority over those in FR's.

B. Containment pressure is expected to be below the HI-3 reset setpoint.

C. Conservation of RWST inventory is required to ensure availability of core cooling.

D. There is no available suction source for the pumps if the recirc sump is unavailable.

Plausibility and Answer Analysis Reason answer is correct: In accordance with the ERG Background Document, Guideline ECA-1.1 uses a less restrictive criteria, which permits reduced spray pump operation depending on RWST level, containment pressure and number of emergency fan coolers operating. The less restrictive criteria for containment spray operation is used in guideline ECA-1.1 since recirculation flow to the RCS is not available and it is very important to conserve RWST water, if possible, by stopping containment spray pumps.

A. Incorrect. Plausible since the EOP-Users Guide states "certain E-, ES- and ECA procedures take precedence over some FR procedures because of their treatment of specific initiating events"; however this is incorrect because this precedence will be identified in a Note or Caution at the beginning of the applicable EOP.

B. Incorrect. Plausible since the RWST minimum volume is designed to "assure sufficient water is available within containment to permit recirculation cooling flow to the core and the reactor remain subcritical in cold conditions" it may be assumed the equilibrium containment presssure once the RWST volume is exhausted will be below the HI-3 reset value.

However this is incorrect because the reason to use the less restrictive criteria for containment spray operation from EOP-ECA-1.1.

C. Correct.

D. Incorrect. Plausible since the Containment Spray pump take a suction from the because RWST or the Containment Recirculation Sump; however this is incorrect because EOP-ECA-1.1 determines the availablility of the Containment Recirculation Sump based on the status of the RHR system not the Containment Spray system.

Thursday, May 19, 2016 5:04:39 PM 113

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 026 Containment Spray / 5 026G2.4.20; Knowledge of the operational implications of EOP warnings, cautions, and notes.

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: 3.8 4.3 Technical

Reference:

EOP-FR-Z.1, Step 3, Rev 0, Page 4 ERG-BKGD-FR-Z.1, Step 3, Rev 2, Page 9 SDD-FR-Z.1, Step 3, Rev 0, Page 2 References to be provided: None Learning Objective: EOP-LP-3.13, Objective 4.b Question Origin: Bank Comments: At HNP ERG 3-CAUTION is incorporated into associated step because it involves operator action.

Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 114

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 044/PREVIOUS/FUNDAMENTAL///NONE//039A3.02/

Given the following plant conditions:

- The unit is operating at 100% power

- A Main Steam line rupture in the Turbine building has occurred

- The crew has manually tripped the Reactor Which ONE of the following completes the statement below?

The Turbine Ventilating valves 1GS-97, 1GS-98 are expected to (1) AND the MSR Non-Return valves 1HD-2, 1HD-3, 1HD-302, 1HD-303 are expected to (2) .

Valve Titles:

Turbine Ventilating valves 1GS-97, HP Turbine Vent to Cond (FCV-01TA-0415B) 1GS-98, HP Turbine Vent to Cond (FCV-01TA-0415A)

MSR Non-Return valves 1HD-2, MSR 1A-NNS Outlet to MSDT 1A-NNS 1HD-3, MSRDT 1A-NNS Outlet to 5-1A-NNS 1HD-302, MSR 1B-NNS Outlet to MSDT 1B-NNS 1HD-303, MSRDT 1B-NNS Outlet to 5-1B-NNS A. (1) shut (2) shut B. (1) shut (2) open C. (1) open (2) shut D. (1) open (2) open Thursday, May 19, 2016 5:04:40 PM 115

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Any Reactor Trip generates a Turbine Trip signal. Since a Turbine Trip signal is present all of the Turbine Throttle valves would be shut and the Auto Stop Trip header would be depressurized causing the Turbine Ventilating valves to OPEN and MSR Non-Return valves to SHUT. 1GS-97 and 1GS-98 automatically open, while 1HD-2, 1HD-3, 1HD-302 and 1HD-303 shut automatically based on the status of the Turbine Throttle valves or the Auto Stop Trip header pressure which are used to determine if the Turbine is tripped or latched.

A. Incorrect. The first part is plausible since with the Turbine tripped 1st stage pressure is reduced to the pressure of the Main Condenser which is less than the 5 psig. The Gland Sealing Steam Spillover Regulator to the condenser to modulates open if header pressure is > 5 psig and therefore the valve would be shut on a turbine trip, however the ventilating valve open to provide a flowpath to the condenser for the steam trapped in the HP turbine. The second part is correct.

B. Incorrect. The first part is plausible since with the Turbine tripped 1st stage pressure is reduced to the pressure of the Main Condenser which is less than the 5 psig. The Gland Sealing Steam Spillover Regulator to the condenser to modulates open if header pressure is > 5 psig and therefore the valve would be shut on a turbine trip, however the ventilating valve open to provide a flowpath to the condenser for the steam trapped in the HP turbine. The second part is plausible since the turbine drain valves automatically open following a turbine trip to provide a drain path for the residual steam trapped in the turbine as this steam begins to condense, however this is .

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since the turbine drain valves automatically open following a turbine trip to provide a drain path for the residual steam trapped in the turbine as this steam begins to condense.

Thursday, May 19, 2016 5:04:40 PM 116

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 039 Main and Reheat Steam / 4 039A3.02; Ability to monitor automatic operation of the MRSS, including: Isolation of the MRSS (CFR: 41.5 / 45.5)

Importance Rating: 3.1 3.5 Technical

Reference:

MT Student text MSR Student text References to be provided: None Learning Objective: Lesson Plan MT, Objective 9 Lesson Plan MSR, Objective 4.e Question Origin: Previous 2014 NRC RO 45 radomly selected Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 117

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 045/BANK/C/A//1364-000870, ALB-014/NONE//039A4.04/

Given the following plant conditions:

- A Reactor Trip and Safety Injection have actuated

- A MSLI has actuated

- Steam Generator parameters have lowered to the following values:

SG NR Level Pressure A 32% 870 psig B 12% 420 psig C 34% 830 psig Which ONE of the following identifies the expected position of the following valves?

(1) 1AF-143, STM TURB AUX FW B Isolation (2) 1MS-70, B SG to AFW Turbine (Assume NO operator actions have been taken)

A. (1) OPEN (2) OPEN B. (1) OPEN (2) CLOSED C. (1) CLOSED (2) CLOSED D. (1) CLOSED (2) OPEN Thursday, May 19, 2016 5:04:40 PM 118

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: FCV-2071B has received an AFW isolation signal due to the Main Steam Isolation signal coincident with high differential steam pressure. No automatic start signal has been generated for the TDAFW pump (only 1 SG low-low level has occurred) so the steam supply valves should be closed.

A. Incorrect. Plausible since the valve controller is set at 100% per OP-137, and a MDAFWP start signal has been generated which would open the FCVs from the MDAFW pumps (not TDAFW), however it received an AFW isolation signal due to the Main Steam Isolation signal coincident with high differential steam pressure. Second half is plausible since this valve would open on low-low level in 2 SGs.

B. Incorrect. Plausible since the valve controller is set at 100% per OP-137, and a MDAFWP start signal has been generated which would open the FCVs from the MDAFW pumps (not TDAFW), however it received an AFW isolation signal due to the Main Steam Isolation signal coincident with high differential steam pressure. Second half is correct since no automatic start signal has been generated for the TDAFW pump (only 1 SG low-low level has occurred) so the steam supply valves should be closed C. Correct.

D. Incorrect. First half is correct since the FCV received an AFW isolation signal due to the Main Steam Isolation signal coincident with high differential steam pressure. Second half is plausible since this valve would open on low-low level in 2 SGs.

Thursday, May 19, 2016 5:04:40 PM 119

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 039 Main and Reheat Steam / 4 039A4.04; Ability to manually operate and/or monitor in the control room: Emergency feedwater pump turbines (CFR: 41.7 / 45.5 to 45.8)

Importance Rating: 3.8 3.9 Technical

Reference:

ALB-014 2-1A, Rev 25, Page 9 Logic Drawing EMDRAC 1364-000870, AFW Isolation logic References to be provided: None Learning Objective: Student Text MSSS, Objective 4 Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 120

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 046/BANK/FUNDAMENTAL//OP-134/NONE//059A3.03/

Given the following plant conditions:

- A plant startup is in progress

- The 'A' Condensate pump will be the first Condensate pump started in accordace with OP-134, Condensate System Which ONE of the following completes the statements below?

The discharge valve for the 'A' Condensate pump must be (1) before the pump motor will energize.

Once the Main Feed pump suction header is pressurized to normal operating pressure, the discharge valve for (2) Condensate pump(s) fully open(s).

A. (1) 10% - 13% open (2) ONLY A B. (1) 10% - 13% open (2) BOTH C. (1) closed (2) ONLY A D. (1) closed (2) BOTH Thursday, May 19, 2016 5:04:40 PM 121

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: OP-134 P&L #9 states: To satisfy Condensate Pump start permissive, the Condensate Pump Discharge Valve has to be open 10 to 13% for the first pump to start and the discharge valve fully open for the second pump to start. With neither Condensate Pump running, taking the Condensate Pump Discharge Valve control switch to the OPEN position and releasing it will cause the valve to open to the 10 to 13% open position. A note in section 5.3 for starting the first condensate pump states: When the breaker for the first condensate pump shuts and MFP suction pressure reaches 210 psig (sensed by PS-2201), the discharge valves for BOTH condensate pumps stroke full open.

A. Incorrect. The first part is correct. The second part is plausible since the A Condensate Pump discharge valve does open after starting the first Condensate pump; however this is incorrect because BOTH Condensate pump discharge valves will automatically fully open when MFW pump suction pressure reaches 210 psig (sensed by PS-2201)..

B. Correct.

C. Incorrect. The first part is plausible since other plant systems must have their discharge valves shut for water hammer concerns prior to starting the pump; however this is incorrect for the operation of the Condensate Pump Breaker. The second part is plausible since the A Condensate Pump discharge valve does open after starting the first Condensate pump; however this is incorrect because BOTH Condensate pump discharge valves will automatically fully open when MFW pump suction pressure reaches 210 psig (sensed by PS-2201).

D. Incorrect. The first part is plausible since other plant systems must have their discharge valves shut for water hammer concerns prior to starting the pump; however this is incorrect for the operation of the Condensate Pump Breaker. The second part is correct.

Thursday, May 19, 2016 5:04:40 PM 122

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 059 Main Feedwater / 4 059A3.03; Ability to monitor automatic operation of the MFW, including: Feedwater pump suction flow pressure (CFR: 41.7 / 45.5)

Importance Rating: 2.5 2.6 Technical

Reference:

OP-134, P&L 9, Rev 58, Page 8 OP-134, Note prior to Section 5.3.2 Step 10, Rev 58, Page 16 References to be provided: None Learning Objective: Student Text CFW, Objective 8.b Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 123

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 047/NEW/C/A//APP-ALB-017/NONE//061A2.05/

Given the following plant conditions:

- The unit is operating at 98% power

- The TDAFW Pump is running Subsequently the following occurs:

- ALB-017-7-3, Aux Feedwater Pump Turbine Gov Control Power Failure, alarms

- ALB-017-7-4, Aux Feedwater Pump Turbine Trip, alarms 30 seconds later Which ONE of the following completes the statement below?

In accordance with APP-ALB-017:

The TDAFW pump governor valve will fail (1) . Performing a corrective action of shutting BOTH 1MS-70 SA and 1MS-72 SB (2) be required.

Valve Titles:

1MS-70 SA, Main Steam B To Aux FW Turbine 1MS-72 SB, Main Steam C To Aux FW Turbine A. (1) shut (2) will B. (1) shut (2) will NOT C. (1) open (2) will D. (1) open (2) will NOT Thursday, May 19, 2016 5:04:40 PM 124

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: APP-ALB-017 window 7-3 is an indication that the TDAFW Pump has loss control power to the governor valve. Without power the governor fails open and is not able to control the turbine speed which results in an overspeed condition on the TDAFW Pump. The APP response provides guidance the the Trip and Throttle valve will shut automatically and the operator is required to manually shut the steam admission valve to the TDAFW Pump 1MS-70 SA and 1MS-72 SB.

A. Incorrect. The first part is plausible since the governor valve is controlled via a leakage assembly similar to the Trip and Throttle valve and the candidate may misapply this operation to the cause of the turbine trip alarm and determine the governor valve is shut. However this is incorrect because the governor valve is normally held open by a spring and a hydraulic operator overcomes the spring pressure to modulate the turbine speed.

The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since the trip and throttle valve operates when the trip setpoint is reached and shut the steam supply to the TDAFW pump. But inaccordance with the APP the corrective action is to shut BOTH 1MS-70 and 1MS-72.

C. Correct.

D. Correct. The first part is correct. The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:40 PM 125

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 061 Auxiliary/Emergency Feedwater / 4 061A2.05; Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Automatic control malfunction (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Importance Rating: 3.1 3.4 Technical

Reference:

APP-ALB-017, Window 7-3, Rev 14, Page 19 and 20 References to be provided: None Learning Objective: Student Text AFW, Objective 9.c Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 126

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 048/BANK/FUNDAMENTAL//OP-155/NONE//062K3.02/

Given the following plant conditions:

- The 1A-SA EDG started automatically due to an undervoltage condition on 6.9-kV Bus 1A-SA.

Subsequently a 1A-SA Emerg Bus Differential and a Low Lube Oil Pressure signal occur.

Based on the conditions above, which ONE of the following identifies the signal(s), if any, that would result in a trip of the 1A-SA EDG?

A. NEITHER Emerg Bus Differential NOR Low Lube Oil Pressure B. BOTH Emerg Bus Differential AND Low Lube Oil Pressure C. Low Lube Oil Pressure ONLY D. Emerg Bus Differential ONLY Thursday, May 19, 2016 5:04:40 PM 127

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The automatic start of the Diesel due to the undervoltage condition on the 1A-SA Safety Bus results in the bypassing of non-emergency diesel generator trips, additionally with Breaker 105, Emergency Bus A-SA To Aux Bus D Tie Breaker, open the non-emergency diesel generator trips are bypassed. The following five Diesel trips remain active during emergency start conditions:

Overspeed Trip Emergency Stop (Manual)

Emergency Bus Differential Trip Loss Of Both Gen Pot CKS Trip Gen Diff Protection Trip A. Incorrect. Plausible since the diesel started from an emergency start signal due to the undervoltage condition on the 1A-SA Safety Bus which results in all non-emergency trips being bypassed; however this is incorrect since the bus differential trip is an emergency trip this signal will not be bypassed.

B. Incorrect. Plausible since both conditions result in the trip of the diesel during normal (Non-emergency) start conditions; however this is incorrect since the diesel started from an emergency start signal due to the undervoltage condition on the 1A-SA Safety Bus which results in all non-emergency trips being bypassed therefore only the bus differential trip is active.

C. Incorrect. Plausible since this condition will result in the trip of the diesel during normal (Non-emergency) start conditions; however this is incorrect since the diesel started from an emergency start signal due to the undervoltage condition on the 1A-SA Safety Bus which results in all non-emergency trips being bypassed therefore only the bus differential trip is active.

D. Correct.

Thursday, May 19, 2016 5:04:40 PM 128

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 062 AC Electrical Distribution / 6 062K3.02; Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following: ED/G (CFR: 41.7 / 45.6)

Importance Rating: 4.1 4.4 Technical

Reference:

APP-DGP-001 Window E-2, Rev 30, Page 34 OP-155, P&L #4, Rev 81, Page 7 References to be provided: None Learning Objective: Student Text DE, Objective 8.a Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 129

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 049/NEW/C/A//DCP ST/NONE/EARLY/063A1.01/SAT At 0800 Maintenance started a discharge test of the 125 VDC battery 1B-SB with an initial test load of 292 amps. The test will be terminated when any cell voltage reaches 2.14 Volts, which is expected to occur at 1200.

Subsequently at 0815 additional load was added to the battery, bringing total DC load to 365 amps.

Considering the additional load on the battery, which ONE of the following identifies the time that the battery will reach the termination criteria?

A. Prior to 1115 B. At 1115 C. After 1115, but before 1200 D. At 1200 Thursday, May 19, 2016 5:04:40 PM 130

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Raising the discharge rate on a battery reduces the battery capacity in a non-linear function such that raising the discharge rate by 1/4, reduces the capacity by more than 25%.

A. Correct.

B. Incorrect. Plausible since the discharge rate has been raised 25%, so it could appear that the capacity would be linearly reduced by 1/4 (4 x .25 = 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, 0815 + 3 hrs = 1115), but this is not correct because it is a non-linear relationship.

C. Incorrect. Plausible since the discharge rate has been raised, so it would appear that the capacity would be reduced based on the time the change in the battery load occurred (15 minutes after the test started) the candidate may misapply this and determine the time to discharge the battery is only reduced by the amount of time between the change in battery load, i.e. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> - 15 minutes = 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 45 minutes, 0800 + 3 hrs and 45 minutes = 1145), but this is not correct because it is a non-linear relationship.

D. Incorrect. Plausible if the candidate misapplies the concept of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> battery rating being affected by the change in the discharge rate and determines the change in the load does not affect the discharge rating and the test termination time does not change; but this is not correct because it is a non-linear relationship Thursday, May 19, 2016 5:04:40 PM 131

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 063 DC Electrical Distribution / 6 063A1.01; Ability to predict and/or monitor changes in parameters associated with operating the DC electrical system controls including: Battery capacity as it is affected by discharge rate (CFR: 41.5 / 45.5)

Importance Rating: 2.5 3.3 Technical

Reference:

Student Text DCP References to be provided: None Learning Objective: Student Text DCP, Objective 3 Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 132

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 050/BANK/FUNDAMENTAL//OP-156.01/NONE//063G2.1.20/

Which ONE of the following completes the statement below?

When placing the 1A-SA battery charger in service, the DC output breaker must be closed before energizing the charger from an AC source to .

A. prevent a possible undervoltage battery charger trip B. check for DC bus grounds before energizing the charger C. warmup the charger internals before placing it in service D. allow the charger output filter capacitors to charge from the battery Plausibility and Answer Analysis Reason answer is correct: When energizing a Battery Charger, the DC Output Breaker and feeder breaker to the associated Distribution Panel should be closed before energizing the charger from an AC source. This allows the output filters to become charged from the battery, and prevents drawing excessive current through the AC input rectifiers. To prevent a possible High Voltage trip, the output filters should be charged a minimum of 30 seconds before closing the AC Input Breaker.

A. Incorrect. Plausible since the Low DC Volt alarm will be recieved during this evolution; however this is incorrect because the alarm is expected to annunciate.

B. Incorrect. Plausible since the procedure has the candidate test the battery charger for grounds; however this is incorrect because the ground test is not performed until both the DC Output and AC input breakers are closed.

C. Incorrect. Plausible since the procedure has the candidate wait a minimum of 30 seconds prior to closing the AC input breaker and during this time the battery charger internal temperature is rising as a by product of I2 r losses; however this is incorrect because the reason for the 30 second minimum is to prevent an inadvertent High Voltage trip.

D. Correct.

Thursday, May 19, 2016 5:04:40 PM 133

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 063 DC Electrical Distribution / 6 063G2.1.20; Ability to interpret and execute procedure steps.

(CFR: 41.10 / 43.5 / 45.12)

Importance Rating: 4.6 4.6 Technical

Reference:

OP-156.01, P&L #1, Rev 37, Page 6 References to be provided: None Learning Objective: Student Text DCP, Objective 4.c Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 134

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 051/NEW/FUNDAMENTAL///NONE//064A1.04/

Given the following plant conditions:

- Post Maintenance Testing is in progress on the 1A-SA EDG. The 1A-SA EDG has been started in accordance with OP-155, Diesel Generator Emergency Power System The following problems are occurring on the EDG during the test:

- A Fuel Oil leak in the Day Tank room is lowering Day Tank level

- A mechanical failure within the EDG is causing crankcase pressure to rise

- A Jacket Water leak is lowering the Jacket Water Standpipe level

- Debris in the lube oil is clogging the lube oil filter Which ONE of the following would cause an automatic trip of the EDG?

A. Low Low Day Tank level B. High Crankcase Pressure C. Low Jacket Water level D. High Lube Oil Filter P Thursday, May 19, 2016 5:04:40 PM 135

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with APP-DGP-001 when an EDG is running during a NON-emergency start when EDG crankcase pressure reaches 2.5 psig the EDG will automaticaly trip.

A. Incorrect. Plausible since this is an alarm that will occur when the Fuel Oil Day Tank level decreases to 45.4% and without fuel the EDG cannot run. There isn't an automatic trip associated with low fuel and with the Low Low Day Tank Level (window E-4) alarm in there is still fuel in the day tank and the EDG will continue to run.

B. Correct.

C. Incorrect. Plausible since the lack of Jacket Water will lead to high lube oil temperatures. Lube Oil high temp (window A-1) will trip the EDG during a Non-emergency run when temperatures reach 195°F but the low level alarm for Jacket Water does not automaticaly cause an EDG trip.

Additionally, low level in the Jacket water system could lead to a low Jacket Water pressure which is also a Non-Emergency EDG trip. But, when the low level alarm comes in there is still adequate level in the Jacet Water Standpipe to support EDG cooling.

D. Incorrect. Plausible since a clogged filter could lead to low lube oil pressure which is a Non-emergency trip (window B-2 setpoint 31 psig) but high lube oil filter P is not a direct automatic Non-emergency trip.

Thursday, May 19, 2016 5:04:40 PM 136

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 064 Emergency Diesel Generator / 6 064A1.04; Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including:

Crankcase temperature and pressure (CFR: 41.5 / 45.5)

Importance Rating: 2.8 2.9 Technical

Reference:

APP-DGP-001, Rev. 31, Window C-1 "Trip High Press Crankcase" References to be provided: None Learning Objective: DE Objective 7 Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 137

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 052/NEW/FUNDAMENTAL//APP-ALB-010, OP-118/NONE//073A4.02/

Given the following plant conditions:

- The unit is operating at 100% power

- ALB-010-4-5, Rad Monitor System Trouble, alarms Which ONE of the following completes the statement below?

The (1) radiation monitoring panel used to confirm the alarm in accordance with APP-ALB-010 AND the display/control key push button will (2) to indicate the channel in alarm.

A. (1) RM-11 (2) blink B. (1) RM-11 (2) be solid C. (1) RM-23 (2) blink D. (1) RM-23 (2) be solid Thursday, May 19, 2016 5:04:40 PM 138

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Only safety related radiation monitors provide input in to ALB-010-4-5. All safety related radiation monitors have an associated RM-23 mounted in a safety cabinet in the Main Control Room. APP-ALB-010 directs the operator to confirm the alarm using the RM-23. If a channel high or alert signal is received, the appropriate HIGH or ALERT indicator goes on and the corresponding backlighted channel display/control key BLINKS. The display/control display key can be extinguished by depressing it, but the indicator remains illuminated until the condition clears.

A. Incorrect. the first part is plausible since the RM-11 provides monitoring capability for safety related radiation monitors; however this is incorrect since the RM-11 does not have the ability to change the status of a safety related radiation monitor, RM-23 is directed to be used by APP-ALB-010. The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since the RM-23 primary channel HIGH, ALERT and OPER lights stays solid to indicate the status of the radiation monitor; however this is not correct because the alarm status on the RM-11 will blink until the alarm is acknowledge on the RM-11 console.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:40 PM 139

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 073 Process Radiation Monitoring / 7 073A4.02; Ability to manually operate and/or monitor in the control room: Radiation monitoring system control panel (CFR: 41.7 / 45.5 to 45.8)

Importance Rating: 3.7 3.7 Technical

Reference:

APP-ALB-010, Window 4-5, Rev 32, Page 23 OP-118, Section 6.2, Rev 35, Page 24 References to be provided: None Learning Objective: Student Text RMS, Objective 2.h Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 140

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 053/BANK/C/A//ALB-002, AOP-022/NONE//076K4.01/

Given the following plant conditions:

- The unit is in Mode 4, performing a cooldown on RHR

- Both trains of CCW are in service

- NSW Pump 'A' is operating

- NSW Pump 'B' is in standby

- Both ESW Pumps are available, but are NOT running Subsequently the following occurs:

- NSW Pump 'A' experiences a sheared shaft Which ONE of the following completes the statement below?

ESW automatically aligns on a low (1) signal to cool (2) train(s) of CCW.

A. (1) flow (2) BOTH B. (1) flow (2) ONLY 'A' C. (1) pressure (2) BOTH D. (1) pressure (2) ONLY 'A' Thursday, May 19, 2016 5:04:40 PM 141

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: With the a shaft shear occurring on the running NSW pump the standby pump will not automatically start. The result will be lowering NSW and ESW header pressure due to the loss of NSW flow. Twenty (20) seconds after the associated ESW header pressure lowers to 53 psig the ESW pump will start and realign the header isolation valves in order to restore pressure. Because both trains of CCW are in service to support RHR operation once ESW pressure and flow are restored the system will supply both trains of CCW.

A. Incorrect. The first part is plausible since a shaft shear results in both low header flow and pressure. Additionally multiple cooling systems (Water cooled Air Handling Units) in the plant have the standby unit automatically start when a low system flow is detected; however this is incorrect because the ESW system automatic start signals are generated based on a low system pressure. The second part is correct since both CCW trains are in service.

B. Incorrect. The first part is plausible since a shaft shear results in both low header flow and pressure. Additionally multiple cooling systems (Water cooled Air Handling Units) in the plant have the standby unit automatically start when a low system flow is detected; however this is incorrect because the ESW system automatic start signals are generated based on a low system pressure. The second part is plausible since NSW 'A' is the service water pump supplying the system at the being of the event candidate misinterpret the system response and determine that only the

'A' header will be effected; however this is incorrect because NSW 'A' pump is supplying both service water headers and both will automatically isolate to supply the associated loads.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since NSW 'A' is the service water pump supplying the system at the being of the event candidate misinterpret the system response and determine that only the

'A' header will be effected; however this is incorrect because NSW 'A' pump is supplying both service water headers and both will automatically isolate to supply the associated loads.

Thursday, May 19, 2016 5:04:40 PM 142

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 076 Service Water / 4 076K4.01; Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Conditions initiating automatic closure of closed cooling water auxiliary building header supply and return valve(s)

(CFR: 41.7)

Importance Rating: 2.5 2.9 Technical

Reference:

AOP-022-BD, Rev 13, Page 2 APP-ALB-002, Window 6-1, 7-1, Rev 52, Page 27, 32 References to be provided: None Learning Objective: Student Text ESWS, Objective 7.b Question Origin: Bank Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 143

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 054/NEW/C/A//AOP-017/NONE//078K4.01/

Given the following plant conditions:

- The plant is operating at 100% power

- The Compressed Air System (CAS) Control Panel is in Sequence 1

- A loss of Auxiliary Bus 1D has occurred

- 'A' EDG is carrying Bus 1A-SA

- A leak is in progress on the Instrument Air system that is causing pressure to lower

- The crew enters AOP-017, Loss of Instrument Air Which ONE of the following completes the statement below?

'A' Air Compressor (1) AND (2) will be controlling the air compressor after it is restored.

A. (1) will start automatically (2) CAS Sequence 1 B. (1) will start automatically (2) the local pressure switch C. (1) must be locally reset to start (2) CAS Sequence 1 D. (1) must be locally reset to start (2) the local pressure switch Thursday, May 19, 2016 5:04:40 PM 144

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The lost of Aux Bus 1D will cause CAS to lose power. With the loss of power auto starts will be generated according to the local low pressure transducer switches. C will start at 101 psig, A at 96 psig, and B at 95 psig. 'A' A/C is supplied by emergency bus 1A1 off of the 1A-SA safety bus and has been lost until operator action is taken to restore, once the control room restores power to the 'A' A/C local operator action is require to reset the loss of power relay to restore 'A' A/C to operation.

A. Incorrect. The first part is plausible since the 'A' A/C is the CAS lead compressor in sequence 1; however this is incorrect because the loss of Auxiliary Bus 1D will disable CAS. With CAS disabled compressors will start off of their local low pressure transducer switches (see attachment 7 of AOP-017).

The second part is plausible since this would be correct if 1D had not lost power.

B.Incorrect. The first part is plausible see A(1). The second part is plausible if the candidate believes CAS Sequence 1 and Local Pressure transducer switch start orders are the same; however this is incorrect because the 'C' A/C local low pressure transducer switch will start it at 101 psig.

C. Incorrect. The first part is correct. The second part is plausible if the candidate believes that because the 'C' A/C is disconnected from CAS and it will always start prior to the 'B' A/C as in sequence 3; however this is incorrect because the CAS controller has lost power and is not able to control the

'B' A/C.

D. Correct.

Thursday, May 19, 2016 5:04:40 PM 145

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 078 Instrument Air / 8 078K4.01; Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Manual/automatic transfers of control (CFR: 41.7)

Importance Rating: 3.1 3.3 Technical

Reference:

AOP-017, Attachment 7, Rev 40, Page 57 OP-151.01, P&L 10, Rev 95, Page 9 References to be provided: None Learning Objective: Student Text ISA, Objective 7 Question Origin: New Comments: Ask Michael about the K/A...we don't have any HNP specific ties with design features and interlocks for securing service air due to a loss of cooling water and we are not able to create a valid HNP specific question.

Phonecon 4/15: HNP states that they are unable to generate a question with plausible distractors on the topic item of design features and interlocks for securing service air due to a loss of cooling water, so I committed to providing a new K/A. T2G1 K4 is at maximum, so not required to stay in K4. Attempted to stay in K4, there are 2 others with > 2.5: .01, .02. Assigned them 1 and 2, randomly chose: 2.

New K/A 078K4.01: Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following:

Manual/automatic transfers of control.

Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 146

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 055/NEW/FUNDAMENTAL//OP-113, CONT ST/NONE//103K4.04/

Which ONE of the following completes the statement below?

The Containment Personnel Airlock (PAL) doors (1) AND (2) interlock to prevent simultaneous operation of both doors in the automatic mode.

A. (1) operate on a common shaft (2) an electrical B. (1) operate on a common shaft (2) a mechanical C. (1) have separate operating stations (2) an electrical D. (1) have separate operating stations (2) a mechanical Thursday, May 19, 2016 5:04:40 PM 147

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Four control panels are provided for automatic operation of the airlock. Control Panels A and D have controls for BOTH doors; whereas, Panels B and C are for the adjacent door only. An electrical interlock is provided to prevent both doors from being opened simultaneously in the automatic mode of operation.

A. Incorrect. The first part is plausible since the Emergency Airlock (EAL) operates on a common shaft; however this is incorrect because the Personnel Airlock (PAL) operates electronically via hydraulic control stations. The second part is plausible since it is the correct type of interlock for the PAL B. Incorrect. The first part is plausible since the Emergency Airlock (EAL) operates on a common shaft; however this is incorrect because the Personnel Airlock (PAL) operates electronically via hydraulic control stations. The second part is plausible since it is the correct type of interlock for the EAL; however this is incorrect because the PAL interlock is an electronic interlock.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since it is the correct type of interlock for the EAL; however this is incorrect because the PAL interlock is an electronic interlock.

Thursday, May 19, 2016 5:04:40 PM 148

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 103 Containment / 5 103K4.04; Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following: Personnel access hatch and emergency access hatch (CFR: 41.7)

Importance Rating: 2.5 3.2 Technical

Reference:

OP-113, P&L #4, Rev 23, Page 6 References to be provided: None Learning Objective: Student Text CONT, Objective 4.b Question Origin: New Comments: None Tier/Group: T2/G1 Thursday, May 19, 2016 5:04:40 PM 149

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 056/NEW/C/A//RODCS STUDENT TEXT/NONE//001K5.10/

Given the following plant conditions:

- The unit is operating at 90% power after a reduction from 100% one hour ago The following indications exist for Power Range % Flux (AFD) and TI-408A, Tavg /

Tref mismatch meters.

Subsequently, the OATC manually withdraws Control Bank 'D' four steps for Tavg control.

Which ONE of the following completes the statement below describing the effects that the rod motion had on the indication for AFD and TI-408A?

The AFD indication became more (1) AND the Tavg / Tref mismatch indication became more (2) .

A. (1) positive (2) negative B. (1) positive (2) positive C. (1) negative (2) positive D. (1) negative (2) negative Thursday, May 19, 2016 5:04:40 PM 150

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The first part of the question deals with AFD. AFD is a measure of the difference of power produced on the upper part of the core compared to power produced in the lower part of the core. If AFD is at zero there is an equal amount of power produced in the upper and lower parts of the core. A positive value indicates that more power is being produced in the upper part of the core where a negative value indicates more power is being produced in the lower part of the core.

Currently AFD is slightly negative (~ negative 1). Control rod withdrawl at 90% will expose more of the upper part of the core to neutrons increasing the power produced in the upper portion of the core. The increase in power in the upper part of the core will cause AFD to become more positive.

The second part of the question deals with a mismatch indication between Tavg and Tref. The indication is a difference in the actual selected median Tavg minus the calculation of value of Tref which is based on Turbine first-stage pressure. As control rods are withdrawn positive reactivity is added to the core which causes the core to produce heat. The increase in heat will cause the measured median Tavg indication to rise which will cause the indication of mismatch of Tavg to Tref to become more positive.

A. Incorrect. The first part is correct. The second part is plausible if the candidate has a misconception about how the mismatch is calculated. If they have it backwards where Tref - Tavg is the readout then as Tavg is increased the mismatch will become more negative.

B. Correct.

C. Incorrect. Plausible since the unit had been ramped down from 100% to 90% one hour ago. The ramp caused a change in xenon conditions which could be thought to impact AFD causing AFD to become more negative and this negative effect could be greater than the positive effect that a rod withdrawl would have on AFD.

D. Incorrect. The first part is plausible (see C.1).

The second part is plausible (see A.2)

Thursday, May 19, 2016 5:04:40 PM 151

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 001 Control Rod Drive / 1 001K5.10; Knowledge of the following operational implications as they apply to the CRDS: Effect of rod motion on core power distribution and RCS temperatures (CFR: 41.5/45.7)

Importance Rating: 3.9 4.1 Technical

Reference:

Student Text RODCS References to be provided: None Learning Objective: RODCS Objective 4.d and Objective 10 Question Origin: New Comments: Per telecom discussion with Mike Donithan on 4-14-2016 the answers "negative" in the (1) and (2) parts of the question are acceptable since xenon effects could cause AFD changes. Negative could also be acceptable on the Tavg - Tref indication since a candiate could have a misconception and believe that the measurement is the difference between Tref and Tavg. Additionally, the mismatch indication does NOT read out in a temperature, it reads out degrees of mismatch. Also discussed LOD and determined LOD to be at least a 2.

Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:40 PM 152

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 057/BANK/C/A//OP-105, MST-I0050/NONE//015K6.02/

Given the following plant conditions:

- A Reactor startup was in progress when the Reactor trips on Source Range High Flux The following conditions existed at the time of the Reactor trip:

- The crew was verifying proper overlap and preparing to block the SR High Flux Trip

- IR Channel N-35 indicated 4 x 10-11 amps

- IR Channel N-36 indicated 7 x 10-11 amps Which ONE of the following could be the cause of the Reactor trip?

A. IR N-35 failed low causing the trip when P-6 cleared.

B. IR N-36 was overcompensated and caused the trip prior to P-6 being satisfied.

C. SR N-31 pulse height discrimination circuit failed causing an artificially high indication.

D. SR N-32 failed low causing the negative rate bistable to trip.

Plausibility and Answer Analysis Reason answer is correct: If the pulse height discriminator is set too low, higher readings will result.

A. Incorrect. Plausible since P-6 automatically unblocks the Source Range NI's when Intermediate Range NI's are below; however this is incorrect since both IR NI's are below the 1 x 10-10 amps therefore P-6 is not yet satisfied.

B. Incorrect. Plausible since IR NI-36 is reading higher than IR NI-35 the candidate may misinterpret the system response to overcompensation; however this is incorrect because an overcompensated NI will result in a lower indicated power level.

C. Correct.

D. Incorrect. Plausible since the RPS system generates a Reactor trip signal due to a negative rate trip from the Power Range NI's; however this is incorrect because the Source Range NI's do not generate a Reactor trip due to the change in flux rate.

Thursday, May 19, 2016 5:04:40 PM 153

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 015 Nuclear Instrumentation / 7 015K6.02; Knowledge of the effect of a loss or malfunction on the following will have on the NIS: Discriminator/compensation circuits (CFR: 41.7 / 45.7)

Importance Rating: 2.6 2.9 Technical

Reference:

MST-I0050, Rev 27, Page 39 References to be provided: None Learning Objective: Student Text NIS, Objective 8.e Question Origin: Bank Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:40 PM 154

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 058/NEW/FUNDAMENTAL//APP-ALB-027/NONE//016K3.10/

Given the following plant conditions:

- The unit is operating at 100% power

- S-2 1A-SA, Primary Shield Cooling Fan is in operation

- FS-01RP-7970S, S-2 Flow Switch fails low Subsequently ALB-027-5-5, Reactor Primary Shield Clg Fans S2 Low-Flow-O/L alarms Which ONE of the following completes the statements below?

In accordance with APP-ALB-027 the S-2 1A-SA control switch white light will be (1) AND S-2 1B-SB, Primary Shield Cooling Fan (2) .

A. (1) ON (2) will start automatically B. (1) ON (2) must be manually started C. (1) OFF (2) will start automatically D. (1) OFF (2) must be manually started Thursday, May 19, 2016 5:04:40 PM 155

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The flow sensor FS-01RP-7970 provides input into a flow switch to actuate the alarm ALB-027-5-5. While the low flow and thermal overload conditions actuate the ALB-027-5-5, only the thermal overload condition will energize the white light on the fan control switch. Because the Primary Shield Cooling fans do not automatically start the APP response is for the operator to manually start the standby Primary Shield Cooling fans.

A. Incorrect. The first part is plausible since the student has to recall the indications on the S-2 fan control switch and this alarm is associated with the white indication light; however this is incorrect because the white indication on the control switch indicates the presence of a thermal overload condition.

The second part is plausible since the containment cooling system fans E80 and 81's for CRDM cooling automatically start the standby fan if a low flow condition occurs; however this is incorrect because the S-2 and S-4 fans do not have an automatic start feature.

B. Incorrect. The first part is plausible see A(1). The second part is correct.

C. Incorrect. The first part is correct. The second part is plausible see A(2).

D. Correct.

Thursday, May 19, 2016 5:04:40 PM 156

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 016 Non-Nuclear Instrumentation / 7 016K3.10; Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: CCS (CFR: 41.7 / 45.6)

Importance Rating: 3.0 3.2 Technical

Reference:

APP-ALB-027-5-5, Rev 11, Page 14 References to be provided: None Learning Objective: Student Text CCS Objective 6 Question Origin: New Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:40 PM 157

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 059/BANK/FUNDAMENTAL//OP-168/NONE//027K2.01/

Which ONE of the following is the power supply for S-1A , Containment Airborne Radioactivity Removal (ARR) Fan?

A. MCC 1A21-SA B. 480V Bus 1A1 C. MCC 1D11 D. 480V Bus 1E2 Plausibility and Answer Analysis Reason answer is correct: In accordance with the electrical lineup checklist for the containment ventilation and relief system the power supply to the S-1A and S-1B are as follows:

1D11-2B Airborne Radioactive Removal Fan S-1 (1A-NNS) (both breakers) 1E11-1E Airborne Radioactive Removal Fan S-1 (1B-NNS) (both breakers)

A. Incorrect. Plausible since fans are powered from 480V MCCs and could be mistaken to be safety-related components.

B. Incorrect. Plausible since fans are powered from 480V nonsafety-related power supplies and could be mistaken to be 480V bus powered vice 480V MCC powered.

C. Correct.

D. Incorrect. Plausible since fans could be mistaken to be safety-related components and could be mistaken to be 480V bus powered vice 480V MCC powered.

Thursday, May 19, 2016 5:04:40 PM 158

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 027 Containment Iodine Removal / 5 027K2.01; Knowledge of bus power supplies to the following: Fans (CFR: 41.7)

Importance Rating: 3.1 3.4 Technical

Reference:

OP-168, Attachment 1, Rev 37, Page 27 References to be provided: None Learning Objective: Student Text CVS, Objective 5.a Question Origin: Bank Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:40 PM 159

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 060/NEW/C/A//EOP-E-1/NONE//028A1.01/

Given the following plant conditions:

- The Post Accident Hydrogen Monitoring System is in standby mode

- A LOCA develops inside Containment

- Containment hydrogen concentration is 0.5%

4 days later the following conditions exist:

- Safety Injection system is aligned for Cold Leg Recirculation

- Containment pressure has lowered to atmospheric

- Containment hydrogen concentration is 5%

Which ONE of the following completes the statement below in accordance with EOP-E-1, Loss Of Reactor Or Secondary Coolant?

The Hydrogen Monitoring System is required to be aligned (1) AND the Hydrogen Purge System (2) allowed to be in service.

A. (1) to continuous sample mode (2) is B. (1) to continuous sample mode (2) is NOT C. (1) for remote dilution panel operations (2) is D. (1) for remote dilution panel operations (2) is NOT Thursday, May 19, 2016 5:04:40 PM 160

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: During the performance of EOP-E-1 once the Low Head and High Head safety injection systems are in cold leg recirculation mode the hydrogen monitoring system is placed in continuous mode. The hydrogen concentration is monitored until the concentration rises to 4% or more at which time the plant staff evaluates additional recovery actions including the use of hydrogen purge in order to reduce hydrogen concentration. With Containment pressurized the Hydrogen Purge system is designed for operation during atmospheric conditions.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the hydrogen concentration is 4% the potential exist to create and explosive enviornment if oxygen from purge air is mixed in the containment volume; however this is incorrect because the hydrogen purge system is designed to dilute the hydrogen concentration and reduce it below 4% during atmospheric conditions.

C. Incorrect. The first part is plausible since a large break LOCA is in progress; however this is incorrect because remote dilution operation is only required during the performance of post accident sampling. The second part is plausible see B(2).

D. Incorrect. The first part is plausible see C(1). The second part is correct.

Thursday, May 19, 2016 5:04:40 PM 161

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 028 Hydrogen Recombiner and Purge Control / 5 028A1.01; Ability to predict and/or monitor changes in parameter (to prevent exceeding design limits) associated with operating the HRPS controls including: Hydrogen concentration (CFR: 41.5 / 45.5)

Importance Rating: 3.4 3.8 Technical

Reference:

EOP-E-1, Step 23, Rev 1, Page 26 References to be provided: None Learning Objective: EOP-LP-3.1, Objective 4.e Question Origin: New Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:40 PM 162

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 061/BANK/FUNDAMENTAL//FSAR 10.4.4/NONE//041G2.1.27/

Which ONE of the following completes the statement below?

The design purpose of the Steam Dump System includes (1) as well as (2) .

A. (1) removing residual heat from the primary following a Turbine trip (2) eliminating the need for rod movement during a secondary load rejection B. (1) removing residual heat from the primary following a Turbine trip (2) maintaining the plant in a Hot Standby conditions above the Point of Adding Heat C. (1) preventing overpressurization of the Steam Generators after an MSIV goes shut (2) eliminating the need for rod movement during a secondary load rejection D. (1) preventing overpressurization of the Steam Generators after an MSIV goes shut (2) maintaining the plant in a Hot Standby conditions above the Point of Adding Heat Thursday, May 19, 2016 5:04:40 PM 163

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with the FSAR, the Steam dump system is designed to perform the following 4 functions:

a) To permit the plant to accept sudden load rejections up to 50 percent external electrical load without incurring a reactor trip or lifting the main steam safety valves.

b) To remove stored energy and residual heat from the primary system following a turbine/reactor trip.

c) To maintain the plant in hot stand-by condition.

d) To permit manual controlled cooldown of the plant to the point where the Residual Heat Removal System can be placed in service.

A. Incorrect. The first part is correct. The second part is plausible since one of the steam dump system functions is to permit the plant to reject up to 50 percent external electrical load; however this is incorrect because the 50 percent load rejection is based on preventing a trip of the reactor or actuation of the MS Safety valves not prevent movement of the rod control system.

B. Correct.

C. Incorrect. The first part is plausible since the stream dumps relieve pressure from the main steam system; however this is incorrect since the steam dump system taps off the MS system down stream of the MSIVs they are not physically able to relieve SG pressure with the MSIVs shut. The second part is plausible since one of the steam dump system functions is to permit the plant to reject up to 50 percent external electrical load; however this is incorrect because the 50 percent load rejection is based on preventing a trip of the reactor or actuation of the MS Safety valves not prevent movement of the rod control system.

D. Incorrect. The first part is plausible since the stream dumps relieve pressure from the main steam system; however this is incorrect since the steam dump system taps off the MS system down stream of the MSIVs they are not physically able to relieve SG pressure with the MSIVs shut. The second part is correct.

Thursday, May 19, 2016 5:04:40 PM 164

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 041 Steam Dump/Turbine Bypass Control / 4 041G2.1.27; Knowledge of system purpose and/or function.

(CFR: 41.7)

Importance Rating: 3.9 4.0 Technical

Reference:

FSAR, Section 10.4.4, Amendment 58, Page 10.4.4-1 References to be provided: None Learning Objective: Student Text SDS, Objective 1 Question Origin: Bank Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:41 PM 165

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 062/BANK/FUNDAMENTAL//AOP-005/NONE//071K4.05/

In accordance with AOP-005, Radiation Monitoring System, which ONE of the following identifies the response to a HIGH radiation (RED) alarm on the WPB Stack 5 PIG monitor (REM-*1WV-3546)?

A. Alarm only (no auto actions)

B. WG Decay Tank E&F to Plant Vent valve (3WG-229) will shut C. WPB Control Room HVAC System swaps to recirculation mode D. Normal WPB supply fans (S-61's & 62's) and WPB nonfiltered exhaust fans (E-59 &

83) trip. Emergency filtration exhaust fans (E-45, 46, 47, & 49) start Plausibility and Answer Analysis Reason answer is correct: In accordance with AOP-005 and AOP-005-BD when rad monitor 1WV-3546 (WPB Stack 5) goes in high alarm 3WG-229, WG Decay Tanks E & F To Plant Vent Vlv should automatically shut.

A. Incorrect. Plausible since there are Plant Vent Stack rad monitors that can go into High Alarm but do not have any auto actions. Example:

RM-21AV-3509-1, Plant Vent Stack 1 WRGM can be in ALERT OR HIGH ALARM with NO auto actions. The action would be to manually start a RAB Emergency Exhaust Fan per OP-172, RAB HVAC B. Correct.

C. Incorrect. Plausible since AOP-005 guidance for a rad monitor in alarm for an affected area is to establish proper ventilation for that area (Attachment 3 step 8). Step 8.b has the MCR operators direct the Radwaste Control Room to align WPB Ventilation using OP-171, Waste Processing Building Heating and Air Conditioning. This alignment is in the recirculation mode but the alignment is not something that swaps automatically.

D. Incorrect. Plausible since RAB and FHB ventilation fans trip on High Radiation conditions and the emergency exhaust filtration automatically aligns with the start of the emergency fans. The WPB supply fans have auto trip features. But the trip signals to the supply fans are overcurrent, low temperature, low flow, or smoke detection. They do NOT trip on High Radiation from REM-1WV-3546.

Thursday, May 19, 2016 5:04:41 PM 166

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 071 Waste Gas Disposal / 9 071K4.05; Knowledge of design feature(s) and/or interlock(s) which provide for the following: Point of release (CFR: 41.7)

Importance Rating: 2.7 3.0 Technical

Reference:

AOP-005, Page 15, Rev. 30, AOP-005-BD, Page 3, Rev. 12 References to be provided: None Learning Objective: AOP-LP-3.9 Objective 4.a Question Origin: Bank Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:41 PM 167

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 063/BANK/C/A//AOP-005-BD/NONE//072K1.04/

Given the following plant conditions:

- Control Room Ventilation is in a normal lineup with 'A' Train fans in operation

- Power is lost to the 'B' Train North MCR Emergency Outside Air Intake (OAI)

Radiation Monitor, RM-3505B2SB Which ONE of the following completes the statements below?

A Control Room Isolation Signal (1) occurred.

The required action in accordance with Technical Specification 3.3.3.1, Radiation Monitoring For Plant Operations is to (2) .

A. (1) has (2) maintain the respective OAI isolated B. (1) has (2) place MCR Ventilation in recirculation with ALL OAIs isolated C. (1) has NOT (2) maintain the respective OAI isolated D. (1) has NOT (2) maintain MCR Ventilation in recirculation with ALL OAIs isolated Thursday, May 19, 2016 5:04:41 PM 168

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: A CRIS has occurred, only one channel is required to trip.

The Tech Spec Action is correct per T.S. table 3.3-6.

A. Correct.

B. Incorrect. A CRIS has occurred, only one channel is required to trip. The Tech Spec Action is incorrect per OWP-RM-01 and T.S. table 3.3-6 but this is plausible because this is the required action for NO MCR OAIs operable.

C. Incorrect. Plausible because for other systems the coincidence for Radiation monitors to cause isolation signals is different. Example: CNMT Vent Isolation is 2 of 4. The Tech Spec Action is correct per OWP-RM-01 and T.S. table 3.3-6.

D. Incorrect. Plausible since for other systems the coincidence for Radiation monitors to cause isolation signals is different. Example: CNMT Vent Isolation is 2 of 4. The Tech Spec Action is also incorrect per OWP-RM-01 and T.S.

table 3.3-6 but this is plausible because this is the required action for NO MCR OAIs operable.

Thursday, May 19, 2016 5:04:41 PM 169

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 072 Area Radiation Monitoring / 7 072K1.04; Knowledge of the physical connections and/or cause effect relationships between the ARM system and the following systems: Control room ventilation (CFR: 41.2 to 41.9 / 45.7 to 45.8)

Importance Rating: 3.3 3.5 Technical

Reference:

AOP-005-BD, page 3, Rev. 12, OWP-RM-01, page 5, Rev. 41 References to be provided: None Learning Objective: RMS Objective 6C Question Origin: Bank Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:41 PM 170

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 064/BANK/C/A//AOP-012, AOP-012-BD/NONE//075A2.02/

Given the following plant conditions:

- The plant is operating at 75% power Subsequently the following occurs:

- Circ Water Pump C breaker trips on an overcurrent condition

- Condenser vacuum is 5.5 inches Hg and degrading

- The crew enters AOP-012, Partial Loss of Condenser Vacuum Which ONE of the following directions is the MINIMUM required in accordance with AOP-012?

A. Verify that the Turbine has Tripped.

B. Verify that the 'C' Circ Water Pump discharge valve 1CW-12 shuts.

C. Verify that the 'C' Circ Water Pump Bearing/Seal Water Pump starts.

D. Dispatch the Outside AO to shut 'C' Circ Water Pump discharge valve 1CW-12.

Plausibility and Answer Analysis Reason answer is correct: In accordance with AOP-012, if a Circulating Water Pump trips, the associated pump discharge valve automatically shuts. The follow up actions of AOP-012 have the status of the automatic action verified using the pump and valve MCB indications.

A. Incorrect. Plausible since the vacuum has degraded past the <60% Turbine load trip setpoint of 5.0 inches Hg; however this is incorrect because the turbine load is 75% and the >60% Turbine load trip setpoint is 7.5 inches Hg.

B. Correct.

C. Incorrect. Plausible since the bearing/seal water pump on other large plant components, RCP's, Main Turbine, etc., automatically start or are continuously in-service to ensure the bearings are adequately cooled during coast down of the component; however this is incorrect because the required seal flow for the CWP's pump seals during coast down is provided by the pump discharge.

D. Incorrect. Plausible since the this is a contingency RNO action; however this is incorrect because this is only directed to be performed if the discharge valve fails to close automatically when the pump trips.

075 Circulating Water / 8 Thursday, May 19, 2016 5:04:41 PM 171

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 075A2.02; Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of circulating water pumps.

(CFR: 41.7 / 45.5 to 45.8)

Importance Rating: 3.2 3.2 Technical

Reference:

AOP-012, Step 10, Rev. 30, Page 5 AOP-012-BD, Section 1.0, Rev 19, Page 2 References to be provided: None Learning Objective: AOP-LP-3.12 Objective 3 Question Origin: Bank Comments: Ask Michael about the K/A...we don't have any ties with Circ Water / Service water at HNP other than they both take a suction from the Cooling Tower basin. We have had this K/A on a 2006 NRC RO exam but the question was written to the ability statement dealing with just the ESW pumps. There wasn't any mention in the question about Circ Water.

Phonecon 3/23: HNP states that neither example will work for their station, so I committed to providing a new K/A. No other 075A4 has IR >2.5. Noted from the Examination Outline cover sheet that RO T2G2 area A2 was not sampled, so randomly chose from there:

New K/A 075A2.02: Circulating Water System - Ability to predict the impacts of loss of circulating water pumps and use procedures to correct, control or mitigate consequences.

Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:41 PM 172

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 065/BANK/C/A//FPT/NONE//086A3.01/

Given the following plant conditions:

- Fire header pressure was 123 psig when a fire occurred on site

- Fire header pressure lowered to 88 psig Which ONE of the following completes the statements below?

The Motor Driven Fire Pump will be (1) .

The Diesel Driven Fire Pump will be (2) .

(Assume NO operator actions have been taken)

A. (1) Off (2) Off B. (1) Running (2) Off C. (1) Off (2) Running D. (1) Running (2) Running Thursday, May 19, 2016 5:04:41 PM 173

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: OP-149, Fire Protection P&L 2 states that the Jockey, Motor Driven, and Diesel Driven Fire Pumps are normally aligned for automatic operation are are subject to start at anytime.

Notes in OP-149 identify auto start setpoints based on lowering Fire Header pressure decreases. If fire header system pressure decreases to < 90 psig (local) or < 95 psig using ERFIS the Motor Driven Fire pump starts. The Diesel Driven Fire Pump auto starts when Fire Header pressure decreases to < 73 psig (local) or < 78 psig using ERFIS with an 8 second time delay. Since Fire Header Pressure has decreased to 88 psig the Motor Driven Fire Pump should have auto started and will be RUNNING but the Diesel Driven Fire Pump auto start setpoint has not been reached so it will be OFF.

A. Incorrect. Plausible if the candidate has a misconception that both the Motor and Diesel Driven Fire Pumps auto start signals are at < 78 psig.

B. Correct.

C. Incorrect. Plausible if the candidate has a misconception that the Diesel Driven Fire Pumps auto starts at 90 psig and the Motor Driven Fire Pump starts at 78 psig.

D. Incorrect. Plausible if the candidate has a misconception that both the Motor and Diesel Driven Fire Pumps auto start signals are at < 90 psig.

Thursday, May 19, 2016 5:04:41 PM 174

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 086 Fire Protection / 8 086A3.01; Ability to monitor automatic operation of the Fire Protection System including: Starting mechanisms of fire water pumps (CFR: 41.7 / 45.5)

Importance Rating: 2.9 3.3 Technical

Reference:

OP-149 Pages 17 and 18, Rev. 72 References to be provided: None Learning Objective: Fire Protection LP, Objective 9 Question Origin: Bank Comments: None Tier/Group: T2/G2 Thursday, May 19, 2016 5:04:41 PM 175

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 066/NEW/FUNDAMENTAL//AD-OP-ALL-1000/NONE//G2.1.14/

Which ONE of the following completes the statements below?

In accordance with AD-OP-ALL-1000, Conduct Of Operations, prior to closing a breaker with a MINIMUM load of (1) a plant announcement is required AND at MINIMUM the announcement will direct plant personnel to stand clear of the associated (2) is required before operating the breaker.

A. (1) 480V (2) electrical switchgear ONLY B. (1) 480V (2) piece of equipment AND electrical switchgear C. (1) 6.9kV (2) electrical switchgear ONLY D. (1) 6.9kV (2) piece of equipment AND electrical switchgear Thursday, May 19, 2016 5:04:41 PM 176

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with AD-OP-ALL-1000, Conduct Of Operation Starting or stopping plant equipment of 4 kv loads or greater that are operated from the Control Room.

(1) When starting or stopping plant equipment, Operations personnel will announce the planned activity with direction for plant personnel to stand clear of the equipment.

(2) For equipment with electrical switchgear 4 kv or greater, the announcement includes direction to stand clear of its associated electrical switchgear.

A. Incorrect. The first part is plausible since the AD-OP-ALL-1000 states Operations personnel make plant annoucements to alert personnel in the plant of changing plant conditions and the candidate may conclude that closing a 480V breaker is a change in plant conditions; however this is incorrect because AD-OP-ALL-1000 requires a plant announcement for the closing operation of a breaker that is rated greater than 4kV. The second part is plausible since breakers are located in various rooms throughout the plant, 286' RAB, 286' Turbine building, etc. the candidate may determine they are required to announce the room location of the switchgear as the minimum to be announced; however this is not correct since the minimum required in accordance with AD-OP-ALL-1000 is to announce stand clear of the switchgear not the entire switchgear room.

B. Incorrect. The first part is plausible see A(1). The second part is correct.

C. Incorrect. The first part is correct. The second part is plausible see A(2).

D. Correct.

Thursday, May 19, 2016 5:04:41 PM 177

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.1 Conduct of Operations G2.1.14; Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.

(CFR: 41.10 / 43.5 / 45.12)

Importance Rating: 3.1 3.1 Technical

Reference:

AD-OP-ALL-1000, Section 5.5.14, Step 2.b, Rev. 5, Page 38 References to be provided: None Learning Objective: PP-LP-2.0, Objective 10.g Question Origin: New Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 178

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 067/BANK/FUNDAMENTAL//OMM-016/NONE//G2.1.18/

Given the following conditions:

- ESOMS is NOT functioning

- The OATC is maintaining a manual narrative log The following log entries have been made:

- 0956 B-SB CSIP trip

- 1005 Started A-SA CSIP per AOP-018

- 1011 Established normal letdown Subsequently: At 1030, the OATC realizes he forgot to make a 0957 entry that letdown had been isolated.

Which ONE of the following identifies a proper entry in accordance with OMM-016, Operator Logs?

A. 0957 Isolated normal letdown B. L.E. 0957 Isolated normal letdown C. 1030 Isolated normal letdown (0957)

D. L.E. 1030 Isolated normal letdown (0957)

Plausibility and Answer Analysis Reason answer is correct: In accordance with OMM-016, Attachment 16, 1.e If it becomes necessary to make a log entry out of chronological order, the log entry MUST be noted with the actual time of the event and marked "L.E."

A. Incorrect. Plausible since ESOMS entry is time stamped when the entry is made so each entry must identify when the actual event occurred. If a late entry is made, the late entry box must be checked. The symbol indicates the difference in the current time and actual time the log entry should have been entered.

B. Correct.

C. Incorrect. Plausible since ESOMS entry time is automatically entered, but each entry must identify when the actual event occurred.

D. Incorrect. Plausible since this would place the entries in the correct order, but late entries are noted with the actual time of the event and marked with "L.E.".

Thursday, May 19, 2016 5:04:41 PM 179

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.1 Conduct of Operations G2.1.18; Ability to make accurate, clear, and concise logs, records, status boards, and reports.

(CFR: 41.10 / 45.12 / 45.13)

Importance Rating: 3.6 3.8 Technical

Reference:

OMM-016 Attachment 16 step 1.e, page 32, Rev. 73 References to be provided: None Learning Objective: Lesson Plan PP-LP-3.10 Objective 3.b Question Origin: Bank Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 180

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 068/BANK/FUNDAMENTAL//AOP-031/NONE//G2.1.8/

Given the following plant conditions:

- The plant is in Mode 6

- Fuel Handlers are waiting for an assembly to be placed in the upender on the Reactor side prior to transferring the assembly to the 'A' Fuel Pool

- A leak in the Spent Fuel Pool is causing Cavity and Spent Fuel Pool levels to lower

- The crew entered and are implementing AOP-031, Loss of Refueling Cavity Integrity Which ONE of the following completes the statements below concerning coordinated activities directed by the Operators in the control room for this event?

Once all fuel assemblies are safely stored, then direct the Fuel Handlers to (1) .

This will be followed by (2) .

A. (1) move the Fuel Transfer Cart to the Fuel Handling Building side (2) dispatching an Operator to shut 1PP-427, Fuel Transfer Tube Gate Valve B. (1) move the Fuel Transfer Cart to the Fuel Handling Building side (2) directing Maintenance to install and inflate Fuel Pool gates to the Unit 1&4 Transfer Canal C. (1) maintain the Fuel Transfer Cart on the Reactor side (2) dispatching an Operator to shut 1PP-427, Fuel Transfer Tube Gate Valve D. (1) maintain the Fuel Transfer Cart on the Reactor side (2) directing Maintenance to install and inflate Fuel Pool gates to the Unit 1&4 Transfer Canal Thursday, May 19, 2016 5:04:41 PM 181

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: AOP-031 has the crew verify that the Transfer Cart is parked in the FHB. The cart is currently on the Reactor side and therefore the MCR crew would direct the Fuel Handling crew to move the Transfer Cart to the Fuel Handling Building side. IF the Fuel Transfer Cart is on the Reactor side the Fuel Transfer Gate Valve 1PP-427 cannot be shut due to the design of the Fuel Transfer system at HNP.

A Correct.

B Incorrect. The first part is correct. The second part is plausible since another procedure will direct Maintenance to install and inflate the Fuel Pool gates; however this is incorrect because this is the direction that would be provided in AOP-041 for a loss of Fuel Pool Level but is not a direction from AOP-031.

C. Incorrect. The first part is plausible since the Fuel Transfer Cart can traverse from the REACTOR side or the FUEL HANDLING BUILDING side during refueling operations the candidate may misapply the required location in order to remove the cart emergency cable from the travel path of the Fuel Transfer Tube Gate Valve. However this is incorrect because the cart is required to be on the FUEL HANDLING BUILDING side of the transfer tube in order to remove the cart emergency cable from the travel path of the Fuel Transfer Tube Gate Valve. The second part of the answer is correct.

D. Incorrect. The first part is plausible see C(1). The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:41 PM 182

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.1 Conduct of Operations G2.1.8; Ability to coordinate personnel activities outside the control room.

(CFR: 41.10 / 45.5 / 45.12 / 45.13)

Importance Rating: 3.4 4.1 Technical

Reference:

AOP-031, Page 37, Rev. 18 References to be provided: None Learning Objective: AOP-LP-031, Objective 3 Question Origin: Bank Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 183

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 069/NEW/FUNDAMENTAL//AD-OP-ALL-1000/NONE//G2.2.21/

A safety related motor operated valve (MOV) is manually backseated using its handwheel due to packing leakage.

Which ONE of the following completes the statements below in accordance with AD-OP-ALL-1000, Conduct of Operations?

The MOV (1) prior to backseating AND (2) required to be manually removed from its backseat prior to performing post-maintenance stroke testing.

A. (1) remains energized (2) is B. (1) remains energized (2) is NOT C. (1) must be de-energized (2) is D. (1) must be de-energized (2) is NOT Thursday, May 19, 2016 5:04:41 PM 184

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: AD-OP-ALL-1000 states that an MOV that is manually backseated must be declared inoperable and prior to testing it must be manually removed from it backseat.

A. Incorrect. The first part is plausible since an MOV can be manually operated and remain operable provided it was not seated or backseated. The second part is correct.

B. Incorrect. The first part is plausible since an MOV can be manually operated and remain operable provided it was not seated or backseated. The second part is plausible since the candidate may not recognize that leaving the valve backseated during testing could affect its closing stroke time.

C. Correct.

D. Incorrect. The first part is correct. The second part of the distractor is plausible since the candidate may not recognize that leaving the valve backseated during testing could affect its closing stroke time.

Thursday, May 19, 2016 5:04:41 PM 185

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.2 Equipment Control G2.2.21; Knowledge of pre-and post-maintenance operability requirements.

(CFR: 41.10 / 43.2)

Importance Rating: 2.9 4.1 Technical

Reference:

AD-OP-ALL-1000, Section 5.6.7, Step 1, Rev. 5, Page 47, 48 References to be provided: None Learning Objective: PP-LP-2.0 Objective 9 Question Origin: New Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 186

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 070/MODIFIED/FUNDAMENTAL//TS TABLE 1.2/NONE//G2.2.35/

Given the following plant conditions:

- An RCS heatup is in progress

- RCS temperature is 358°F

- 1B-SB EDG is declared INOPERABLE due to failure of the shutdown relay Which ONE of the following identifies (1) the current plant OPERATIONAL MODE and (2) the Technical Specification requirements regarding Mode changes?

A. (1) Mode 3 (2) Change to Mode 2 may be performed provided the TS 3.8.1, AC Sources -

Operating, Action Statements for 1B-SB EDG inoperability are satisfied.

B. (1) Mode 3 (2) Change to Mode 2 may NOT performed.

C. (1) Mode 4 (2) Change to Mode 3 may be performed provided the TS 3.8.1, AC Sources -

Operating, Action Statements for 1B-SB EDG inoperability are satisfied.

D. (1) Mode 4 (2) Change to Mode 3 may NOT performed.

Plausibility and Answer Analysis Reason answer is correct: Mode 3 is defined in Tech Spec Table 1.2 as a plant condition where the average coolant temperature is > 350°F Tavg. With an RCS Tavg of 358°F Tech Spec Mode 3 applies. Tech Spec 3.8.1.1 states that in Modes 1-4 as a minimum two seperate and independent diesel generators must be operable. With the 1B-SB EDG declared inoperable and is required to be returned to operational within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the unit be placed in HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (already below HSB - Mode

3) and in Cold Shutdown (Mode 5) within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Since the EDG is inoperable Tech Spec 3.0.4 applies. A change of operational modes shall not be made when the conditions for the LCO are not met and the associated action requires a shutdown if they are not met within a specified time interval.

A. Incorrect. The first part is correct The second part allowing a Mode change is plausible because some Tech Specs indicate TS 3.0.4 is not applicable.

In this instance, 3.0.4 does apply, and even though action requirements are met, the LCO does not have an indefinite time requirement as defined by TS section 3.0.

B. Correct.

Thursday, May 19, 2016 5:04:41 PM 187

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal C. Incorrect. The first part is plausible since Hot Stutdown is Mode 4, which is 350°F >

Tavg > 200°F and the temperature of 358°F is >350°F. The second part allowing a Mode change is plausible because some Tech Specs indicate TS 3.0.4 is not applicable. In this instance, 3.0.4 does apply, and even though action requirements are met, the LCO does not have an indefinite time requirement as defined by TS section 3.0.

D. Incorrect. The first part is plausible since Hot Stutdown is Mode 4, which is 350°F >

Tavg > 200°F and the temperature of 358°F is >350°F. The second part is correct as the actions are correct since they represent the wording in Tech Spec 3.0.4 which does not allow for a Mode change with an inoperable component that has an LCO with an shutdown action time.

Original question:

Thursday, May 19, 2016 5:04:41 PM 188

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.2 Equipment Control G2.2.35; Ability to determine Technical Specification Mode of Operation.

(CFR: 41.7 / 41.10 / 43.2 / 45.13)

Importance Rating: 3.6 4.5 Technical

Reference:

TS Table 1.2, Operational Modes, TS 3.0.4, TS 3.8.1, References to be provided: None Learning Objective: TS LP-2.0/3.0/5.0/8.0 Objective 3.a and 4.a Question Origin: Modified - 2014 NRC RO 68 Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 189

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 071/NEW/C/A//APP-ALB-008/6-B-401 0210//G2.2.41/

Which ONE of the following completes the statements below?

The ASI pump will automatically start (1) after seal injection flow is lost to the RCPs.

Based on CAR-2166-B-401 Sheet 0210, with the 2-3/210A Control Relay contacts CLOSED the ASI pump will continue to run once the CS-210.1, ASI Pump is returned to the AUTO position from the START position because the (2) relay is energized.

(Reference provided)

A. (1) 2 minutes and 30 seconds (2) 49/MR B. (1) 2 minutes and 30 seconds (2) 42X C. (1) 2 minutes and 45 seconds (2) 49/MR D. (1) 2 minutes and 45 seconds (2) 42X Thursday, May 19, 2016 5:04:41 PM 190

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with APP-ALB-008 window 2-2 if the RCP Seal Injection flow remains below 4 gpm for 2 minutes and 30 seconds, the ASI SQUIB valves (1ASI-21 & 1ASI-22) are actuated and the ASI Pump starts 15 seconds later.

Review of CWD 2166 B-401 0210 determines in both the START and AUTO switch positions contact 5-6 is closed to energize the 42 and 42Xrelays when the 2-3/210A relay contacts 3-5 are closed. With the 2-3/210A relay contacts closed the power applied to the 42X relay will close the 42X 9-10 contacts maintaining the 42 relay energized until CS-210.1 is taken to STOP A. Incorrect. The first part is plausible since the ASI system squib valve will actuate at 2 minutes and 30 seconds after the RCP Seal injection flow is below 4 gpm; however this is incorrect because the ASI pump does not start until 2 minutes and 45 seconds have elasped. The second part is plausible since 49/MR contact is required to be closed in order for the 42 and 42X relays to be energize; however the CWD is shown in the de-energized state, therefore the 49/MR contact in the ASI Pump motor starting circuit is open when the 49/MR is energized.

B. Incorrect. The first part is plausible see A(1). The second part is correct.

C. Incorrect. The first part is correct. The second part is plausible see A(2).

D. Correct.

Thursday, May 19, 2016 5:04:41 PM 191

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.2 Equipment Control G2.2.41; Ability to obtain and interpret station electrical and mechanical drawings.

(CFR: 41.10 / 45.12 / 45.13)

Importance Rating: 3.5 3.9 Technical

Reference:

APP-ALB-008,Window 2-2, Rev 24, Page 9 6-B-401 0210 References to be provided: 6-B-401 0210 Learning Objective: Lesson Plan PSPR-3.1, Objective 5 Question Origin: New Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 192

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 072/BANK/FUNDAMENTAL//OP-106/NONE//G2.3.14/

Which ONE of the following is a condition that would result in excessive radiation exposure rates in the Containment (Keyway) Sump Area during a refueling outage?

A. Movement of irradiated fuel in the reactor vessel.

B. Withdrawal of the Incore Detectors from the core.

C. Draining the RCS to mid-loop prior to core off load.

D. A leak in the Auxiliary Building results in lowering Reactor Cavity level.

Plausibility and Answer Analysis Reason answer is correct: In accordance with OP-106, This procedure provides the radiological controls and approvals required per IER 11-41, Unplanned Personnel Exposures from Highly Radioactive In-Core Components. A radiation hazard exists in the containment during operation of the incore instrumentation system and when placing the detectors in storage.

A. Incorrect. Plausible because the candidate may not understand location differences inside containment. Shielding would be provided by the Reactor Cavity water level. In addition, the distance between a worker in the sump and a fuel assembly in transit would increase as the fuel assembly was removed from the core.

B. Correct.

C. Incorrect. Plausible because the candidate may not understand location differences inside containment. Shielding would be provided by the Reactor Cavity water level. In addition, the distance between a worker in the sump and a fuel assembly in transit would increase as the fuel assembly was removed from the core.

D. Incorrect. Plausible if the candidate did not understand how this leak affects RCS level and dose rates. A leak in the Aux Building could not cause RCS level to drop below the bottom of the RCS loops. In addition, lowering cavity level would affect dose rates in the region above the vessel flange.

Someone in the sump would be in a location below the vessel flange and would be unaffected.

Thursday, May 19, 2016 5:04:41 PM 193

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.3 Radiation Control G2.3.14; Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

(CFR: 41.12 / 43.4 / 45.10)

Importance Rating: 3.4 3.8 Technical

Reference:

OP-106, P&L #1, Rev 22, Page 4 References to be provided: None Learning Objective: EOP-LP-3.2 Objective 2.a Question Origin: Bank Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 194

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 073/PREVIOUS/FUNDAMENTAL///NONE//G2.3.15/

The following radiation monitors are in service:

- REM-3502A, Containment RCS Leak Detection

- REM-3502B, Containment Pre-Entry Purge Subsequently a Containment Isolation Phase 'A' actuation occurs.

Which ONE of the following describes the effect on these monitors?

REM-3502A REM-3502B A. remains in service remains in service B. remains in service is isolated C. is isolated remains in service D. is isolated is isolated Plausibility and Answer Analysis Reason answer is correct: A Phase 'A' Containment Isolation signal will shut sample panel valves 1SP-916, 1SP-16, 1SP-918 and 1SP-939. When these valves are shut Radiation monitor REM-3502A will not have any flow. REM-3502B does not have isolation valves that receive a Phase A signal and will remain unisolated when a Phase A signal is generated.

A. Incorrect. Plausible since 3502B remains in service on a Phase A, but 3502A isolates.

B. Incorrect. Plausible since one of the monitors isolates on a Phase A, but it is 3502A.

C. Correct.

D. Incorrect. Plausible since 3502A isolates on a Phase A, but 3502B remains in service.

Thursday, May 19, 2016 5:04:41 PM 195

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.3 Radiation Control G2.3.15; Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

(CFR: 41.12 / 43.4 / 45.9)

Importance Rating: 2.9 3.1 Technical

Reference:

PLP-116, Page 19, Rev. 56, Student Text Radiation Monitoring, Page 35 References to be provided: None Learning Objective: LP RMS Objective 6.b Question Origin: Previous 2014 NRC RO 70 radomly selected Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 196

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 074/BANK/C/A//EOP USERS GUIDE/NONE//G2.4.17/

While conducting a cooldown during the implemention of the EOPs due to a small break LOCA the following trends are observed:

Time RCS Temperature (°F) RCS Pressure (psig) 1400 435 462 1402 433 460 1404 431 458 1406 429 456 Subsequently the CRS has asked if "RCS pressure is stable or lowering".

Which ONE of the following identifies the correct response in accordance with the EOP User's Guide?

A. STABLE because RCS subcooling is rising B. STABLE because the crew is controlling the RCS pressure reduction C. LOWERING even though RCS subcooling is rising D. LOWERING because the crew cannot control the RCS pressure reduction Thursday, May 19, 2016 5:04:41 PM 197

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with the EOP User's Guide RCS Pressure should be considered STABLE because RCS subcooling is rising with an operator controlled cooldown in progress.

A. Correct.

B. Incorrect. It is correct for RCS Pressure to be considered STABLE because RCS subcooling is rising with an operator controlled cooldown in progress and it is plausible to consider the RCS pressure reduction to be controlled since an controlled RCS cooldown is in prorogress however this is incorrect because no attempt to control RCS Pressure is made during a RCS cooldown.

C. Incorrect. It is plausible for RCS Pressure to be considered LOWERING since the parameter values are lowering however this is incorrect because an operator controlled cooldown in progress and with subcooling rising RCS Pressure would be considered to be stable.

D. Incorrect. It is plausible for RCS Pressure to be considered LOWERING since the parameter values are lowering however this is incorrect because an operator controlled cooldown in progress and it is plausible to consider the RCS pressure reduction to be controlled since an controlled RCS cooldown is in prorogress however this is incorrect because no attempt to control RCS Pressure is made during a RCS cooldown.

Thursday, May 19, 2016 5:04:41 PM 198

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.4 Emergency Procedures / Plan G2.4.17; Knowledge of EOP terms and definitions.

(CFR: 41.10 / 45.13)

Importance Rating: 3.9 4.3 Technical

Reference:

EOP-Users Guide, Step 6.5, Rev. 46, Page 36 References to be provided: None Learning Objective: EOP-LP3.19, Objective 4.d Question Origin: Bank Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 199

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC RO 075/NEW/FUNDAMENTAL//APP-ALB-014/NONE//G2.4.50/

Given the following plant conditions:

- The unit is operating at 100% power

- ALB-014-1-1B, SG A NR LVL/SP HI/LO DEV alarms

- 'A' Steam Generator level is 51% and lowering

- 'B' and 'C' Steam Generator levels are 57% and stable

- FCV-478, SG A is throttling SHUT Based on these conditions which ONE of the following actions should be the MINIMUM performed in accordance with the APP-ALB-014?

A. take manual control of FCV-478, SG A and restore SG level B. perform the immediate actions of AOP-010, Feedwater Malfunctions C. wait for CRS permission before taking manual control of FCV-478, SG A D. wait for control bands and trip limits before taking manual control of FCV-478, SG A Plausibility and Answer Analysis Reason answer is correct: In accordace with the alarm procedure for SG A narrow range level going out of normal band of 52% to 62% the operator should check steam flow and feed flow for deviation and is the assoicated SG Flow Control Valve (FCV) is not sufficiently controlling level, switch the control to manal and restore level to normal (57% narrow range)

A. Correct.

B. Incorrect. Plausible since AOP-010 provides operator actions for Feedwater Regulator Valves that are NOT properly operating in AUTO by controling SG level with the Feedwater Regular Valves in MANUAL and maintaining SG levels between 52% and 62% but the expectations found in AD-OP-ALL-1000 for procedure compliance section 5.17.2 states: Written procedures are not necessary for situations where conditions exist which may require timely actions due to failure of automatic control systems or uncertain equipment status (e.g., taking manual control of hand/auto stations, position of selector switches, ect.) Operators at HNP are trained to take manual control of the FCV for a SG that has a manfunctioning controller rather than implementing AOP-010. The guidance for taking manual control is also found in OMM-001, Operations Adminstrative Requirements, Attachment 13. "During a transient situation, the RO/BOP may take manual control of a controller to prevent a transient or trip. The CRS will provide necessary guidance to the RO/BOP to stabilize the plant per this procedure for the affected controller. " Student text for SG Water Level Control states:

AOP-010 - A failure of the SGWLCS can result in inadequate feed flow.

This problem can be addressed by AOP-010. If it is due to an instrument Thursday, May 19, 2016 5:04:41 PM 200

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal or controller failure, the operator can take manual control per OMM-001 guidance.

C. Incorrect. Plausible since in accordance with AD-OP-ALL-1000 section 5.18.3, 2.b (1) Controllers/components and bypasses should be placed in manual control using the following guidance: Prior to taking manual actions, if known, the abnormal condition and the cause shall be communicated to the CRS. Although there is a communication from the individual that is placing the automatic controller to manual there isn't a need to wait for permission from the CRS to do so. In accordance with OMM-001, Operations Administrative Requirements section 5.1.11 2, the CRS shall direct control limits whenever controllers are placed in manual. The RO and BOP must understand these limits and implement the actions when the limits are reached. These actions must be taken without delay. This does not imply that CRS permission is required to initially place the controller in manual to prevent a transient or trip.

OMM-001, Attachment 13. "During a transient situation, the RO/BOP may take manual control of a controller to prevent a transient or trip. The CRS will provide necessary guidance to the RO/BOP to stabilize the plant per this procedure for the affected controller. "

D. Incorrect. Plausible since in accordance with AD-OP-ALL-1000 section 5.18.3, 2.c (1) Placing control systems in manual may require periodic manual adjustments in order to maintain desired plant conditions. The following shall be discussed with the CRS prior to placing any portion of a control system in manual: (b) control bands.

Although the control bands are discussed with the CRS it is NOT required to wait for the CRS to provide these control bands prior to placing the controller in manual. In accordace with OMM-001 the CRS shall provide control bands whenever controllers are placed in manual. This does not imply that CRS permission is required to initially place the controller in manual to prevent a transient or trip.

Thursday, May 19, 2016 5:04:41 PM 201

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.4 Emergency Procedures / Plan G2.4.50; Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

(CFR: 41.10 / 43.5 / 45.3)

Importance Rating: 4.2 4.0 Technical

Reference:

APP-ALB-014-1-1B, Page 4, Rev. 25 References to be provided: None Learning Objective: SGWLC Objective 5.d Question Origin: New Comments: None Tier/Group: T3 Thursday, May 19, 2016 5:04:41 PM 202

Appendix C Job Performance Measure Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301009H401 Task

Title:

Initiate Emergency Boration JPM No.: 2016 HNP NRC Exam Following a Reactor Trip (AOP-002) Simulator JPM CR a K/A

Reference:

APE024 AA1.17 RO 3.9 SRO 3.9 ALTERNATE PATH - YES Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: _________________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant

• The plant was at 100% power when the A MFW pump tripped

• The crew performed a manual Reactor Trip in accordance with AOP-010, Feedwater Malfunctions Initial Conditions:

• The crew completed the immediate actions of EOP E-0, Reactor Trip or Safety Injection and have transitioned to ES-0.1, Reactor Trip Response

• RCS temperature has been stabilized in accordance with ES-0.1 step 4

• Your position is the OATC Initiating Cue:

• You have the responsibility for the Foldout items in ES-0.1

• Continue ES-0.1 starting with step 5 Allow the candidate to use the procedures from the DO NOT READ TO THE Simulator for this JPM. You will need to have pre-made EXAMINEE: copies of ES-0.1 and AOP-002 ready for replacements after the JPM is complete.

2016 HNP NRC Exam Simulator JPM CR a Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Task Standard: Emergency Boration flow 30 GPM with Charging flow 30 GPM Required Materials: None General

References:

ES-0.1, Reactor Trip Response, Rev. 2 AOP-002, Emergency Boration, Rev. 24 Handout: Use simulator copy of EOP-ES-0.1 and AOP-002 Time Critical Task: No Validation Time: 8 minutes Critical Step Justification The operator must start one of the two BA pumps. A boric acid pump has to be running to deliver boric acid flow. The Boric Acid Tank provides the most preferred source of borated water for emergency Step 3 boration. If a Boric Acid Pump cannot be started, the RNO directs the operator forward to the step for establishing the flow path from the RWST.

After the emergency boration flow path cannot be established due to 1CS-278 failing to open the operator must establish an alternate boration valve lineup to establish a boration. Flow is established from Step 5 the BA Tank via a BAT pump through FCV-113A and FCV-113B to CSIP suction. This provides a method for the control room operator to use the preferred flow path if 1CS-278 cannot be used.

A flow rate of > 30 gpm ensures that the boron concentration and Step 6 required flow of the action statements of LCOs 3.1.1.1 and 3.1.1.2 are being met.

2016 HNP NRC Exam Simulator JPM CR a Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet 2016 NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-164

• Password noinstants

• Go To Run

• Turn volume down / range Source Range audio counts as needed to reduce distraction from source range audio counts

• (IF NEEDED) The 86 relays should roll when the simulator is placed in run. If not then run the APP file Roll 86 Gen or they can be manually overridden with override LOs XGAO018A GEN LOCKOUT G1A-TRIP COIL ON XGBO017A GEN LOCKOUT G1B-TRIP RELAY ON

• GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

The following setup information is how this exam IC was developed.

• Reset to IC-19

• Go to run

• Insert a malfunction to prevent two control rods from inserting on the trip

• CRF16a 220.0 4 (shutdown bank A Rod N-7)

• CRF16b 220.0 27 (control bank A Rod F-14)

• Insert IRF CVC161 ENGAGED to prevent 1CS-278 from opening

• Shut 1CS-8 (60 gpm letdown orifice)

• Reduce flow on FCV-121 to < 30 gpm (somewhere close to 20 gpm)

• Place a trip of the A MFW Pump on Trigger 1 (IMF CFW16A)

• Go to run, insert Trigger 1 then manually trip the Reactor

• Verify immediate action conditions are met

• Secure the TDAFW pump and adjust AFW flows to obtain SG levels > 25%

• Stabilize RCS temperature within the required range of ES-0.1, Step 4

• Acknowledge and reset annunciator alarms

• Freeze and snap to exam IC 2016 HNP NRC Exam Simulator JPM CR a Rev. 2

Appendix C Page 4 of 8 Form ES-C-1 PERFORMANCE INFORMATION Simulator Operator: When directed by the Lead Examiner go to Run.

START TIME:

ES-0.1, Step 5 Performance Step: 1 Check Feed System Status:

• RCS temperature - less than 564F

• Verify Feed Reg valves - SHUT

• Check feed flow to SGs - GREATER THAN 210 KPH Standard: Verifies RCS temperature indication less than 564F YES Verifies each Feed Reg Valve indicating SHUT YES Verifies feed flow to SGs greater than 210 KPH YES Comment:

ES-0.1, Step 6 Performance Step: 2 Check control rod status:

• Check DRPI - available

• Verify all control rods - fully inserted Standard:

• Determines DRPI available by indicating lights on AEP-1

• Determines two rods stuck fully out

• Takes RNO path to AOP-002 Evaluator Note: Applicant may go to AEP-1 to determine which rods are stuck.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR a Rev.2

Appendix C Page 5 of 8 Form ES-C-1 PERFORMANCE INFORMATION AOP-002, Step 1 Performance Step: 3 VERIFY a Boric Acid (BA) Pump RUNNING.

Standard: Starts one BA Pump Comment:

AOP-002, Step 2.a Alternate Path Begins Performance Step: 4 ESTABLISH boration flowpath using 1CS-278 as follows:

• OPEN 1CS-278, Emergency Boric Acid NO Addition Standard: Identifies 1CS-278 will not open.

Informs CRS 1CS-278 will NOT open.

Takes RNO path to Step 3 Evaluator Cue: Acknowledge 1CS-278 will not open.

Comment:

AOP-002, Steps 3.a,b Performance Step: 5 ESTABLISH boration flowpath using FCV-113A/B as follows:

OPEN the following valves

• 1CS-283, Boric Acid To Boric Acid Blender FCV-113A

• 1CS-156, Make Up To CSIP Suction FCV-113B VERIFY at least 30 gpm boric acid flow to CSIP suction on recorder panel or ERFIS point FCS0113A.

Standard: Locates MCB switches then turns switch to OPEN for

• 1CS-283

• 1CS-156 Verifies flow indicated on recorder panel or ERFIS Comment: Candidate may use recorder FI-113A vice ERFIS.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR a Rev.2

Appendix C Page 6 of 8 Form ES-C-1 PERFORMANCE INFORMATION AOP-002, Step 4 Performance Step: 6 VERIFY and MAINTAIN at least 30 gpm charging flow to RCS (FI-122A.1) until required boration is completed.

Standard: Verifies flow indicated on FI-122A.1 as < 30 gpm.

With FCV-121 in manual candidate increases demand to increase flow to 30 gpm flow to CSIP suction on FI-121A.1 Comment:

After the candidate has established and verified flow to CSIP suction on FI-121A.1 Evaluator Cue: Announce: I have the shift, END OF JPM Inform Simulator Operator to place the Simulator in Freeze.

Comment:

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR a Rev.2

Appendix C Page 7 of 8 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam Simulator JPM CR a Initiate Emergency Boration Following a Reactor Trip (ES-0.1 and AOP-002)

Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam Simulator JPM CR a Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• The plant was at 100% power when the A MFW pump tripped

• The crew performed a manual Reactor Trip in accordance with AOP-010, Feedwater Malfunctions Initial Conditions:

• The crew completed the immediate actions of EOP E-0, Reactor Trip or Safety Injection and have transitioned to ES-0.1, Reactor Trip Response

• RCS temperature has been stabilized in accordance with ES-0.1 step 4

• Your position is the OATC Initiating Cue:

• You have the responsibility for the Foldout items in ES-0.1

• Continue ES-0.1 starting with step 5 2016 HNP NRC Exam Simulator JPM CR a Rev. 2

Appendix C Page 1 of 14 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 004016H101 Task

Title:

Place Excess Letdown In Service JPM No.: 2016 HNP NRC Exam Simulator JPM CR b K/A

Reference:

004 A4.06 3.6 RO/ 3.1 SRO ALTERNATE PATH - NO Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The unit is operating at 100% power MOL

• Normal letdown needs to be secured for maintenance due to a Initial Conditions:

problem with PCV-145

• PCV-145 is in manual

• You are the OATC and have been directed by the CRS to establish Excess Letdown to the VCT per OP-107, Section 8.2.

Initiating Cue:

• Excess letdown has not been in service during this refueling cycle The candidates should be briefed outside of the Simulator prior to performing this JPM. Provide them with a copy of the procedure and inform them that ALL initial conditions are satisfied.

Evaluator Note:

This will allow them to review the Precautions and Limitations associated with OP-107 and have time for a task preview of the steps to accomplish establishing Excess Letdown. Expect that the candidates will take about 10-15 minutes to complete this review.

2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 2 of 14 Form ES-C-1 Worksheet Task Standard: Excess letdown is established with proper flow and temperature Required Materials: None General

References:

OP-107, Rev. 113 Handout: OP-107, Rev. 113, Prerequisites, P&Ls, and Section 8.2, Excess Letdown Heat Exchanger Operation Time Critical Task: No Validation Time: 15 minutes Critical Step Justification Step 12 Excess Letdown flow cannot be established if 1CS-466, EXCESS LETDOWN TO VCT/RCDT, is NOT positioned to the RCDT.

Step 13 Excess Letdown flow cannot be established if 1CS-461, EXCESS LETDOWN valve is NOT opened.

Step 14 Excess Letdown flow cannot be established if 1CS-460, EXCESS LETDOWN valve is NOT opened Step 17 Exceeding procedural parameters limits for outlet temperatures or pressure could damage the Excess Letdown Heat Exchanger OR the RCDT for this flow path.

Step 19 Exceeding procedural parameters limits for outlet temperatures or pressure could damage the Excess Letdown Heat Exchanger and for this flow path the excess pressure would go to the RCDT.

Exceeding procedural parameters limits for outlet temperatures or Step 22 pressure could damage the Excess Letdown Heat Exchanger and for this flow path the high pressure will lift the Letdown relief which discharges to the PRT.

2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 3 of 14 Form ES-C-1 Worksheet 2016 NRC Exam JPM b - SIMULATOR SETUP Simulator Operator

• Reset to IC-165

• Password noinstants

• Place RED Off Normal placard on PCV-145

• Go to RUN

• Silence and Acknowledge annunciators

• GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

The following setup information is how IC-165 was developed.

• Initial Simulator IC was IC-19

• GO to RUN

• Place PCV-145 in manual

• Silence Acknowledge and Reset Annunciators

• FREEZE and Snap these conditions to your exam IC 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 4 of 14 Form ES-C-1 PERFORMANCE INFORMATION Simulator Operator: When directed by the Lead Examiner go to Run.

START TIME:

Performance Step: 1 OBTAIN PROCEDURE Standard: Obtains OP-107 and reviews P & Ls and Section 8.2 for Excess Letdown Heat Exchanger Operation. Reviews and verifies initial conditions are satisfied.

Evaluator Cue: Initial conditions have been established Comment:

OP-107, Section 8.2, Note prior to step 1 Performance Step: 2 NOTE: Normally Excess Letdown will go to the VCT. However, if plant conditions warrant, the RCDT may be selected. When the Excess Letdown line has been flushed, the VCT position can then be re-selected.

NOTE: If Excess Letdown is to remain in service for sufficient time for dilution or boration to be necessary then VCT level should be lowered to accommodate the expected level increase before placing Excess Letdown in service NOTE: Placing Excess Letdown in service will result in increased dose rates in the Seal Water Heat Exchanger Room.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 5 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Caution prior to step 1 Performance Step: 3 Caution Excess Letdown operation during times of BTRS operation may result in damage to the RCP seals (due to increased contaminants and higher pH water). This should not prevent any AOP or EOP actions. The Responsible Engineer for RCP or CVCS may provide additional guidance if needed.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

OP-107, Section 8.2, Step 1 Performance Step: 4 INFORM Radwaste Control Room to monitor Seal Water Filter P while Excess Letdown is in service.

Standard: Contacts RW Control Room operator to monitor Seal Water Filter P while Excess Letdown is in service Acknowledge request to monitor Seal Water Filter P while Simulator Operator:

Excess Letdown is in service Comment:

OP-107, Section 8.2, Step 2.a Performance Step: 5 PLACE the excess letdown heat exchanger in operation as follows:

VERIFY 1CC-188, CCW TO EXCESS LETDOWN HEAT EXCHANGER, is open.

Standard: Locates MCB switch for 1CC-188, CCW TO EXCESS LETDOWN HEAT EXCHANGER, verifies it is open Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 6 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Step 2.b Performance Step: 6 VERIFY 1CC-202 SB, CCW FM EXCESS LTDN & RCDT HEAT EXCHANGERS, is open.

Standard: Locates MCB switch for 1CC-202 SB, CCW FM EXCESS LTDN

& RCDT HEAT EXCHANGERS, verifies it is open.

Comment:

OP-107, Section 8.2, Step 2.c Performance Step: 7 VERIFY 1CC-176, CCW TO EXCESS LTDN & RCDT HEAT EXCHANGERS, is open.

Standard: Locates MCB switch for 1CC-176, CCW TO EXCESS LTDN &

RCDT HEAT EXCHANGERS, verifies it is open.

Comment:

OP-107, Section 8.2, Note prior to step 3 Performance Step: 8 NOTE: Flushing the excess letdown line to the RCDT is required if the boron concentration in the excess letdown line from the RCS isolation valves to 1CS-466 is unknown or differs from RCS concentration. The volume of this line is 74 gallons. Two volumes (148 gallons) should be adequate to prevent unexpected reactivity changes in the RCS when flow is aligned to the VCT.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 7 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Caution prior to step 3 Performance Step: 9 Caution: 1CS-464, HC-137 EXCESS LTDN FLOW is rated for 1500 psid. Anytime that 1CS-464 is exposed to greater than 1500 psid, leakby should be expected.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

OP-107, Section 8.2, Step 3.a Performance Step: 10 IF excess letdown flow is to be aligned to the RCDT, THEN PERFORM the following:

NOTIFY Radwaste Control Room of expected RCDT level change.

Standard: Contacts RW Control Room and informs the operator to expect RCDT level change.

Simulator Operator: RW Operator acknowledges Comment:

OP-107, Section 8.2, Step 3.b Performance Step: 11 VERIFY 1CS-464, HC-137 EXCESS LTDN FLOW is shut (potentiometer to zero).

Standard: Operator verifies HC-137 EXCESS LTDN FLOW is shut (potentiometer to zero).

Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 8 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Step 3.c Performance Step: 12 PLACE 1CS-466, EXCESS LETDOWN TO VCT/RCDT, to the RCDT position.

Standard: Operator locates MCB switch and places 1CS-466, EXCESS LETDOWN TO VCT/RCDT, to the RCDT position.

Comment:

OP-107, Section 8.2, Step 4 Performance Step: 13 PLACE 1CS-461, EXCESS LETDOWN to OPEN.

Standard: Operator locates MCB switch and places 1CS-461, EXCESS LETDOWN valve to OPEN.

Comment:

OP-107, Section 8.2, Step 5 Performance Step: 14 PLACE 1CS-460, EXCESS LETDOWN to OPEN.

Standard: Operator locates switch and places 1CS-460, EXCESS LETDOWN valve to OPEN.

Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 9 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Note prior to Step 6 Performance Step: 15 NOTE: Seal Water Flow should be observed on FR-154A and FR-154B when adjusting 1CS-464, HC-137 EXCESS LTDN FLOW for the following reasons:

• RCP No 1 seal leakoff flow will be affected, and

• The possibility exists of lifting the 150 psi safety on the excess letdown/No. 1 seal return line.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

OP-107, Section 8.2, Caution prior to Step 6 Performance Step: 16 Do NOT exceed 174°F outlet temperature as indicated on TI-139.

Do NOT exceed 150 psig as indicated on PI-138.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 10 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Step 6 Performance Step: 17 ADJUST 1CS-464, HC-137 EXCESS LTDN FLOW as necessary to establish excess letdown flow, and not exceed the following parameters:

• 174°F outlet temperature as indicated on TI-139

• 150 psig as indicated on PI-138 Standard: Operator adjusts 1CS-464, HC-137 EXCESS LTDN FLOW to establish excess letdown flow while not exceeding 174°F outlet temperature as indicated on TI-139 and 150 psig as indicated on PI-138 until > 148 gallons have been flushed to the RCDT.

Examiner Cue: After adjustments to 1CS-464 have been made establishing (NOTE: This should be Excess letdown to RCDT cue the applicant:

enough time for the applicant to determine Time compression is being used; approximately 5 minutes that an adequate flush have elapsed since 1CS-464 has been opened.

has been completed.)

Comment:

OP-107, Section 8.2, Step 7.a Performance Step: 18 IF excess letdown flow is to be aligned to the VCT, THEN PERFORM the following:

VERIFY 1CS-464, HC-137 EXCESS LTDN FLOW is shut (potentiometer to zero).

Standard: Locates and verifies 1CS-464, HC-137 EXCESS LTDN FLOW is SHUT Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 11 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Step 7.b Performance Step: 19 PLACE 1CS-466, EXCESS LETDOWN TO VCT/RCDT, to the VCT position.

Standard: Locates MCB switch and places 1CS-466, EXCESS LETDOWN TO VCT/RCDT, to the VCT position.

Comment:

OP-107, Section 8.2, Note Prior to Step 7.c Performance Step: 20 NOTE: Seal Water Flow should be observed on FR-154A and FR-154B when adjusting 1CS-464, HC-137 EXCESS LTDN FLOW for the following reasons:

• RCP No 1 seal leakoff flow will be affected, and

• The possibility exists of lifting the 150 psi safety on the excess letdown/No. 1 seal return line.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

OP-107, Section 8.2, Caution Prior to Step 7.c Performance Step: 21 CAUTION :Do NOT exceed 174°F outlet temperature as indicated on TI-139.

CAUTION : Do NOT exceed 150 psig as indicated on PI-138.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 12 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-107, Section 8.2, Step 7.c Performance Step: 22 ADJUST 1CS-464, HC-137 EXCESS LTDN FLOW as necessary to establish excess letdown flow and not exceed the following parameters:

• 174°F outlet temperature as indicated on TI-139.

• 150 psig as indicated on PI-138.

Standard: Locates MCB control for 1CS-464, HC-137 EXCESS LTDN FLOW to establish flow and adjusts excess letdown flow while not exceeding 174°F outlet temperature as indicated on TI-139 or 150 psig as indicated on PI-138.

Comment:

NOTE: It may be necessary to ask the candidate if Excess Letdown has been placed in service IF they do not report to the CRS after Excess Letdown has clearly been established.

Examiner Cue: After Excess Letdown has been established and reported to the CRS then:

Announce: I have the shift, END OF JPM Inform Simulator Operator to place the Simulator in Freeze.

After the candidate has established Excess letdown within Terminating Cue: temperature and pressure limits and/or Excess letdown flow is > Charging flow JPM is complete.

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 13 of 14 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Simulator JPM b Establish Excess Letdown to the VCT OP-107, Section 8.2, Excess Letdown Heat Exchanger Operation Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam JPM CR b Rev. 1

Appendix C JPM CUE SHEET Form ES-C-1

• The unit is operating at 100% power MOL

• Normal letdown needs to be secured for maintenance due to a Initial Conditions:

problem with PCV-145

• PCV-145 is in manual

• You are the OATC and have been directed by the CRS to establish Excess Letdown to the VCT per OP-107, Section 8.2.

Initiating Cue:

• Excess letdown has not been in service during this refueling cycle 2016 NRC Exam JPM CR b Rev. 1

Appendix C Page 1 of 20 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301150H601 Task

Title:

Transfer To Hot Leg Recirculation JPM No.: 2016 HNP NRC Exam Simulator JPM CR c K/A

Reference:

EPE 011 EA1.11 RO 4.2 / SRO 4.2 ALTERNATE PATH - YES Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The plant was operating at 100% power and subsequently has experienced a Large Break LOCA.

Initial Conditions:

• The ESF equipment is operating and presently aligned per EOP-ES-1.3, Transfer to Cold Leg Recirculation.

• Your position is the OATC

• 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> have passed since the LOCA occurred.

• The CRS directs you to implement EOP-ES-1.4, Transfer Initiating Cue: Between Cold Leg and Hot Leg Recirculation and perform steps 1 - 5 to transfer to Hot Leg recirculation.

• The BOP will acknowledge annunciators not associated with your task.

2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 2 of 20 Form ES-C-1 Worksheet Task Standard: Transfer to hot leg recirculation is accomplished IAW EOP-ES-1.4.

Required Materials: None General

References:

EOP-ES-1.4, Rev. 0 Handout: Use simulator copy of EOP-ES-1.4 Time Critical Task: No Validation Time: 20 minutes Critical Step Justification Without shutting 1SI-340 and 1SI-341 RHR flow would continue to flow to the Step 6 cold legs Without opening 1SI-359 RHR flow would not be lined up to deliver flow to the Step 7 hot legs therefore hot leg recirculation would not occur.

Secures CSIP flow to prevent dead head conditions (no mini-flow protection)

Step 11 during flow path realignment Must reopen 1SI-52 and restart A CSIP to re-establish cold leg High Head Steps 14 and 15 Safety Injection flow in accordance with RNO action Secures CSIP flow to prevent dead head conditions (no mini-flow protection)

Step 18 during flow path realignment Without shutting 1SI-3 AND 1SI-4 CSIP flow would continue to flow to the RCS Step 19 cold legs Step 20 Must open 1SI-86 to establish flow path for B CSIP to hot legs Step 21 Restarts CSIP to re-establish High Head Safety Injection flow to the hot legs Step 24 Secures CSIP flow to deadheading CSIP during subsequent valve lineup Step 25 Must shut 1SI-52 to prevent CSIP flow to cold legs Step 26 and 27 Must open miniflow isolation valves to relieve pressure locking on 1SI-107 Step 28 Must restart the A CSIP to establish flow to the hot legs Step 29 Must open valve to establish a Hot leg Recirc flow path Step 30 Must secure mini-flow lineup to complete A CSIP flow path to the hot legs Step 31 Must secure mini-flow lineup to complete A CSIP flow path to the hot legs 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 3 of 20 Form ES-C-1 Worksheet 2016 NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-166

• Password noinstants

• (IF NEEDED) The 86 relays should roll when the simulator is placed in run. If not then run the APP file Roll 86 Gen or they can be manually overridden with override LOs XGAO018A GEN LOCKOUT G1A-TRIP COIL ON XGBO017A GEN LOCKOUT G1B-TRIP RELAY ON

• Put this file in the AMS folder: 2016NRCJPMcCAEP

• Go to RUN

• Silence and Acknowledge annunciators

• GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

The following setup information is how to develop the exam IC. Since the JPM recreates a pressure locked valve (1SI-107 will not open under pressure) a CAEP and several triggers need to be created.

• Initial Simulator IC was IC-19

• Tie a LB LOCA to Trigger 1 o imf RCS01A 0 100

• You will need to create a CAEP file and use this file to allow the candidate to take the switch for 1SI-107 to OPEN without the valve opening when tried on the first attempt. The CAEP file will delete the malfunction that has the valve wheel engaged and set to 0 (shut). I named the CAEP 2016NRCJPMcCAEP. In the future you can name it whatever you desire. Simulator exam security did not allow me to leave this CAEP in the CAEP folder so I saved it to my portable hard drive and loaded it for validation and the examination.

• CAEP 2016NRCJPMcCAEP o dmf sis083 o mrf sis082 (n 0 0) DISENGAGE

• Create 2 trigger files (these files are now saved on the Simulator so you will NOT have to recreate these trigger files - this is just what I did. Makes sure you have spaces before and after the equal signs and dashes for the name of the file.)

o CSIP_A_switch_to_START xa2i127 == 3 o 1SI-107OPEN

@xa1i102lJISlDI.value==4

• Now open the Event Trigger Summary (ET) to assign some files to triggers 1 + 2 o Trigger 1 - click assign file - scroll through the files and assign this file CSIP_A_switch_to_START o Click on link command and type this in the box trg= 2 APP! 2016NRCJPMcCAEP o Trigger 2 - click assign file - scroll through the files and assign this file 1SI-107OPEN 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 4 of 20 Form ES-C-1 Worksheet Continued on next page:

IC development for JPM CR c (continued)

• Insert a malfunction to maintain 1SI-107 shut (engage manual hand wheel and set valve position to 0 - shut) o irf sis082 (n 0 0) ENGAGE o irf sis083 (n 0 0) 0 00:00:00 0 o

GO to RUN

• Run trigger 1 (LB LOCA) this will cause a rapid Reactor Trip and SI

• Perform foldout actions of E-0 to trip all RCPs when RCS pressure is < 1400 psig and SI flow is > 200 gpm

• Perform additional actions of E-0 (including energizing 1A1 and 1B1 and adjusting AFW flows as necessary) then transition to E-1

• Energize power to CSIP cross connect valves o Run AMS file cvc\E-0 Att 3 CSIP suct & disc valve power

• Energize Accumulator discharge valves o Run AMS file cvc\path-1 Att. 6

• Continue with E-1 steps up to wait for 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

• Allow RWST level to decrease until < 23.4%

• With RWST level < 23.4% perform EOP ES-1.3 Cold Leg Recirc Lineup o Cold Leg Recirc Lineup established IAW ES-1.3

• Silence Acknowledge and Reset Annunciators

• FREEZE and Snap these conditions to your exam IC 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 5 of 20 Form ES-C-1 PERFORMANCE INFORMATION During the performance of this JPM a RED path condition may occur for RCS Integrity. This is expected since the RCS has depressurized from the LB LOCA event. Since RCS pressure is < 230 psig and RHR Hx header flow is > 1000 gpm the actions are to return to procedure and step in Evaluator Note: effect.

IF the candidate stops progressing with the procedure in effect cue them that EOP-FR-P.1, Response To Imminent Pressurized thermal Shock has already been addressed.

Continue with your actions.

Simulator Operator: When directed by the Lead Examiner go to Run.

START TIME:

OBTAIN PROCEDURE Performance Step: 1 Procedure EOP- ES-1.4 obtained Standard: Locates a copy of EOP-ES-1.4 and goes to Step 1 Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 6 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Note prior to Step 1 Performance Step: 2 NOTE: IF an RHR pump and/or CSIP has been secured to mitigate blockage of the associated recirculation sump, THEN it should NOT be restarted during implementation of this procedure. All valve alignments; however, should be performed.

NOTE: Monitoring for degraded recirculation sump performance and evaluation of potential mitigating actions is to continue during implementation of and following transition from this procedure Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

The recirc sump levels are >142 and there are no signs of degraded performance. IF the candidate takes more than a Evaluator Cue: minute to evaluate the recirculation sumps:

Cue: Another Operator will monitor the Containment Recirc Sumps for indications of Sump Blockage or Degradation Comment:

EOP-ES-1.4, Step 1 Performance Step: 3 Check Charging System Status:

a. Check charging line - ISOLATED Standard: Locates charging line flow indicator or isolation valve indicators and determines that flow through charging line is isolated.

Proceeds to step 2 of EOP-ES-1.4 Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 7 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Step 2 Performance Step: 4 Check SI Systems - ALIGNED FOR COLD LEG RECIRCULATION Standard: Determines that SI systems are aligned for Cold Leg Recirculation. (Also part of turnover)

Comment:

EOP-ES-1.4, Note prior to Step 3 Performance Step: 5 Steps 3, 4 AND 5 will transfer the SI system from cold leg recirculation to hot leg recirculation.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

EOP-ES-1.4, Step 3.a Performance Step: 6 Align RHR Pumps For Hot Leg Recirculation:

a. Shut low head SI to cold leg valves:

• 1SI-340

• 1SI-341 Standard: 1. Locates 1SI-340 control switch, verifies that the control power is ON and 1SI-340 is SHUT

2. Locates 1SI-341 control switch, energizes control power

3. Takes switch for 1SI-341 out of pull to lock then takes the switch to SHUT IF required to cue candidate due to confusion as to why 1SI-340 is powered and shut tell them:

Examiners NOTE:

1SI-340 has its power on and is shut from the Cold Leg Recirc alignment that has already been completed.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 8 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Step 3.b Performance Step: 7 b. Open low head SI to hot leg valve:

• 1SI-359 Standard: 1. Locates 1SI-359 control switch

2. Energizes control power

3. Takes switch for 1SI-359 out of pull to lock then takes switch to OPEN Comment:

EOP-ES-1.4, Caution prior to Step 4 Performance Step: 8 CAUTION: Simultaneous flow through two injection headers by one CSIP may cause pump run out (as indicated by oscillating discharge pressure).

Standard: Operator reads and placekeeps at any procedure caution (initials, checks or circle/slash)

Comment:

EOP-ES-1.4, Step 4.a Performance Step: 9 Align Train A CSIP For Hot Leg Recirculation:

a. Check Train A CSIP - RUNNING Standard: Identifies CSIP A is running Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 9 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Step 4.b Performance Step: 10 b. Check alternate high head SI to cold leg valve - OPEN

• 1SI-52 Standard: Identifies control power is on and 1SI-52 is OPEN Comment: This is the way the valve was aligned for Cold Leg Recirc.

EOP-ES-1.4, Step 4.c Performance Step: 11 c. Stop Train A CSIP.

Standard: Locates CSIP A control switch and takes the pump to STOP.

(May inform CRS that A CSIP will be stopped)

Comment:

IF CRS is informed that A CSIP will be stopped Evaluator Cue:

acknowledge the communication.

EOP-ES-1.4, Step 4.d Performance Step: 12 d. Shut alternate high head SI to cold leg valve:

• 1SI-52 Standard: Locates 1SI-52 control switch (control power is already ON) and takes the valve to SHUT Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 10 of 20 Form ES-C-1 PERFORMANCE INFORMATION NOTE: 1SI-107 is failed closed (handwheel is engaged and valve position set to 0% open). Do NOT delete the valve position at 0% open or engage the handwheel. IF the failure is deleted the valve will immediately go back to OPEN.

• Evaluator /

Simulator Operator Information: An AMS file with a conditional trigger has been developed to delete the failure of 1SI-107 (handwheel engaged and position at 0% open). The malfunction for 1SI-107 is modified when the A CSIP switch is taken to START which takes place AFTER the candidate attempts to open 1SI-107.

EOP-ES-1.4, Step 4.e Performance Step: 13 e. Open alternate high head SI to hot leg valve:

• 1SI-107 Standard: a. Locates 1SI-107 control switch

b. Energizes control power

c. Takes switch for 1SI-107 out of pull to lock then takes the switch to OPEN Identifies that 1SI-107 WILL NOT OPEN. Informs the CRS that 1SI-107 will not open and implements the RNO for step 4.e Comment:

Acknowledge any communication that 1SI-107 will not open Evaluator Cue: and communications with alternate path activities (1SI-52 and A CSIP restart).

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 11 of 20 Form ES-C-1 PERFORMANCE INFORMATION Alternate Path starts here RNO EOP-ES-1.4, Step 4.e Performance Step: 14 e. Perform the following:

1) Reopen alternate high head SI to cold leg valve:

1SI-52

2) Consult the plant operations staff to evaluate use of Attachment 1 to open the alternate high head SI to hot leg valve while continuing with this procedure.

Standard: Locates control switch for 1SI-52 and takes switch to OPEN.

Contacts plant operations staff (or informs CRS to contact plant operations staff) to evaluate use of Attachment 1 Comment:

CRS acknowledge need to contact plant operations staff OR Evaluator OR Simulator if plant staff is contacted then acknowledge request to Operator Cue:

evaluate use of Attachment 1.

EOP-ES-1.4, Step 4.f Performance Step: 15 f. Restart the Train A CSIP.

Standard: Locates CSIP A control switch and takes the pump to START.

(May inform the CRS that A CSIP will be restarted)

Comment:

IF CRS is informed that A CSIP will be started acknowledge Evaluator Cue:

the communication.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 12 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Step 5.a Performance Step: 16 Align Train B CSIP For Hot Leg Recirculation:

a. Check Train B CSIP - RUNNING Standard: Identifies CSIP B is running Comment:

EOP-ES-1.4, Step 5.b Performance Step: 17 b. Check any BIT outlet valve - OPEN

• 1SI-3

• 1SI-4 Standard: Identifies 1SI-3 and 1SI-4 are OPEN Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 13 of 20 Form ES-C-1 PERFORMANCE INFORMATION With the current plant configuration when the B CSIP is stopped the Alternate Seal Injection (ASI) timers will start due to RCP seal water flow decreasing to < 4 gpm. (Flow is provided by the B CSIP only at this time).

If flow is not restored by restarting the B CSIP within 2 minutes and 30 seconds the ASI squib valves will actuate and the ASI pump will start 15 seconds later. This will not Evaluator Note: change the hot leg injection line up butif the ASI pump starts the candidate may refer to AOP-018, the annunciator directions, or OP-185 to secure the ASI pump operation.

IF the candidate stops performing the actions of EOP- ES-1.4 to address the ASI system then Cue:

Another operator will address the ASI system response.

EOP-ES-1.4, Step 5.c Performance Step: 18 c. Stop Train B CSIP.

Standard: Locates CSIP B control switch and takes the pump to STOP.

(May inform CRS that B CSIP will be stopped)

Comment:

IF CRS is informed that B CSIP will be stopped Evaluator Cue:

acknowledge the communication.

EOP-ES-1.4, Step 5.d Performance Step: 19 d. Shut BIT outlet valves:

• 1SI-3

• 1SI-4 Standard: Locates 1SI-3 and 1SI-4 control switches and takes the valves to SHUT Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 14 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Step 5.e Performance Step: 20 e. Open high head SI to hot leg valve:

• 1SI-86 Standard: a. Locates 1SI-86 control switch

b. Energizes control power

c. Takes 1SI-86 switch out of pull to lock then takes switch to OPEN Comment:

EOP-ES-1.4, Step 5.f Performance Step: 21 f. Restart the Train B CSIP.

Standard: Locates CSIP B control switch and takes the pump to START.

(May inform CRS that B CSIP will be started)

NOT CRITICAL: Reports to CRS that A CSIP is still in Cold Leg Recirc due to 1SI-107 not opening and B CSIP is now in Hot Leg Recirc line up.

Comment:

Acknowledge any communications then cue below:

CRS to candidate:

Evaluator Cue: Plant operations staff has completed an evaluation of using Attachment 1 and the directions are to perform Attachment 1. Start with step 1 to attempt to open 1SI-107.

Report the position of 1SI-107 to CRS when complete.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 15 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Attachment 1 Note prior to step 1 Performance Step: 22 NOTE: This attachment provides guidance to open high head injection valves that are pressure locked. The effects of pressure locking are relieved by operating the associated CSIP.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

EOP-ES-1.4, Attachment 1 Caution prior to step 1 Performance Step: 23 CAUTION: CSIPs should NOT be operated unless the associated normal miniflow isolation valves are open.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

EOP-ES-1.4, Attachment 1, step 1 Performance Step: 24 To Open Alternate Hot Leg Valve (1SI-107) Perform The Following:

a. Stop Train A CSIP.

Standard: Locates MCB control switch for A CSIP and places switch to stop.

(May inform CRS that A CSIP will be stopped)

Comment:

IF CRS is informed that A CSIP will be stopped Evaluator Cue:

acknowledge the communication.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 16 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Attachment 1, step 1 Performance Step: 25 b. Shut alternate high head SI to cold leg valve:

1SI-52 Standard: Locates 1SI-52 control switch and takes the valve to SHUT Comment:

Ensure that the Simulator Operator has ran the AMS file that

• Evaluator: will delete the malfunction of 1SI-107 when the switch is take to OPEN.

EOP-ES-1.4, Attachment 1, step 1 Performance Step: 26 c. Open the common CSIP normal miniflow isolation valve:

1CS-214 Standard: Locates 1CS-214 control switch and takes switch to OPEN Comment:

EOP-ES-1.4, Attachment 1, step 1 Performance Step: 27 d. Open the associated Train A CSIP normal miniflow isolation valve:

1CS-182 (CSIP 1A-SA) 1CS-210 (CSIP 1C-SAB)

Standard: Locates 1CS-182 control switches and takes each switch to OPEN Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 17 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Attachment 1, step 1 Performance Step: 28 e. Restart the Train A CSIP.

Standard: Locates the control switch for A CSIP and STARTS the pump.

(May inform CRS that A CSIP will be started)

Comment:

IF CRS is informed that A CSIP will be started acknowledge Evaluator Cue:

the communication.

EOP-ES-1.4, Attachment 1, step 1 Performance Step: 29 f. Open alternate high head SI to hot leg valve:

1SI-107 Standard: Locates 1SI-107 control switch and takes it to OPEN (Valve opens)

Informs CRS that 1SI-107 has opened.

Comment:

Evaluator Cue: Acknowledge communication.

EOP-ES-1.4, Attachment 1, step 1 Performance Step: 30 g. Shut the common CSIP normal miniflow isolation valve:

1CS-214 Standard: Locates 1CS-214 and takes switch to SHUT Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 18 of 20 Form ES-C-1 PERFORMANCE INFORMATION EOP-ES-1.4, Step 6 Performance Step: 31 h. Shut the associated Train A CSIP normal miniflow isolation valve:

1CS-182 (CSIP 1A-SA) 1CS-210 (CSIP 1C-SAB)

Standard: Locates 1CS-182 control switches and takes each switch to SHUT Reports to CRS that 1SI-107 has opened and the A CSIP is in the Hot Leg Recirc lineup Comment:

CRS acknowledge report.

After the candidate has reported completion of Attachment 1 step 1: Evaluation on this JPM is complete.

Evaluator Cue:

I have the shift, END OF JPM Inform Simulator Operator to place the Simulator in Freeze.

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 19 of 20 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam Simulator JPM CR c Transfer to Hot Leg Recirculation IAW EOP-ES-1.4.

Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• The plant was operating at 100% power and subsequently has experienced a Large Break LOCA.

Initial Conditions:

• The ESF equipment is operating and presently aligned per EOP-ES-1.3, Transfer to Cold Leg Recirculation.

• Your position is the OATC

• 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> have passed since the LOCA occurred.

• The CRS directs you to implement EOP-ES-1.4, Transfer Initiating Cue: Between Cold Leg and Hot Leg Recirculation and perform steps 1 - 5 to transfer to Hot Leg recirculation.

• The BOP will acknowledge annunciators not associated with your task.

2016 HNP NRC Exam Simulator JPM CR c Rev. 2

Appendix C Page 1 of 12 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301068H401 Task

Title:

Perform a Max Rate Cooldown JPM No.: 2016 HNP NRC Exam for a SG Tube Rupture (E-3) Simulator JPM CR d K/A

Reference:

041 A4.08 RO 3.0 SRO 3.1 ALTERNATE PATH - YES Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The unit was operating at 100% power when a SG Tube Rupture developed in the A SG.

• The crew manually tripped the Reactor and actuated a manual Safety Injection.

• The crew has transitioned from EOP-E-0, Reactor Trip Or Safety Initial Conditions:

Injection to EOP-E-3, Steam Generator Tube Rupture.

• The TD AFW Pump has tripped and cause is being investigated

• The CRS has determined that there is at least one intact SG available for RCS cooldown.

• Your position is the BOP.

• ERFIS is NOT Available.

Initiating Cue:

• The CRS has directed you to continue progressing through EOP-E-3 starting at step 28.

2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 2 of 12 Form ES-C-1 Worksheet Task Standard: Determine the required core exit temperature based on the lowest ruptured SG pressure then cooldown the RCS to this target temperature utilizing the Steam Dumps and SG PORVs.

Required Materials: None General

References:

EOP-E-3, Steam Generator Tube Rupture, Rev. 1 Handout: Use simulator copy of EOP-E-3 Time Critical Task: No Validation Time: 15 minutes Critical Step Justification Critical to determine the required core exit temperature is 495°F by identifying that ERFIS is NOT available and the lowest ruptured SG pressure (C SG) is 900 to 1000 psig. Without the correct temperature the Step 2 cooldown will be either stopped too soon which will require additionally cooldown later OR later than required and delay making progress through E-3.

Critical to place Steam Dumps to Steam Pressure mode to allow manual Step 5 control of Steam dumps to establish a Max Rate RCS cooldown.

Critical to momentarily place Both Steam Dump Interlock Bypass Switches Step 7 To INTLK BYP if these switches are not taken to this position the Steam Dumps will not open.

Critical to dump steam from intact SGs to condenser at Maximum Rate Step 9 which will decrease RCS temperature which will enable the operator to depressurize the RCS to equal the ruptured SG pressure.

Critical to identify that the MAX RATE cooldown has stopped and the Steam Dumps have closed. Then reestablishes Max cooldown using SG PORVs.

Step 10 If this is not recognized then the RCS will not be able to be depressurized to the ruptured SG pressure to stop the tube leakage.

Critical to stop the RCS cooldown by securing the B and C SG PORVs prior to either Core Exit TCs reaching 480°F or 470°F if using Thot Step 12 indications for cooldown (due to readability of scale) to prevent RCS overcooling which would delay the RCS depressurization to ruptured SG pressure.

Setting the SG PORV controllers correctly prevents a delay in RCS Step 13 depressurization to ruptured SG pressure.

2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 3 of 12 Form ES-C-1 Worksheet 2016 HNP NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-167

• Password noinstants

• (IF NEEDED) The 86 relays should roll when the simulator is placed in run. If not then run the APP file Roll 86 Gen or they can be manually overridden with override LOs XGAO018A GEN LOCKOUT G1A-TRIP COIL ON XGBO017A GEN LOCKOUT G1B-TRIP RELAY ON

• Go to run

• Turn OFF ERFIS in SFC and power down the MCR EFRIS screens

• Set RVLIS screens: top screen to Vessel Level, bottom screen to T/Cs

• Silence and Acknowledge annunciators

• Place STAR placards on A SG AFW isolation valves 1AF-55, 1AF-137 and on A SG PORV controller GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

To recreate the IC setup for this JPM:

• Initial Simulator IC was IC-19

• Go to run

• Insert a 420 gpm SG Tube Leak on the A SG o imf sgn05a (n 00:00:00 00:00:00) 420 00:00:00 0

• Manually trip the Rx and insert a Safety Injection

• Perform actions of E-0 then transition to E-3

• Perform actions of E-3 up to step 28 (determine required core exit temperatures based on lowest ruptured SG pressure)

• Isolate A SG and shut A SG MSIV

• Allow the Simulator to run to obtain the following conditions (approximately) o RCS temp 546° o PZR Pressure ~1800 psig o A SG Pressure ~ 940 psig (To get the above PZR pressure we opened the PZR Spray a few times to lower RCS pressure down otherwise you will have problems with RCS pressure causing an SI signal and a distraction that you would not want during the JPM. Also opened the A SG PORV to bring down pressure to be at the low end of the 900 - 1000 range.

Leave the TDAFW Pump in operation until B and C SG pressures are < 900 # then trip the TDAFW pump)

• Trip the TDAFW Pp to prevent a cooldown using the TDAFW Pp o Imf cfw01c (n 00:00:00 00:00:00) true

• Develop Conditional Trigger (1) to cause the Steam Dumps to fail closed o Assign a conditional trigger (SDC_OPEN) when Steam Dumps output pushbutton is increased the Steam Dumps with a 5 second delay.

Dumps will close in ~ 30 seconds from when the output button is pushed.

o Imf mss07 (1 00:00:00 00:00:00) 3 0.0

• Silence Acknowledge and Reset Annunciators

• Freeze and Snap these conditions to your exam IC 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 4 of 12 Form ES-C-1 PERFORMANCE INFORMATION Simulator Operator: When directed by the Lead Examiner go to Run.

Performance Step: 1 OBTAIN PROCEDURE Standard: Obtains current copy of E-3, Steam Generator Tube Rupture (May review steps prior to step 28)

Prepares to proceed with E-3 step 28.

Each candidate may take a few minutes here to review the steps that have been previously completed in EOP-E-3.

Evaluator Note:

The JPM start time should begin when they have completed the review procedure review.

Comment:

START TIME:

EOP-E-3, Step 28 Performance Step: 2 Determine required core exit temperature based on lowest ruptured SG pressure:

Standard: Obtains SG A pressure from MCB pressure instruments and determines that the pressure is between 900 and 1000 psig Determines ERFIS is NOT available and the required core exit temperature is 495°F Comment: Critical to determine required temperature is 495°F.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 5 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-3, Step 29 Performance Step: 3 Condenser Available For Steam Dump:

Standard: Determines that the B and C MSIVs are OPEN Condenser is Available Steam Dump Control is Available Comment:

EOP-E-3, Step 30 Performance Step: 4 Place Steam Dump Pressure Controller In MANUAL AND Lower Output To 0%.

Standard: Locates Steam Dump Pressure Controller, places controller to manual and lowers output to 0%

Comment:

EOP-E-3, Step 31 Performance Step: 5 Place Steam Dump Mode Select Switch In STEAM PRESS.

Standard: Locates Steam Dump Mode Selector switch and turns switch to the right from Tavg to STEAM PRESS Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 6 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-3, Step 32 Performance Step: 6 Check RCS Temperature < 553°F (P-12)

Standard: Identifies that RCS temperature is < 553°F Locates BPLP-4-4 and determines that Low-Low Tavg Steam Dump Blocked (P-12) is ON Comment:

EOP-E-3, Step 33 Performance Step: 7 Momentarily Place Both Steam Dump Interlock Bypass Switches To INTLK BYP.

Standard: Locates the Steam Dump Interlock Bypass switches and turns BOTH switches to INTLK BYP Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 7 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-3, Step 34 Performance Step: 8 Check LOW-LOW STEAM DUMP (P-12) BYPASSED Status Light - ILLUMINATED Standard: Locates BPLB-5-4 and checks that the LOW-LOW STEAM DUMP (P-12) BYPASSED Status Light is ILLUMINATED Comment:

EOP-E-3, Step 35 Performance Step: 9 Dump Steam From Intact SGs To Condenser At Maximum Rate.

Standard: Raises Steam Dump controller to OPEN the Steam Dumps to start a Max Rate Cooldown.

Monitors RCS temperature during cooldown until Core Exit TCs (or B or C Loop WR Thot) are at the required target temperature.

Comment: When PK-464 (Steam Dump controller) raise pushbutton is depressed the steam dumps will fail shut in ~ 30 seconds.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 8 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-3, Step 35 continued - ALTERNATE PATH Begins Performance Step: 10 Identifies that the Cooldown has stopped and the Steam Dumps have closed. Determines SG PORV must be used to continue Max Rate Cooldown.

Standard: Informs the CRS that the Steam Dumps have closed (steam dump control is Unavailable) and restores the RCS Max Rate Cooldown by opening the SG PORVs in accordance with step 29 RNO actions.

RNO Step 29 now applies: Dump steam from intact SGs at maximum rate using any of the following:

(Listed in order of preference):

a) SG PORVs - Locates controls for ONLY the B and C SG PORVs and places controls to manual then 100% to fully open BOTH PORVs Verifies that the cooldown rate is again established and RCS temperature is lowering towards the core exit temperature of 495°F at the maximum rate.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 9 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-3, Step 36 Performance Step: 11 Core Exit TCs - LESS THAN REQUIRED TEMPERATURE Standard: Identifies that RVLIS or Thot temperatures indicate LESS THAN 495°F Informs the CRS that the core exit temperatures are less than the required temperature.

CRS acknowledges that the core exit temperatures are less Evaluator Cue:

than the required temperature.

Comment:

EOP-E-3, Step 37 Performance Step: 12 Stop RCS Cooldown Standard: Secures the cooldown by closing the B and C SG PORVs.

Comment: Critical to stop the RCS cooldown by securing the B and C SG PORVs

• IF using Core Exit TCs then prior to Core Exit TCs reaching 480°F

• IF using Thot indications then prior to Thot reaching 470°F (since these indications are not digital are reading in 10°F increments)

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 10 of 12 Form ES-C-1 PERFORMANCE INFORMATION In the next step the candidate can either set the PORV controller to AUTO with the setpoint range of 42% to 48.5%

OR Evaluator Note: Stabilize Core Exit Temperatures between 480°F - 495°F with the PORV controller in MANUAL. IF the manual PORV control option is chosen then the JPM can be ended when the Evaluator is satisfied with post-cooldown temperature control.

EOP-E-3, Step 38 Performance Step: 13 Maintain Core Exit TCs Less Than Required Temperature.

Standard: Sets the B and C SG PORV controllers to 48.5% to maintain Core Exit TCs < 495°F.

NOTE: A setting of 48.5% corresponds to 495°F (Tcold)

Acceptable PORV Controller setting range: 48.5% - 42%

(where 42% corresponds to 480°F Tcold)

OR Adjusts PORV controllers for B and C SG PORV Manually to control CET within a range of 470°F - 495°F.

CRS acknowledges that the core exit temperatures are being maintained as required. I have the shift.

Evaluator Cue:

End of JPM Direct Simulator Operator to place the Simulator in Freeze.

Comment:

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 11 of 12 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam Simulator JPM CR e Perform a Max Rate Cooldown for a SG Tube Rupture EOP-E-3, Steam Generator Tube Rupture Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C JPM CUE SHEET Form ES-C-1

• The unit was operating at 100% power when a SG Tube Rupture developed in the A SG.

• The crew manually tripped the Reactor and actuated a manual Safety Injection.

• The crew has transitioned from EOP-E-0, Reactor Trip Or Safety Initial Conditions:

Injection to EOP-E-3, Steam Generator Tube Rupture.

• The TD AFW Pp has tripped and cause is being investigated

• The CRS has determined that there is at least one intact SG available for RCS cooldown.

• You are the BOP.

• ERFIS is NOT Available.

Initiating Cue:

• The CRS has directed you to continue progressing through EOP-E-3 starting at step 28.

2016 HNP NRC Exam Simulator JPM CR d Rev. 1

Appendix C Page 1 of 14 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 005016H101 Task

Title:

Align the RHR System for ECCS JPM No.: 2016 HNP NRC Exam Mode Simulator JPM CR e K/A

Reference:

005 A4.01 RO 3.6 SRO 3.4 ALTERNATE PATH - NO Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The crew is performing a plant startup in accordance with GP-002, Normal Plant Heatup From Cold Solid To Hot Subcritical Mode 5 to Mode 3.

Initial Conditions:

• A RHR train is operating in cooldown alignment

• B RHR Train is in ECCS alignment

• The last satisfactory run of OST-1216 has the throttle position of 1CC-146 at 50 degrees open

• Your position is the BOP

• The CRS has directed you to align the A RHR System for ECCS Mode prior to RCS temperature exceeding 350°F (ERFIS)

Initiating Cue:

• An Extra RO will be monitoring the Secondary Systems

• The previous shift has started placing the A Train RHR System in ECCS Mode per OP-111 Section 7.2 and have turned over the lineup at step 19. Since B RHR pump is in ECCS mode steps 26-34 can be marked NA for the B Pump The students should be briefed outside of the Simulator prior to performing this JPM. Provide them with a copy of OP-111 with P&Ls, ALL initial conditions satisfied and Section 7.2 steps 1-18 completed.

Evaluator Note:

This will allow them to review the Precautions and Limitations associated with OP-111 and have time for a task preview of the steps. Expect that the candidates will take about 10 - 15 minutes to complete this review.

2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 2 of 14 Form ES-C-1 Worksheet Task Standard: A RHR pump is aligned for ECCS Mode of operation Required Materials: None General

References:

OP-111, Section 7.2.2 Rev. 59 Handout: OP-111, Rev. 59, Prerequisites, P&Ls, and Section 7.2.2, Restoring the RHR System to ECCS Mode place kept through step 18 Time Critical Task: No Validation Time: 15 minutes Critical Step Justification Step 2 Stopping the A RHR is required to prevent deadheading the pump during the ECCS valve alignment Isolates RHR from the CVCS Letdown system to prevent diversion of Step 5 ECCS flow from the RWST Step 12 Opening 1SI-322 SA aligns the A RHR suction to the RWST and is part of the ECCS valve alignment requirements Step 14 In order to operate 1SI-340 SA the control power must be turned on.

Without control power the valve would have to be manually operated.

Step 15 Opening 1SI-340 SA aligns the A RHR train discharge path to the RCS Cold legs and is required to be opened to allow flow while in the ECCS valve alignment.

Step 22 Opening 1RH-30 aligns the A RHR discharge path to the RCS Cold legs and is required to be opened to allow flow while in the ECCS valve alignment.

2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 3 of 14 Form ES-C-1 Worksheet 2016 NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-168

• Password noinstants

• Go to run

• Silence and Acknowledge annunciators NOTE: A EXTRA RO should be stationed near the FW controls to act as if they are monitoring the Secondary side of the plant. THERE SHOULDNT have to be any adjustments made during the performance of the JPM. The A RHR train should be placed in ECCS mode prior to the RCS exceeding 350°F.

GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

To recreate the IC setup for this JPM:

• Initial Simulator IC was IC-2

• The B RHR pump is currently in ECCS Mode and the A RHR is ready to be placed in ECCS Mode

• Using OP-111 Section 7.2.2 complete steps 1-18 for the A RHR Train with conditions satisfied to stop the A RHR pump in step 19

• Ensure that the IC is at a temperature as cool as possible prior to starting the alignment of the A Train RHR to ECCS mode to prevent exceeding 350° during the time it takes to align the system. You must also be well below 249°F (limit in GP-002 P&L 38) prior to shutting the RCS loop suction valves to the RHR system.

• Silence Acknowledge and Reset Annunciators

• Freeze and Snap these conditions to your exam IC 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 4 of 14 Form ES-C-1 PERFORMANCE INFORMATION Simulator Operator: When directed by the Lead Examiner go to Run.

START TIME:

OP-111 Section 7.2.2 Caution prior to Step 19 Performance Step: 1 Failure of equipment to secure in the following step will result in the associated EDG being inoperable. Tech Spec 3.8.1.1 is applicable until the breaker for the affected load is opened.

Standard: Operator reads and placekeeps at any procedure note or caution (initials, checks or circle/slash)

Comment:

OP-111 Section 7.2.2, Step 19 Performance Step: 2 STOP RHR PUMP A-SA.

Standard: Locates MCB control switch for the A RHR pump and takes switch to STOP the A RHR pump Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 5 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 20 Performance Step: 3 SHUT AND LOCK 1RH-26, RHR Header A To CVCS Letdown Isol Vlv.

Standard: Contacts Aux Operator to shut and lock 1RH-26, RHR Header A To CVCS Letdown Isol Vlv Acknowledge request to shut and lock shut and lock 1RH-26, Simulator Communicator:

RHR Header A To CVCS Letdown Isol Vlv Simulator Operator: Shut 1RH-26 and report to communicator that 1RH-26 is shut Report that 1RH-26, RHR Header A To CVCS Letdown Isol Vlv Simulator Communicator:

is shut and locked Comment:

OP-111 Section 7.2.2, Step 21 Performance Step: 4 ADJUST PK-145.1 setpoint as required by present plant conditions.

Standard: Adjusts PK-145.1 setpoint to maintain present pressure Comment:

OP-111 Section 7.2.2, Step 22 Performance Step: 5 SHUT 1CS-28, RHR LETDOWN HC-142.1 Standard: Locates switch and shuts 1CS-28, RHR Letdown HC-142.1 Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 6 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 23 Performance Step: 6 THROTTLE AND LOCK 1CC-146, RHR HX A Outlet Throttle Valve, to the position as determined by the last satisfactory run of OST-1216 or OST-1316 AND DOCUMENT per OPS-NGGC-1308.

Contacts Aux Operator and directs 1CC-146, RHR HX A Outlet Standard: Throttle Valve to be positioned to 50° open and then locked and starts documentation of 1CC-146 position.

Acknowledge request to throttle open 1CC-146 to 50° and then lock the valve in that position.

Simulator Communicator: (no simulator valve position changes need to be made)

Wait 10 - 20 seconds and report back that 1CC-146 has been throttled to 50° and locked.

Another Operator will complete the documentation of Evaluator Cue:

1CC-146 position per OPS-NGGC-1308.

Comment:

OP-111 Section 7.2.2, Step 24 Performance Step: 7 IF two Trains of CCW are in service, THEN REFER to OP-145, Securing the Second CCW Pump While Supplying Both RHR Heat Exchangers, PRIOR to performing the next step.

Standard: Identifies ONLY one Train of CCW is in service and N/As step for securing the second Train.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 7 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 25 Performance Step: 8 SHUT 1CC-147, CCW FROM RHR HEAT EXCHANGER A-SA.

Standard: Locates switch and shuts 1CC-147, CCW from RHR Heat Exchanger A-SA.

Comment:

OP-111 Section 7.2.2, Note prior to Step 26 Performance Step: 9 NOTE: Steps 7.2.2.26 through 7.2.2.34 place both RHR Trains in ECCS lineup at the same time. Both RHR Trains of components are to be aligned together unless only one RHR Train is in service. Steps for the RHR Train that is already in ECCS Mode may be marked N/A.

Standard: Operator reads and placekeeps at any procedure note (initials, checks or circle/slash)

Comment:

OP-111 Section 7.2.2, Caution prior to Step 26 Performance Step: 10 CAUTION: Both RHR Loops must be pressurized greater than 50 psig to prevent a water hammer if stagnant portions of the RHR System are still hot when 1SI-322 SA and 1SI-323 SB are opened.

Standard: Operator reads and placekeeps at any procedure caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 8 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 26 Performance Step: 11 VERIFY the following:

A RHR Loop is pressurized greater than 50 psig on PI-600A.

B RHR Loop is pressurized greater than 50 psig on PI-600B.

Standard: Locates PI-600A and PI-600B and identifies PI-600A is > 50 psig and PI-600B is < 50 psig (B RHR is in ECCS alignment)

Steps 26-34 can be marked NA for the B RHR pump.

Evaluator Cue: (If needed) B RHR pump is in ECCS mode steps 26-34 can be marked NA for the B Pump Comment: B RHR loop is in ECCS alignment pressure (< 50 psig on PI-600B)

OP-111 Section 7.2.2, Step 27 Performance Step: 12 OPEN 1SI-322 SA, RWST TO RHR PUMP A-SA.

Standard: Locates switch and OPENS 1SI-322 SA, RWST TO RHR PUMP A-SA Comment:

OP-111 Section 7.2.2, Step 28 Performance Step: 13 OPEN 1SI-323 SB, RWST TO RHR PUMP B-SB.

Standard: N/As step 28 per NOTE prior to step 26 (B Train is already in ECCS Mode)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 9 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 29.a Performance Step: 14 PERFORM the following for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG:

a. VERIFY 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG CONT PWR & VLV POS is in ON.

Standard: Locates control power and valve position switch for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG CONT PWR & VLV POS and takes switch out of PULL TO LOCK then places switch to ON Comment:

OP-111 Section 7.2.2, Step 29.b Performance Step: 15 PERFORM the following for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG:

b. OPEN 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG.

Standard: Locates switch and OPENS 1SI-340 SA, LOW HEAD SI TRAIN A COLD LEG Comment:

OP-111 Section 7.2.2, Step 29.c Performance Step: 16 PERFORM the following for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG:

c. PLACE 1SI-340 control switch to PULL TO LOCK Standard: Locates control power and valve position switch for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG CONT PWR & VLV POS and places switch to PULL TO LOCK Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 10 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 29.d Performance Step: 17 PERFORM the following for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG:

d. PLACE 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG CONT PWR & VLV POS, to OFF.

Standard: Locates control power and valve position switch for 1SI-340 SA, LOW HEAD SI TRAIN A TO COLD LEG CONT PWR & VLV POS and places switch to OFF Comment:

OP-111 Section 7.2.2, Step 30.a, b, c, d Performance Step: 18 PERFORM the following for 1SI-341 SB, LOW HEAD SI TRAIN B TO COLD LEG:

a. VERIFY 1SI-341 SB, LOW HEAD SI TRAIN B TO COLD LEG CONT PWR & VLV POS is in ON.

b. OPEN 1SI-341 SB, LOW HEAD SI TRAIN B TO COLD LEG.

c. PLACE 1SI-341 control switch to PULL TO LOCK

d. PLACE 1SI-341 SB, LOW HEAD SI TRAIN B TO COLD LEG CONT PWR & VLV POS, to OFF.

Standard: N/As step 30 per NOTE prior to step 26 (B Train is already in ECCS Mode)

Comment:

OP-111 Section 7.2.2, Step 31 Performance Step: 19 OPEN 1SI-326 SA, LOW HEAD SI TRAIN A TO HOT LEG CROSSOVER.

Standard: Locates switch and opens 1SI-326 SA, LOW HEAD SI TRAIN A TO HOT LEG CROSSOVER Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 11 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 32 Performance Step: 20 OPEN 1SI-327 SB, LOW HEAD SI TRAIN B TO HOT LEG CROSSOVER.

Standard: N/As step 32 per NOTE prior to step 26 (B Train is already in ECCS Mode)

Comment:

OP-111 Section 7.2.2, Step 33 Performance Step: 21 OPEN AND LOCK the following breakers:

1B21-SB-5B Supply Breaker To 1RH-1.

1B21-SB-11A Supply Breaker To 1RH-39.

1A21-SA-7B Supply Breaker To 1RH-2.

1A21-SA-8A Supply Breaker To 1RH-40.

Standard: Contacts Aux Operator and directs Operator to Open and then Lock the following breakers:

1B21-SB-5B Supply Breaker To 1RH-1.

1A21-SA-7B Supply Breaker To 1RH-2.

N/A for B Train which is already in ECCS Mode 1B21-SB-11A Supply Breaker To 1RH-39.

1A21-SA-8A Supply Breaker To 1RH-40.

Acknowledge request to open and then lock breakers:

1B21-SB-5B Supply Breaker To 1RH-1.

1A21-SA-7B Supply Breaker To 1RH-2.

Simulator Communicator:

After the Simulator Operator completes opening the breakers report back that the breakers are open and locked.

Open the breakers to:

1B21-SB-5B Supply Breaker To 1RH-1.

Simulator Operator: 1A21-SA-7B Supply Breaker To 1RH-2.

Inform the Simulator Communicator after completion.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 12 of 14 Form ES-C-1 PERFORMANCE INFORMATION OP-111 Section 7.2.2, Step 34 Performance Step: 22 Open the following:

a. 1RH-30, RHR HEAT XCHG A OUT FLOW CONT HC-603A1.

b. 1RH-66, RHR HEAT XCHG B OUT FLOW CONT HC-603B1.

Standard: Locates switch and opens:

a. 1RH-30, RHR HEAT XCHG A OUT FLOW CONT HC-603A1.

b. N/A B Train already in ECCS alignment Comment: After candidate completes opening 1RH-30 the RHR system alignment for A Train is in ECCS Mode.

Evaluation on this JPM is complete.

Evaluator Cue: END OF JPM Direct Simulator Operator to place the Simulator in Freeze.

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Page 13 of 14 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam JPM CR e Align the RHR System for ECCS Mode OP-111 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• The crew is performing a plant startup in accordance with GP-002, Normal Plant Heatup From Cold Solid To Hot Subcritical Mode 5 to Mode 3.

Initial Conditions:

• A RHR train is operating in cooldown alignment

• B RHR Train is in ECCS alignment

• The last satisfactory run of OST-1216 has the throttle position of 1CC-146 at 50 degrees open

• Your position is the BOP

• The CRS has directed you to align the A RHR System for ECCS Mode prior to RCS temperature exceeding 350°F (ERFIS)

Initiating Cue:

• An Extra RO will be monitoring the Secondary Systems

• The previous shift has started placing the A Train RHR System in ECCS Mode per OP-111 Section 7.2 and have turned over the lineup at step 19. Since B RHR pump is in ECCS mode steps 26-34 can be marked NA for the B Pump 2016 HNP NRC Exam Simulator JPM CR e Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301135H601 Task

Title:

Manually Align Containment Spray JPM No.: 2016 HNP NRC Exam Simulator JPM CR f K/A

Reference:

026 A4.01 (4.5/4.3) ALTERNATE PATH - YES Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• A plant event is in progress

• RCS pressure is lowering and Containment pressure is rising

• An automatic Reactor Trip and SI have been occurred Initial Conditions:

• EOP-E-0, Reactor Trip or Safety Injection Loss Of Reactor or Secondary Coolant is being implemented

• Immediate actions of E-0 have just been completed

• Your position is the OATC

• The CRS directs you to continue with EOP-E-0 starting at step 5 to Initiating Cue: stabilize the plant

• You are responsible for all of EOP-E-0 foldout items.

• The BOP will silence annunciators 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Task Standard: Containment Spray has been manually initiated and aligned for Containment Spray operation and ALL 3 RCPs are stopped.

Required Materials: None General

References:

EOP-E-0, Loss Of Reactor or Secondary Coolant Rev. 4 Handout: Simulator copy of EOP-E-0 Time Critical Task: No Validation Time: 10 minutes CRITICAL STEP JUSTIFICATION The RCP trip criteria should be monitored continuously while in EOP-E-0. Securing ALL 3 RCPs when EOP-E-0 foldout RCP Trip Step 1 Criteria is met is a required operator action to reduce the amount of RCS mass pumped out of the break.

Starting one train of Containment Spray system will reduce Containment pressure. The combination of a CT pump running and the Containment Step 11 Ventilation system prevents the Containment from exceeding the design pressure limit during a LBLOCA.

An automatic Phase B has occurred when Containment Pressure exceeded 10 psig. A Phase B actuation causes all CCW cooling to the Step 12 RCPs to automatically isolate. The RCPs cannot operate for prolonged periods when CCW flow to the RCP heat exchangers does not exist.

Securing the RCPs will prevent motor damage due to excessive heat.

2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet 2016 HNP NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-169

• Password noinstants

• (IF NEEDED) The 86 relays should roll when the simulator is placed in run. If not then run the APP file Roll 86 Gen or they can be manually overridden with override LOs XGAO018A GEN LOCKOUT G1A-TRIP COIL ON XGBO017A GEN LOCKOUT G1B-TRIP RELAY ON

• Go to RUN

• Silence and Acknowledge annunciators

• GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

The following setup was used to create this JPM exam snap Initial Simulator IC to IC-19, 100% power Go To Run Defeat automatic and manual Containment Spray actuation

• imf ZRPK505A FAIL_ASIS

• imf ZRPK505B FAIL_ASIS

• Initiate a LBLOCA o imf rcs01b 10%

• Stay in RUN until Containment Pressure exceeds 10 psig and is rising Go to FREEZE and save IC conditions 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 4 of 12 Form ES-C-1 PERFORMANCE INFORMATION This JPM will cause multiple annunciators to alarm based on the SI event that is occurring. There should be a person to silence annunciators to reduce the noise level initially Simulator Operator: with this JPM.

When directed by the Lead Examiner go to Run.

START TIME:

Foldout criteria will be met for RCP Trip Criteria and the candidate may identify that it is met quickly.

Examiners Note: It is critical to secure the RCPs prior to the direction of securing the RCPs at step 16 RNO since the candidate is tasked with ALL EOP-E-0 foldout items.

E-0, Step 5 Performance Step: 1 a. Review Foldout page

b. Evaluate EAL Matrix RCP TRIP CRITERIA IF both of the following occur, THEN stop all RCPs:

• SI flow- GREATER THAN 200 GPM

• RCS pressure - LESS THAN 1400 PSIG Standard: Checks SI flow and RCS pressure on MCB indicators, ERFIS, or Recorder Panel.

Identifies that both SI flow is > 200 gpm and RCS pressure is

< 1400 psig and foldout RCP Trip Criteria is met.

• Locates A, B, and C RCP control switches and STOPS all RCPs.

(May provide crew update that all RCPs have been secured and Containment is Adverse, pressure > 3 psig)

Continues to monitor remaining EOP-E-0 fold out items Comment:

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 5 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-0, Step 6 Performance Step: 2 Verify CSIPs - ALL RUNNING Standard:

• Locates and checks running indication for the A and B CSIPs (indicating lights, amps, flow, status lights, ect) and verifies that BOTH pumps are running Comment:

EOP-E-0, Step 7 Performance Step: 3 Verify RHR Pumps - ALL RUNNING Standard:

• Locates and checks running indication for the A and B RHR Pumps (indicating lights, amps, flow, status lights, ect) and verifies that BOTH pumps are running Comment:

EOP-E-0, Step 8 Performance Step: 4 Safety Injection flow - GREATER THAN 200 GPM Standard:

• Locates and checks Safety Injection flow rates and identifies that Safety Injection flow is > 200 gpm Comment:

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 6 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-0, Step 9 Performance Step: 5 RCS Pressure - LESS THAN 230 PSIG - NO Standard:

• Locates indications of RCS pressure and determines that pressure is > 230 psig

• GO TO STEP 12 Comment:

EOP-E-0, Step 12 Performance Step: 6 Main Steam Line Isolation - ACTUATED Standard:

• Identifies that a Main Steam Line Isolation should have actuated based on Containment pressure > 3 psig Comment:

EOP-E-0, Step 13 Performance Step: 7 Verify All MSIVs AND Bypass Valves - SHUT Standard:

• Locates indications for A, B, and C MSIV and MSIV Bypass valves and verifies that ALL are shut Comment:

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 7 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-0, Step 14 Performance Step: 8 Any SG pressure - 100 PSIG LOWER THAN PRESSURE IN TWO OTHER SGs Standard:

• Locates indications for A, B, and C SG pressures and identifies that pressures are approximately equal and NO SG is 100 psig lower than the other two SGs

• RNO step 14. GO TO Step 16 Comment:

EOP-E-0, Step 16 Performance Step: 9 (* Continuous Action Step)

Check CNMT Pressure - HAS REMAINED LESS THAN 10 PSIG Standard:

• Locates indications for Containment pressure and determines that Containment pressure has NOT remained < 10 psig

• RNO step 16. Perform the following:

a) Verify CNMT spray - ACTUATED b) Stop all RCPs Comment:

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 8 of 12 Form ES-C-1 PERFORMANCE INFORMATION EOP-E-0, Step 16 RNO Performance Step: 10 a) Verify CNMT spray - Actuated Standard:

• Locates and checks indication for CONTAINMENT SPRAY PUMPS A-SA and B-SB and/or checks ALB-001/4-1 CONTAINMENT SPRAY ACTUATION annunciator.

• Determines that neither CNMT spray pump is running

• Attempts to manually actuate Containment Spray by using the Containment Spray Actuation Switches.

• Checks indication to see if CNMT Spray Pumps have started. (NO neither pump has started)

Comment:

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 9 of 12 Form ES-C-1 PERFORMANCE INFORMATION Performance Step: 11 Alternate Path begins here Manually start Containment Spray System (Evaluator: The order is not specified in the procedure)

Standard: Manually start Containment Spray using the following steps:

• Takes control switch for CONTAINMENT SPRAY PUMP A-SA TO START

• Takes control switch for 1CT-50 to OPEN

• Takes control switch for 1CT-12 to OPEN

• Takes control switch for CONTAINMENT SPRAY PUMP B-SB to START

• Takes control switch for 1CT-88 to OPEN

• Takes control switch for 1CT-11 to OPEN Verifies that both Containment Spray pumps are operating correctly and have flow indication.

Provides report to CRS that both A and B CT systems have been manually actuated.

Evaluator Acknowledge report that both Containment Spray pumps Communication: have been manually started.

Comment: Critical to start either A or B CT Pump and open 1CT-50 and 1CT-12 or 1CT-88 and 1CT-11.

EOP-E-0, Step 16 RNO (continued)

Performance Step: 12 b) Stop all RCPs Standard: Locates control switches for RCP-A, RCP-B, and RCP-C and takes them to STOP or verifies that the RCPs have been secured earlier when E-0 foldout RCP Trip Criteria was met.

Informs the CRS that all RCPs have been secured.

Comment: This step also procedurally stops the RCPs but the action to secure the RCPs should have been addressed with EOP-E-0 foldout.

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 10 of 12 Form ES-C-1 PERFORMANCE INFORMATION CRS acknowledges that the RCPs have been secured.

Another Operator will continue with EOP-E-0.

Evaluator Cue:

Announce I have the shift END OF JPM Direct the Simulator Operator to go to Freeze.

Comment:

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Page 11 of 12 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 NRC Exam Simulator JPM f Manually Align Containment Spray In accordance with EOP-E-0 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 NRC Exam Simulator JPM f Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• A plant event is in progress

• RCS pressure is lowering and Containment pressure is rising

• An automatic Reactor Trip and SI have been occurred Initial Conditions:

• EOP E-0, Reactor Trip or Safety Injection Loss Of Reactor or Secondary Coolant is being implemented

• Immediate actions of E-0 have just been completed

• Your position is the OATC

• The CRS directs you to continue with EOP-E-0 starting at step 5 to stabilize the plant Initiating Cue:

• You are responsible for all of EOP-E-0 foldout items.

• The BOP will silence annunciators 2016 NRC Exam Simulator JPM f Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 301098H401 Task

Title:

Restoration of Offsite Power to JPM No.: 2016 HNP NRC Exam Emergency Buses (EOP ECA-0.0, Simulator JPM CR g Attachment 1)

K/A

Reference:

055 EA1.07 RO 4.3 SRO 4.5 ALTERNATE PATH - YES Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The unit was operating at 100% power

• A EDG is under clearance due to a failure that caused the Generator field to not flash during OST-1013 Subsequently:

• A failure of a transmission line on the Duke grid resulted in the Initial Conditions: cascading trip of several units which resulted in low grid frequency

• The HNP unit has experienced a loss of offsite power

• B EDG failed to start. The problem is being investigated

• The crew entered ECA-0.0, Loss Of All AC Power

• The load dispatcher has contacted HNP and informed the MCR that the grid is now stable

• Your position is the BOP

• The CRS has directed you to restore offsite power to a (one) AC emergency bus using ECA-0.0 Attachment 1.

Initiating Cue:

• The Load Dispatcher has given permission to restore offsite power to 6.9 KV buses and to reset any tripped Start Up XFMR lockout relays.

Examiner: Provide the candidate a copy of ECA-0.0 Attachment 1.

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet Task Standard: Bus 1B-SB energized from the SUT Required Materials: None General

References:

EOP-ECA-0.0, Attachment 1, Rev. 3 Handout: Use simulator copy of EOP-ECA-0.0 Time Critical Task: NO Validation Time: 15 Minutes CRITICAL STEP JUSTIFICATION Critical to close Start Up XFMR B(A) To Aux Bus E(D) Breaker Step 17.a (17.b) 121(101), without the breaker being closed power cannot be restored to Emergency Bus B-SB(A-SA).

Critical to close breaker 124(104) for Aux Bus E(D) To Emergency Bus Step 19.a (19.b) BSB(A-SA), without the breaker being closed power cannot be restored to Emergency Bus B-SB(A-SA).

Critical to close tie breaker 125(105) for Emergency Bus BSB(A-SA) To Step 22.a (22.b) Aux Bus E(D), without the breaker being closed power cannot be restored to Emergency Bus B-SB(A-SA).

Critical to close Emergency Bus BSB(A-SA) To XFMR B1SB(A1-SA)

Breaker B1 ASB(A1 A-SA) and Emergency Bus BSB(A-SA) To XFMR Step 24.a (24.b)

B3SB(A3-SA) Breaker B3 ASB(A3-SA) to supply power to safeguards emergency equipment.

Critical to close 6.9 KV Emergency Bus BSB(A-SA) To XFMR Step 25.a (25.b) B2SB(A2-SA) Breaker B2 ASB(A2-SA) to supply power to safeguards emergency equipment.

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet 2016 NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-170

• Password noinstants

• Hang clearance tags on 1A-EDG

• Protect Equipment IAW OMM-001 o Protected Train Equipment Tags on:

B-SB EDG Start Switch B-SB Fuel Oil Transfer Pump Switch Breaker 52-1, Breaker 52-2 and Breaker 52-3

• (IF NEEDED) The 86 relays should roll when the simulator is placed in run. If not then run the APP file Roll 86 Gen or they can be manually overridden with override LOs XGAO018A GEN LOCKOUT G1A-TRIP COIL ON XGBO017A GEN LOCKOUT G1B-TRIP RELAY ON

• Go to RUN

• Silence and Acknowledge annunciators GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

The following setup information is how this exam IC was developed o Reset to IC-19 o Place 1A-EDG under clearance

• IRF DSG005 (n 0 0) LOCAL

• IRF DSG006 (n 0 0) MAINTAIN o Fail Emergency Bus A-SA to Aux Bus D Tie Breaker 105 SA ASIS (this will not allow the breaker to be manually closed from the MCB switch)

• IOR XD1I066 (n 0 0) ASIS o Fail Emergency Bus B-SB to Aux Bus E Tie Breaker 125 SB ASIS (this will not allow the breaker to be manually closed from the MCB switch)

• IOR XD1I075 (n 0 0) ASIS o Fail 1B-SB EDG to start

• IMF DSG01 (n 0 0) B o Loss of Offsite Power (trigger 1)

• IMF EPS01 (1 0 0) W/O_DELAY 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Job Performance Measure Form ES-C-1 Worksheet JPM IC development - continued o Since Attachment 1 allows the operator to choose energizing either bus A or B, malfunctions were developed to fail breakers 105 and 125 ASIS. The JPM is written to have ONLY one of the buses energize due to an problem with the opposite train breaker (alternate path development). When the candidate first attempts to close either breaker 105 or breaker 125 the breaker they initially choose will NOT close. They will then have to restore power to the other bus. The conditional triggers will clear the other breakers failure when the first breaker switch is taken to the CLOSE position.

o Create 2 trigger files (note these files will NOT need to be recreated I have saved them to the Simulator trigger file this is just how I did it)

• Breaker104toclose

@xbbi073lJISlDI.value==3

• Breaker124toclose

@xbbi077lJISlDI.value==3 o Open ET (Event Trigger Summary) o On trigger 2 - click assign file then type in the following

• Breaker104toclose o Click - link command - then type in the following

• dor xd1i075 (n 0 0) ASIS o On trigger 3 - click assign file then type in the following

• Breaker124toclose o Click - link command - then type in the following

• dor xd1i066 (n 0 0) ASIS o Place the Simulator in Run - insert Trigger 1

• Isolate Letdown

• Adjust TDAFW flow to maintain AFW flow > 210 KPPH and NR levels between 25% to 50% (this may require adjusting TDAFW pump speed as necessary to raise flow)

• Place the EDG 1B-SB emergency stop switch to EMERG STOP o Delete the Loss of Offsite Power malfunction

• DMF EPS01 o FREEZE and SNAP these conditions to your exam IC 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 5 of 23 Form ES-C-1 PERFORMANCE INFORMATION Simulator Operator: When directed by the Lead Examiner go to Run.

START TIME:

EOP ECA-0.0 Step 9 Directs energizing AC Emergency Buses from Offsite Power using Attachment 1 The attachment allows flexibility of energizing Emergency Bus A with steps 2-8 or B with steps 9-15. There isnt a fault indicated on either bus so a candidate should NOT be suspecting that either bus has a fault. Butyou never know.

Since the JPM is going to be ran as an ALTERNATE PATH the candidate has the choice of attempting to re-energize either bus first. Either choice will yield a failure of energizing the first bus Evaluator Note: but will have a success path for energizing the second bus.

Since there could be a decision made by the candidate on which bus to restore first the JPM has a Part A (steps 2-8) and Part B steps 9-15).

IF the candidate starts with trying to energize the A bus (more than likely) use Part A of the JPM.

IF the candidate starts with trying to energize the B bus (least likely - maybe suspects a fault due to failure of EDG B to start) use Part B.

Common step for Part A and Part B EOP-ECA-0.0 Attachment 1 - RESTORATION OF OFFSITE POWER TO EMERGENCY BUSES Caution prior to step 1 Performance Step: 1 CAUTION Tripping of a Start Up XFMR lockout relay indicates a major fault on the XFMR. Reenergizing the XFMR may cause additional damage and should NOT be done without dispatcher's permission.

Standard: Operator reads and placekeeps at any procedure caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 6 of 23 Form ES-C-1 PERFORMANCE INFORMATION Common step for Part A and Part B EOP ECA-0.0, Attachment 1, Step 1.a, b Performance Step: 2 Obtain Load Dispatcher's permission prior to performing the following:

a. Restoring offsite power to 6.9 KV buses

b. Resetting any tripped Start Up XFMR lockout relays Standard: Information provided by CRS stated that the Load Dispatcher has provided permissions to restore offsite power to the 6.9 KV buses and reset any tripped Startup XFMR lockout releays Comment:

EOP ECA-0.0, Attachment 1, Caution prior to Step 2 Performance Step: 3 CAUTION An AC Bus should NOT be reenergized if it is suspected the bus may be faulted.

Standard: Operator reads and placekeeps at any procedure caution (initials, checks or circle/slash)

Comment:

EOP ECA-0.0, Attachment 1, Note prior to Step 2 Performance Step: 4 NOTE Steps 2 through 8 restore power to Bus ASA and Steps 9 through 15 restore power to Bus BSB.

Standard: Operator reads and placekeeps at any procedure note (initials, checks or circle/slash)

Comment:

Part A, Energizing the A Emergency Bus first starts on the next page Part B, Energizing the B Emergency Bus first starts on page 14

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 7 of 23 Form ES-C-1 PERFORMANCE INFORMATION PART A - Attempting restoration of power to the A Emergency Bus first PART B - Attempting restoration of power to the B Emergency Bus first (go to page 14)

EOP ECA-0.0, Attachment 1, Step 2.a Performance Step: 5.a On Start Up XFMR Protective Relay Panel 1A, verify offsite power to Start Up XFMR A:

a. Verify the Start Up XFMR 1A Lockout SU 1A Relay is reset.

Standard: Locates Startup XFMR 1A Lockout SU 1A Relay and verifies that the relay is reset. (Relay is reset)

Comment:

EOP ECA-0.0, Attachment 1, Step 2.b Performance Step: 6.a b. Verify closed any of the following switch yard tie breakers to energize Start Up XFMR A:

• Breaker 522

• Breaker 523 Standard: Locates tie breaker switches for Startup XFMR A

• Breaker 522 (Verifies already closed)

• Breaker 523 (Not required to be closed but maybe closed w/o consequences)

Comment:

EOP ECA-0.0, Attachment 1, Step 3.a Performance Step: 7.a Restore offsite power to 6.9 KV Aux Bus D:

a. Place Start Up XFMR To Aux Buses A & D Synchronizer control switch to BREAKER 101 position.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses A & D and places switch to Breaker 101 position Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 8 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 3.b Performance Step: 8.a b. Close Start Up XFMR A To Aux Bus D Breaker 101.

Standard: Locates switch for Start Up XFMR A To Aux Bus D Breaker 101 and places switch to CLOSE. (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 3.c Performance Step: 9.a c. Place Start Up XFMR To Aux Buses A & D Synchronizer control switch to OFF.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses A & D and places switch to OFF Comment:

EOP ECA-0.0, Attachment 1, Step 4 Performance Step: 10.a Verify Aux Bus D To Emergency Bus ASA Breaker 104 CLOSED Standard: Locates Aux Bus D to Emergency Bus A-SA Breaker 104 switch and takes switch to CLOSE (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 5 Performance Step: 11.a Verify Diesel Generator ASA Breaker 106 A SA OPEN Standard: Locates Diesel Generator ASA Breaker 106 A SA switch and verifies breaker is Open (GREEN LIGHT LIT)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 9 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 6.a Performance Step: 12.a Energize 6.9 KV Bus ASA:

a. Place Emergency Bus ASA To Aux Bus D Synchronizer control switch to SYNC.

Standard: Locates Synchronizer control switch for Emergency Bus ASA To Aux Bus D and places control to SYNC Comment:

EOP ECA-0.0, Attachment 1, Step 6.b Performance Step: 13.a b. Close Emergency Bus ASA To Aux Bus D Tie Breaker 105.

Standard: Locates switch for Emergency Bus ASA To Aux Bus D Tie Breaker 105 and takes switch to CLOSE.

(GREEN LIGHT STAYS LIT) - Reports to CRS that Emergency Bus ASA To Aux Bus D Tie Breaker 105 will not close (may dispatch AO to investigate)

Acknowledge report that Emergency Bus ASA To Aux Bus Evaluator Cue:

D Tie Breaker 105 will not close.

IF AO is dispatched: Acknowledge and repeat back Simulator Communicator:

communications to investigate breaker IF needed to get the candidate back on task: Ask for an Evaluator NOTE: estimation on when power will be restored to an Emergency Bus.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 10 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 9.a - Alternate Path Begins Restoration of power from the Start Up XFMR 1B to the B-SB Emergency Bus Performance Step: 14.a On Start Up XFMR Protective Relay Panel 1B, verify offsite power to Start Up XFMR B:

a. Verify the Start Up XFMR 1B Lockout SU 1B Relay is reset.

Standard: Locates Startup XFMR 1B Lockout SU 1B Relay and verifies that the relay is reset. (Relay is reset)

Comment:

EOP ECA-0.0, Attachment 1, Step 9.b Performance Step: 15.a b. Verify closed any of the following switch yard tie breakers to energize Start Up XFMR B:

• Breaker 5213

• Breaker 5214 Standard: Locates tie breaker switches for Startup XFMR A

• Breaker 5213 (Verifies already closed)

• Breaker 5214 (Not required to be closed but maybe closed w/o consequences)

Comment:

EOP ECA-0.0, Attachment 1, Step 10.a Performance Step: 16.a Restore offsite power to 6.9 KV Aux Bus E:

a. Place Start Up XFMR To Aux Buses B & E Synchronizer control switch to BREAKER 121 position.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses B & E and places switch to Breaker 121 position Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 11 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 10.b Performance Step: 17.a b. Close Start Up XFMR B To Aux Bus E Breaker 121.

Standard: Locates switch for Start Up XFMR B To Aux Bus E Breaker 121 and places switch to CLOSE. (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 10.c Performance Step: 18.a c. Place Start Up XFMR To Aux Buses B & E Synchronizer control switch to OFF.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses B & E and places switch to OFF Comment:

EOP ECA-0.0, Attachment 1, Step 11 Performance Step: 19.a Verify Aux Bus E To Emergency Bus BSB Breaker 124 CLOSED Standard: Locates Aux Bus D to Emergency Bus B-SB Breaker 124 switch and takes switch to CLOSE (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 12 Performance Step: 20.a Verify Diesel Generator BSB Breaker 126 B SB OPEN Standard: Locates Diesel Generator BSB Breaker 126 B SB switch and verifies breaker is Open (GREEN LIGHT LIT)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 12 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 13.a Performance Step: 21.a Energize 6.9 KV Bus BSB:

a. Place Emergency Bus BSB To Aux Bus E Synchronizer control switch to SYNC.

Standard: Locates Synchronizer control switch for Emergency Bus BSB To Aux Bus E and places control to SYNC Comment:

EOP ECA-0.0, Attachment 1, Step 13.b Performance Step: 22.a b. Close Emergency Bus BSB To Aux Bus E Tie Breaker 125.

Standard: Locates switch for Emergency Bus BSB To Aux Bus E Tie Breaker 125 and takes switch to CLOSE.

(RED LIGHT LIT Comment:

EOP ECA-0.0, Attachment 1, Step 13.c Performance Step: 23.a c. Place Emergency Bus BSB To Aux Bus E Synchronizer control switch to OFF.

Standard: Locates Synchronizer control switch for Emergency Bus BSB To Aux Bus E and places control to OFF Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 13 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 14 Performance Step: 24.a Close the following 6.9 KV breakers:

• Emergency Bus BSB To XFMR B1SB Breaker B1 ASB

• Emergency Bus BSB To XFMR B3SB Breaker B3 ASB Standard:

• Locates control switch for Emergency Bus BSB To XFMR B1SB Breaker B1 ASB and places control to CLOSE (RED LIGHT LIT)

• Locates control switch for Emergency Bus BSB To XFMR B3SB Breaker B3 ASB and places control to CLOSE (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 15 Performance Step: 25.a Verify 6.9 KV Emergency Bus BSB To XFMR B2SB Breaker B2 ASB CLOSED Standard:

• Locates control switch for 6.9 KV Emergency Bus BSB To XFMR B2SB Breaker B2 ASB and places control to CLOSE (RED LIGHT LIT)

Informs CRS that power is restored to Emergency Bus B-SB Acknowledge any reports:

After the 6.9 KV Emergency Bus BSB power is restored:

Evaluation on this JPM is complete.

Evaluator Cue:

I have the shift, END OF JPM Inform Simulator Operator to place the Simulator in Freeze.

Comment:

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 14 of 23 Form ES-C-1 PERFORMANCE INFORMATION PART B - Attempting restoration of power to the B Emergency Bus first Restoration of power from the Start Up XFMR 1B to the B-SB Emergency Bus Performance Step: 5.b On Start Up XFMR Protective Relay Panel 1B, verify offsite power to Start Up XFMR B:

a. Verify the Start Up XFMR 1B Lockout SU 1B Relay is reset.

Standard: Locates Startup XFMR 1B Lockout SU 1B Relay and verifies that the relay is reset. (Relay is reset)

Comment:

EOP ECA-0.0, Attachment 1, Step 9.b Performance Step: 6.b b. Verify closed any of the following switch yard tie breakers to energize Start Up XFMR B:

• Breaker 5213

• Breaker 5214 Standard: Locates tie breaker switches for Startup XFMR A

• Breaker 5213 (Verifies already closed)

• Breaker 5214 (Not required to be closed but maybe closed w/o consequences)

Comment:

EOP ECA-0.0, Attachment 1, Step 10.a Performance Step: 7.b Restore offsite power to 6.9 KV Aux Bus E:

d. Place Start Up XFMR To Aux Buses B & E Synchronizer control switch to BREAKER 121 position.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses B & E and places switch to Breaker 121 position Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 15 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 10.b Performance Step: 8.b e. Close Start Up XFMR B To Aux Bus E Breaker 121.

Standard: Locates switch for Start Up XFMR B To Aux Bus E Breaker 121 and places switch to CLOSE. (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 10.c Performance Step: 9.b f. Place Start Up XFMR To Aux Buses B & E Synchronizer control switch to OFF.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses B & E and places switch to OFF Comment:

EOP ECA-0.0, Attachment 1, Step 11 Performance Step: 10.b Verify Aux Bus E To Emergency Bus BSB Breaker 124 CLOSED Standard: Locates Aux Bus D to Emergency Bus B-SB Breaker 124 switch and takes switch to CLOSE (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 12 Performance Step: 11.b Verify Diesel Generator BSB Breaker 126 B SB OPEN Standard: Locates Diesel Generator BSB Breaker 126 B SB switch and verifies breaker is Open (GREEN LIGHT LIT)

Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 16 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 13.a Performance Step: 12.b Energize 6.9 KV Bus BSB:

d. Place Emergency Bus BSB To Aux Bus E Synchronizer control switch to SYNC.

Standard: Locates Synchronizer control switch for Emergency Bus BSB To Aux Bus E and places control to SYNC Comment:

EOP ECA-0.0, Attachment 1, Step 13.b Performance Step: 13.b e. Close Emergency Bus BSB To Aux Bus E Tie Breaker 125.

Standard: Locates switch for Emergency Bus BSB To Aux Bus E Tie Breaker 125 and takes switch to CLOSE.

(GREEN LIGHT STAYS LIT) - Reports to CRS that Emergency Bus BSB To Aux Bus E Tie Breaker 125 will not close (may dispatch AO to investigate)

Acknowledge report that Emergency Bus BSB To Aux Evaluator Cue:

Bus E Tie Breaker 125 will not close.

IF AO is dispatched: Acknowledge and repeat back Simulator Communicator:

communications to investigate breaker IF needed to get the candidate back on task: Ask for an Evaluator NOTE: estimation on when power will be restored to an Emergency Bus.

Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 17 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 2.a - Alternate Path Begins Restoration of power from the Start Up XFMR 1A to the A-SA Emergency Bus EOP ECA-0.0, Attachment 1, Step 2.a Performance Step: 14.b On Start Up XFMR Protective Relay Panel 1A, verify offsite power to Start Up XFMR A:

a. Verify the Start Up XFMR 1A Lockout SU 1A Relay is reset.

Standard: Locates Startup XFMR 1A Lockout SU 1A Relay and verifies that the relay is reset. (Relay is reset)

Comment:

EOP ECA-0.0, Attachment 1, Step 2.b Performance Step: 15.b b. Verify closed any of the following switch yard tie breakers to energize Start Up XFMR A:

• Breaker 522

• Breaker 523 Standard: Locates tie breaker switches for Startup XFMR A

• Breaker 522 (Verifies already closed)

• Breaker 523 (Not required to be closed but maybe closed w/o consequences)

Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 18 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 3.a Performance Step: 16.b Restore offsite power to 6.9 KV Aux Bus D:

d. Place Start Up XFMR To Aux Buses A & D Synchronizer control switch to BREAKER 101 position.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses A & D and places switch to Breaker 101 position Comment:

EOP ECA-0.0, Attachment 1, Step 3.b Performance Step: 17.b e. Close Start Up XFMR A To Aux Bus D Breaker 101.

Standard: Locates switch for Start Up XFMR A To Aux Bus D Breaker 101 and places switch to CLOSE. (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 3.c Performance Step: 18.b f. Place Start Up XFMR To Aux Buses A & D Synchronizer control switch to OFF.

Standard: Locates Synchronizer control switch for Start Up XFMR To Aux Buses A & D and places switch to OFF Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 19 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 4 Performance Step: 19.b Verify Aux Bus D To Emergency Bus ASA Breaker 104 CLOSED Standard: Locates Aux Bus D to Emergency Bus A-SA Breaker 104 switch and takes switch to CLOSE (RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 5 Performance Step: 20.b Verify Diesel Generator ASA Breaker 106 A SA OPEN Standard: Locates Diesel Generator ASA Breaker 106 A SA switch and verifies breaker is Open (GREEN LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 6.a Performance Step: 21.b Energize 6.9 KV Bus ASA:

b. Place Emergency Bus ASA To Aux Bus D Synchronizer control switch to SYNC.

Standard: Locates Synchronizer control switch for Emergency Bus ASA To Aux Bus D and places control to SYNC Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 20 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 6.b Performance Step: 22.b c. Close Emergency Bus ASA To Aux Bus D Tie Breaker 105.

Standard: Locates switch for Emergency Bus ASA To Aux Bus D Tie Breaker 105 and takes switch to CLOSE.

(RED LIGHT LIT)

Comment:

EOP ECA-0.0, Attachment 1, Step 6.c Performance Step: 23.b a. Place Emergency Bus ASA To Aux Bus D Synchronizer control switch to OFF.

Standard: Locates Synchronizer control switch for Emergency Bus ASA To Aux Bus D and places control to OFF Comment:

EOP ECA-0.0, Attachment 1, Step 7 Performance Step: 24.b Close the following 6.9 KV breakers:

• Emergency Bus ASA To XFMR A1SA Breaker A1 ASA

• Emergency Bus ASA To XFMR A3SA Breaker A3 ASA Standard:

• Locates control switch for Emergency Bus ASA To XFMR A1SA Breaker A1 ASA and places control to CLOSE (RED LIGHT LIT)

• Locates control switch for Emergency Bus ASA To XFMR A3SA Breaker A3 ASA and places control to CLOSE (RED LIGHT LIT)

Comment:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 21 of 23 Form ES-C-1 PERFORMANCE INFORMATION EOP ECA-0.0, Attachment 1, Step 8 Performance Step: 25.b Verify 6.9 KV Emergency Bus ASA To XFMR A2SA Breaker A2 ASA CLOSED Standard:

• Locates control switch for 6.9 KV Emergency Bus ASA To XFMR A2SA Breaker A2 ASA and places control to CLOSE (RED LIGHT LIT)

Informs CRS that power is restored to Emergency Bus A-SA Acknowledge any reports:

After the 6.9 KV Emergency Bus ASA power is restored:

Evaluation on this JPM is complete.

Evaluator Cue:

I have the shift, END OF JPM Inform Simulator Operator to place the Simulator in Freeze.

Comment:

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 22 of 23 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam Simulator JPM g Restoration of Offsite Power to Emergency Buses EOP ECA-0.0, Loss Of All AC Power, Attachment 1 Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• The unit was operating at 100% power

• A EDG is under clearance due to a failure that caused the Generator field to not flash during OST-1013 Subsequently:

• A failure of a transmission line on the Duke grid resulted in the Initial Conditions:

cascading trip of several units which resulted in low grid frequency

• The HNP unit has experienced a loss of offsite power

• B EDG failed to start. The problem is being investigated

• The crew entered ECA-0.0, Loss Of All AC Power

• The load dispatcher has contacted HNP and informed the MCR that the grid is now stabile

• Your position is the BOP

• The CRS has directed you to restore offsite power to a (one) AC emergency bus using ECA-0.0 Attachment 1.

Initiating Cue:

• The Load Dispatcher has given permission to restore offsite power to 6.9 KV buses and to reset any tripped Start Up XFMR lockout relays.

2016 HNP NRC Exam Simulator JPM CR g Rev. 2

Appendix C Page 1 of 15 Form ES-C-1 Worksheet Facility: Harris Nuclear Plant Task No.: 088018H101 Task

Title:

Restoring the Control Room Area JPM No.: 2016 HNP NRC Exam HVAC System to Normal After a Simulator JPM CR h Control Room Isolation Signal K/A

Reference:

APE067 AA1.05 RO 3.0 SRO 3.1 ALTERNATE PATH - YES Examinee: ________________________ NRC Examiner: _________________

Facility Evaluator: ________________________ Date: ________

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

• The plant is operating at 100% power when a fire occurred at the Dedicated Shutdown Diesel Generator during testing.

• The smoke from the fire caused a Control Room Ventilation Isolation signal to occur.

Initial Conditions:

(Smoke detected at the normal intake Zone 1-150)

• The Fire Brigade has put the fire out and the smoke has been cleared.

• Your position is the BOP.

• The CRS has directed you to restore the Control Room Area HVAC System to normal in accordance with OP-173, Control Initiating Cue:

Room Area HVAC System, Section 8.4.

• The initial conditions are satisfied and the HVAC system is in operation per section 8.1 of OP-173.

2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 2 of 15 Form ES-C-1 Worksheet Task Standard: Place the Control Room Area HVAC system in normal operation Required Materials: None General

References:

OP-173, Control Room Area HVAC System, Rev. 37, ALB-030-6-4, Rev. 35 Handout: Use simulator copy of OP-173, Rev. 37 AND APP-ALB-030, Rev 35 Time Critical Task: No Validation Time: 15 minutes CRITICAL STEP JUSTIFICATION Must reset both trains of Control Room ventilation or the system cannot Step 4 be taken out of the Emergency filtration lineup.

Must open the normal intake valves to return system lineup to normal Step 7 flow path for operation.

Must start a normal exhaust fan to obtain flow and obtain correct damper Step 10 alignment.

Must stop both emergency filtration fans to return to normal filtration Step 12 lineup and to shift dampers back to normal lineup.

Must shut emergency exhaust Recirc dampers to complete normal Step 13 alignment of Control Room ventilation system.

Step 18 Must start the standby fan to re-establish Main Control Room ventilation.

2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 3 of 15 Form ES-C-1 Worksheet 2016 NRC Exam - SIMULATOR SETUP Simulator Operator

• Reset to IC-171

• Password noinstants

• Go to RUN

• Ensure RM-11 is NORMAL

• Silence and Acknowledge annunciators GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner. (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

To recreate the IC setup for this JPM:

• Initial Simulator IC was IC go to RUN

• Enter a Control Room Isolation Signal (CRIS) (the JPM has the signal in due to smoke but there isnt a relay that can be individually used so use the one for Radiation) o irf rms011 (n 00:00:00 00:00:00) 0.0005 00:00:00 o irf rms013 (n 00:00:00 00:00:00) 0.0005 00:00:00

• After the fans and dampers have completed switching positions reset the rad monitors to 1e-7 or they will not clear (value is < the alarm setpoint on RM-11) o mrf rms011 (n 00:00:00 00:00:00) 1e-007 00:00:00 o mrf rms013 (n 00:00:00 00:00:00) 1e-007 00:00:00 o Reset RM-11 back to normal

• Start the Motor Driven Fire Pump o irf msc029 (n 0 0 ) START

• Place 2 alarms to on to simulate a fire

• Fire Detection System Trouble o ian xn30a07 (n 0 0) ALARM_ON

• Reflash Fire Pump System Trouble o ian xn30b07 (n 0 0) ALARM_ON

• Create a conditional Trigger to trip AH-15 A SA when the control switch for Emergency Filtration Recirc Damper CA-D61 SB is taken to SHUT To create the conditional trigger:

- Go to malfunctions

- Find Control Room Normal Support Fan AH-15-1A assign Trigger 1 with switch to STOP and GREEN light OFF o ior xdi085 (1 00:00:00 00:00:00) STOP o ior xd2o085 (1 00:00:00 00:00:00) OFF

- Wait for ALB 30-3-3 CNT Room Air Low P to alarm

- Find Annuciator ALB-030-6-4 assign Trigger 1 to alarm on o ior xn30d06 (1 00:00:00 00:00:00) ALARM_ON

- Go to triggers o Click on Trigger 1 o Click on Assign File o Choose CZD61Shut o (source file should now have CZD61Shut)

• Silence, Acknowledge and Reset the annunciators

• Freeze and Snap and save these conditions to your exam IC 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 4 of 15 Form ES-C-1 PERFORMANCE INFORMATION Simulator Operator: When directed by the Lead Examiner go to Run.

START TIME:

Performance Step: 1 Previews procedure Standard: OP-173 and refers to Section 8.4 IF asked by candidate Initial Conditions have been satisfied:

Control Room Isolation Signal is clear and the Control Room Evaluator Cue: Area HVAC System in operation per Section 8.1, 8.2 or 8.3 Comment:

OP-173, Section 8.4 Note prior to Step 1 Performance Step: 2 NOTE: The following Step will cause ALB-030/1-1, Control Room Isolation Train A, and ALB-030/2-1, Control Room Isolation Train B, to clear Standard: Operator reads and placekeeps any note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 5 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 8.4 Caution prior to Step 1 Performance Step: 3 CAUTION: Failure of equipment to secure in this section will result in the associated EDG being inoperable. Tech Spec 3.8.1.1 is applicable until the breaker for the affected load is opened.

Standard: Operator reads and placekeeps any note or caution (initials, checks or circle/slash)

Comment:

OP-173, Section 8.4 Step 1 Performance Step: 4 PLACE the CONTROL ROOM ISOL TRAIN A and B RESET switches to RESET.

• CONTROL ROOM ISOL TRAIN A RESET

• CONTROL ROOM ISOL TRAIN B RESET Standard: Locates and momentarily operates the Control Room Isol Train A reset switch to reset and the Train B reset switch to reset Operator resets alarm using reset button on MCB (May report to CRS that the Control Room Isolation signal for Train A and B are clear)

Evaluator Cue: Acknowledge any report.

Comment: Note: this will cause ALB-030/1-1 and ALB-030/2-1 alarms to clear.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 6 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 8.4, Step 2 Performance Step: 5 Shut any of the following EMER FILT SOUTH (NORTH)

OUTSIDE AIR INLET valves that are open, (any that are not open may be marked N/A):

• EMER FILT SOUTH OUTSIDE AIR INLET 1CZ-9 SA

• EMER FILT SOUTH OUTSIDE AIR INLET 1CZ-10 SB

• EMER FILT NORTH OUTSIDE AIR INLET 1CZ-11 SA

• EMER FILT NORTH OUTSIDE AIR INLET 1CZ-12 SB Standard: Locates valves and verifies ALL outside air inlets are closed.

(Position indication lights are all green.)

Comment:

OP-173, Section 8.4, Note prior to Step 3 Performance Step: 6 NOTE: Performing steps 8.4.2.3 through 8.4.2.6 quickly will minimize excessive pressurization of the Main Control Room Standard: Operator reads and placekeeps any note or caution (initials, checks or circle/slash).

Comment:

OP-173, Section 8.4, Step 3 Performance Step: 7 OPEN the following Control Normal Outside Air Intake Valves:

• NORMAL INTAKE 1CZ-1 SA

• NORMAL INTAKE 1CZ-2 SB Standard: Locates and Opens NORMAL INTAKES 1CZ-1 SA and 1CZ-2 SB (Green Light goes off and Red Light comes on)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 7 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 8.4, Step 4 Performance Step: 8 If more than one NORMAL SUPPLY FAN AH-15 ASA (BSB) is running, stop one fan.

Standard: Identifies ONLY one NORMAL SUPPLY FAN is running (AH-15 ASA) and there is no need to stop a fan Comment:

OP-173, Section 8.4, Step 5 Verify associated valves/dampers align for the stopped train as follows:

AH-15 IN CZ-D1 (CZ-D2) Shut (indication) on SLB-5 (6)

Performance Step: 9 AH-15 IN CZ-25 (CZ-26) Shut (indication) on SLB-5 (6)

CONT ROM NORMAL Shut RECIRC DAMPER CZ-D69 SA (CZ-D70 SB)

Standard: Locates and verifies the associated valves/dampers aligned for the stopped train.

AH-15 IN CZ-D1 (CZ-D2) Shut (indication) on SLB-5 (6)

AH-15 IN CZ-25 (CZ-26) Shut (indication) on SLB-5 (6)

CONT ROM NORMAL Shut RECIRC DAMPER CZ-D69 SA (CZ-D70 SB)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 8 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 8.4, Step 6 Performance Step: 10 Start NORMAL EXHAUST FAN E-9 A (B)

Standard: Takes Normal Exhaust Fan E-9 A fan switch to start, confirms RED light is lit Comment:

OP-173, Section 8.4, Step 7 Performance Step: 11 Verify the following valves/dampers are aligned as indicated:

E-9A(B) IN CZ-D6 (CZ-D7) Open (located on SLB-7)

E-9A(B) OUT CZ-D12 (CZ-13) Modulates (located on SLB-7)

NORMAL EXHAUST Open 1CZ-3 SA and 1CZ-4 SB Standard: Verifies:

E-9A IN CZ-D6 Open (located on SLB-7)

E-9A OUT CZ-D12 Modulates (located on SLB-7)

NORMAL EXHAUST Open 1CZ-3 SA and 1CZ-4 SB Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 9 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 8.4, Steps 8, 9 and 10 Performance Step: 12 If running, THEN STOP BOTH EMERGENCY FILTRATION FAN R-2 A-SA and R-2 B-SB and verify:

R2 INLET CZ-23 (CZ-24) Shut [located on SLB-5 (6)]

R2 DISCH CZ-21 (CZ-22) Shut [located on SLB-5 (6)]

EMERGENCY FILTRATION Shut DISCHARGE 1CZ-19 SA and 1CZ-20 SB Standard: Locates and stops both EMERGENCY FILTRATION FANS R-2 A-SA and R-2 B-SB (critical to stop fans) and verifies:

R2 INLET CZ-23 (CZ-24) Shut [located on SLB-5 (6)]

R2 DISCH CZ-21 (CZ-22) Shut [located on SLB-5 (6)]

And verifies that EMERGENCY FILTRATION DISCHARGE Shut 1CZ-19 SA and 1CZ-20 SB Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 10 of 15 Form ES-C-1 PERFORMANCE INFORMATION NOTE: Alternate Path Starts Here When the candidate shuts CZ-D61 SB a conditional trigger Lead Examiner and will trip the running Control Room Emergency Supply Fan Simulator Operator:

breaker (AH-15 ASA breaker 1A36-SA-5A )

OP-173, Section 8.4, Step 11 Performance Step: 13 Shut the EMERGENCY FILTRATION RECIRC dampers EMERGENCY FILTRATION RECIRC DAMPER CZ-D66 SA and EMERGENCY FILTRATION RECIRC DAMPER CZ-D61 SB Standard: Locates the control switches and SHUTS the EMERGENCY FILTRATION RECIRC dampers CZ-D66 SA (Green Light On) and CZ-D61 SB (supply breaker 1A36-SA-5A trips)

Green light On Comment:

Supply breaker 1A36-SA-5A for AH-15 ASA trips open Performance Step: 14 Annunciator ALB-030-6-4, CONT ROOM HVAC NORMAL SUPPLY FANS AH-15 LOW FLOW-O/L Standard: Acknowledges alarm and identifies that AH-15A SA has lost green indication on MCB and reports information to CRS.

Pulls APP and reviews response for alarm The CRS acknowledges the report. If the candidate identifies the AH-15A SA has tripped and determines the Evaluator Cue: APP will be addressed after the task to restore ventilation is complete as the CRS direct the candidate to address the APP before continuing with the task.

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 11 of 15 Form ES-C-1 PERFORMANCE INFORMATION Response to ALB-030-6-4, CONT ROOM HVAC NORMAL SUPPLY FANS AH-15 LOW FLOW-O/L Performance Step: 15 PERFORM Corrective Actions:

a. CHECK AH-15 fans status indication on MCB.

b. IF fan is tripped, THEN PERFORM the following:

(1) START the standby fan using OP-173, Control Room Area HVAC System.

(2) IF white fan trouble light is LIT, THEN DISPATCH an operator to check overload relays on 1A36-SA-5A or 1B36-SB-3A.

(3) DISPATCH an operator to check for tripped breaker on 1A36-SA-5A or 1B36-SB-3A.

c. CHECK damper alignment on MCB for CZ-D1SA-1, CZ-D2SB-1, CZ-25 and CZ-26.

d. IF alb-030-6-3 is ALARMING, THEN REFER TO ALB-030-6-3 Standard:

• Identifies that AH-15 ASA has tripped and the white fan trouble light is NOT lit

• Start the standby fan:

o Obtains a copy of OP-173 section 5.1, Control Room Area HVAC and starts the standby fan Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 12 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 5.1, Startup of Normal Supply and Exhaust Fans Performance Step: 16 PERFORM Corrective Actions:

(1) START the standby fan using OP-173, Control Room Area HVAC System.

Section 5.1.1 Initial Conditions

1. Attachment 1 is complete

2. Attachment 2 is complete Standard: Reviews initial conditions of Attachment 1 and Attachment 2 for completeness Lead Evaluator Cue: Using time compression the Initial Conditions are met.

Standard: Initials section 5.1.1 step 1 and 2 for initial conditions complete Comment:

OP-173, Section 5.1.2, Notes prior to step 1 Performance Step: 17 NOTE: The following Steps align Train A Control Room Area HVAC components to service. Train B nomenclature is in parenthesis.

NOTE: If Swapping Control Room Ventilation Fans, it is preferable to secure the running fan first, then start the desired fans with this section.

Standard:

• Operator reads and placekeeps any note or caution (initials, checks or circle/slash)

Comment:

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 13 of 15 Form ES-C-1 PERFORMANCE INFORMATION OP-173, Section 5.1.2, step 1 Performance Step: 18 START the NORMAL SUPPLY FAN AH-15 BSB Standard: Locates MCB switch for AH-15 BSB and takes switch to start Comment:

OP-173, Section 5.1.2, step 2 Performance Step: 19 VERIFY that the following components are aligned properly:

AH-15 IN CZ-D2 ..Open SLB-6 AH-15 DISCH CZ-26 ..Open SLB-6 NORMAL INTAKE 1CZ-1 SA and 1CZ-2 SB.. .Open CONT RM NORMAL RECIRC DAMPER CZ-D70 SB Open NORMAL EXHAUST FAN E-9B.. Running E-9B IN CZ-D7 .Open SLB-7 E-9B OUT CZ-D13 .Modulates SLB-7 NORMAL EXHAUST 1CZ-3 SA and 1CZ-4 SB ..Open Standard: VERIFIES that the following components are aligned properly:

AH-15 IN CZ-D2 ..Open SLB-6 AH-15 DISCH CZ-26 ..Open SLB-6 NORMAL INTAKE 1CZ-1 SA and 1CZ-2 SB.. .Open CONT RM NORMAL RECIRC DAMPER CZ-D70 SB Open NORMAL EXHAUST FAN E-9B.. Running E-9B IN CZ-D7 .Open SLB-7 E-9B OUT CZ-D13 .Modulates SLB-7 NORMAL EXHAUST 1CZ-3 SA and 1CZ-4 SB ..Open After verification of step 2 components CUE: Another Operator will continue with any remaining Evaluator Cue:

ventilation restoration. I have the shift. END OF JPM Direct Simulator Operator to place the Simulator in Freeze.

STOP TIME:

Simulator Operator: When directed by the Lead Examiner then go to Freeze.

- Denotes Critical Steps 2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C Page 14 of 15 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2016 HNP NRC Exam Simulator JPM h Restoring the Control Room Area HVAC System to Normal After a Control Room Isolation Signal OP-173, Control Room Area HVAC System Examinees Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiners Signature: Date:

2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix C JPM CUE SHEET Form ES-C-1

• The plant is operating at 100% power when a fire occurred at the Dedicated Shutdown Diesel Generator during testing.

• The smoke from the fire caused a Control Room Ventilation Isolation signal to occur.

Initial Conditions:

(Smoke detected at the normal intake Zone 1-150)

• The Fire Brigade has put the fire out and the smoke has been cleared.

• Your position is the BOP.

• The CRS has directed you to restore the Control Room Area HVAC System to normal in accordance with OP-173 Control Initiating Cue:

Room Area HVAC System, Section 8.4.

• The initial conditions are satisfied and the HVAC system is in operation per section 8.1 of OP-173.

2016 HNP NRC Exam Simulator JPM CR h Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 Facility: Harris Nuclear Plant Scenario No.: 1 Op Test No.: 05000400/2016301 Examiners: Operators: SRO:

OATC:

BOP:

Initial Conditions: IC-19: 100%, MOL

• The unit is operating at 100% power

• The following equipment is under clearance o B MD AFW Pump o B NSW pump o Boric Acid Transfer Pump B-SB

• Engineering reports that there are high motor vibrations on the A Heater Drain Pump.

• Directions from the Operations Manager are to conservatively reduce power Turnover:

IAW GP-006 at 4 DEH units/min to 95% then remove A Heater Drain Pump from service IAW OP-136 Section 7.1.

• Plant risk condition is YELLOW due to shut down.

• Initiate a MANUAL Turbine trip prior to automatic Low Steam Line SI signal Critical Tasks:

• Establish condensate flow to the SGs before RCS bleed and feed is required Event Malf. No. Event Type* Event Description No.

R - RO/SRO 1 N/A Lower power to stop A HD Pump (to 95% power)

N - BOP/SRO 2 CRF008 I - RO/SRO T-ref Controller fails low (AOP-001)

I - BOP/SRO 3 LT:486 SG B Controlling Level Channel fails LO TS - SRO 4 LT-115 I - RO/SRO VCT Level Channel 115 fails LOW (AOP-003)

C - RO/SRO 5 SIS03C C Accumulator nitrogen leak TS - SRO 6 TUR24A C - BOP/SRO EHC pump shaft shear with standby auto start failure CFW16A MFW Pump A trip (AOP-010 to E-0) 7 M - ALL CFW16B MFW Pump B trip (2 minute delay) - Loss Of Heat Sink 8 TUR02 C - BOP/SRO AUTO Turbine Trip fails (manual successful) 9 CFW01B C - BOP/SRO MD AFW Pump A breaker trips when started 10 CFW01C C - BOP/SRO TD AFW Pump trips when running (FR-H.1)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 Engineering has reported that the A Heater Drain Pump has high vibrations and is recommending that the pump be secured as soon as possible. Directions from the Operations Manager are to conservatively reduce power to 95% IAW GP-006 at 4 DEH units/min then remove A Heater Drain Pump from service IAW OP-136 section 7.1. All required notifications have been made to individuals concerning the reason for the downpower. Plant risk condition is YELLOW due to the upcoming shutdown.

The following equipment is under clearance:

• B MDAFW Pump is under clearance for pump packing repairs. The pump has been inoperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and will be restored to operable status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Tech Spec 3.7.1.2 LCO Action a and Tech Spec 3.3.3.5.b Action c applies. 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO or HSB within the next 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, HSD following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Equipment is under clearance continued:

B MDAFW Pump - (continued) - Tech Spec 3.3.3.5.b Action c

• B NSW Pump under clearance for shaft inspection. The pump has been under clearance for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Inspection and return to service are expected to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Boric Acid Transfer Pump B-SB is under clearance due to breaker blown control power fuses. Has been under clearance for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The problem with the breaker has been repaired and the clearance will be removed later this shift. Tech Spec 3.3.3.5.b which is a 7 day LCO and 3.1.2.2 applies (3.1.2.2 is for tracking only). OWP-CS-05 has been completed.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Boric Acid Transfer Pump B-SB (Tech Spec)

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Event 1 The first evolution for the crew is to commence a power reduction at 4 DEH Units / min from 100% to 95% in preparations of securing the A Heater Drain Pump. It is expected that the SRO will conduct a reactivity brief.

Verifiable actions: The RO will borate as necessary to lower power and monitor automatic operation of rod control. The BOP will operate the DEH Main Turbine controls as necessary to reduce turbine load. The BOP will secure the A Heater Drain pump.

Event 2 Failure of the Tref Processor (fails low). The crew should enter AOP-001 and carry out the immediate actions.

Verifiable actions: The OATC will perform the immediate actions of AOP-001 by verifying that

<2 rods are dropped (no rods have dropped), place Rod Control in MANUAL and then verify no rod motion. With concurrence from the SRO the OATC will restore Tave to pre-failure conditions by withdrawing the rods in manual.

Event 3 SG B controlling level transmitter LT-486 fails low. Flow and level will rise as observed on B SG FI-486, FI-487 and NR level instruments LI-483, LI-484 and LI-485.

Verifiable actions: The BOP should report and respond to annunciator ALB-014-5-3A Steam Gen B NR Low Level. The BOP should take manual control of the B SG flow control valve and lower the output of the M/A station to restore level to 57% Narrow Range.

The SRO should provide trip limits and level bands IAW OMM-001 Attachment 13. The SRO directs the implementation of OWP-RP-06.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Event 3 Tech Spec Evaluation: The SRO should evaluate Tech Specs 3.3.1, Reactor Trip Instrumentation, Table 3.3-1 Items 13 Steam Generator Water Level -Low - Low and item 14 SG Water Level - Low Coincident With Steam/Feedwater Flow Mismatch, Action 6 for both and Tech Spec 3.3.2, ESF Instrumentation, Table 3.3-3 Item 5b, Action 19 and Tech Spec 3.3.3.6 Accident Monitoring.

• NOTE: IF the crew does not respond to the low water level in the SG a High level

(> 78%) will develop which will cause an automatic Turbine trip and a Reactor trip since power is > 10% (REACTOR TRIP TURBINE TRIP P7) .

An automatic Reactor Trip for this event would create an unanticipated critical task.

(See Note after critical task justification statements for details on unanticipated critical tasks.)

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Event 4 Failure of VCT level transmitter LT-115 (low). This failure will cause an automatic make up to the VCT to initiate. The crew should identify the failed level transmitter then enter AOP-003, Malfunction of Reactor Makeup Control. There are no immediate actions associated with AOP-003.

Verifiable actions: The OATC will respond to the Reactor Makeup Control system malfunction and place the Reactor Makeup Water Control Switch to STOP. This action will secure the unneeded VCT makeup caused by the level transmitter failure. The SRO will contact Maintenance to investigate and repair the failure.

Event 5 C Accumulator nitrogen leak causes pressure to decrease until annunciator ALB 01-9-1 alarms.

Verifiable actions: The crew should respond to the low pressure condition and restore the Accumulator pressure to normal. The OATC will verify open 1SI-287, ACCUMULATORS &

PRZ PORV N2 SUPPLY then OPEN the ACCUM N2 SUPPLY/VENT for the C Accumulator:

1SI-297 for ACCUMULATOR C N2 Supply & Vent.

Tech Spec Evaluation: The SRO should declare the C Accumulator inoperable per Tech Spec 3.5.1, due to being connected to Non-Safety piping (a one hour action statement in Modes 1 through 3 above 1000 psig). Additionally, The SRO should evaluate Tech Spec 3.5.1 for Accumulator pressure if pressure gets below Tech Spec operability limit.

3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a. The isolation valve open with power supply circuit breaker open,

b. A contained borated water volume of between 66 and 96% indicated level,

c. A boron concentration of between 2400 and 2600 ppm, and

d. A nitrogen cover-pressure of between 585 and 665 psig ACTION:a. With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Event 6 DEH pump shaft shears and auto start failure of the standby DEH pump. When DEH pressure decreases to < 1600 psig annunciator ALB-020-4-2B EH Fluid Low Press will alarm.

Verifiable actions: The BOP should manually start the standby DEH pump prior to system pressure reducing to <1500 psig. If the standby pump is not started prior to DEH pressure decreasing to < 1500 psig they should identify that the standby pump auto start feature has also malfunctioned. Once the standby pump is started DEH system pressure will return to normal.

With DEH pressure decreasing the crew should dispatch an Aux Operator to investigate the DEH system for indications of leakage or pump failure.

• NOTE: IF the crew does not respond to this event by starting the standby DEH pump they would enter AOP-038, Rapid Downpower when DEH fluid pressure drops to

<1500 psig. Entering AOP-038 is ONLY appropriate IF the standby pump is not available. IF the crew fails to start the standby DEH pump system pressure will continue to slowly decrease. When system pressure decreases to 1150 psig an automatic Turbine Trip on Auto Stop Oil Pressure will occur. (ALB-018 window 3-4). Without any actions the crew would create an unanticipated critical task by allowing the Turbine trip to occur which would then cause an automatic Reactor Trip (Reactor Trip Turbine Trip P-7).

(See Note after critical task justification statements for details on unanticipated critical tasks.)

Event 7 - MAJOR (Leading to Loss of Heat Sink)

Main Feedwater Pump A trips with Reactor Power > 90%. The crew should respond to ALB-016-1-4, FW Pump A/B O/C Trip -Gnd or Bkr Fail to Close and AOP-010, Feedwater Malfunctions and perform the immediate actions.

Verifiable actions: The OATC should manually trip the Reactor IAW AOP-010 actions for loss of a Feed Water pump with initial Reactor power above 90%.

Two minutes after the A MFW pump trips the B MFW pump will trip. The crew should recognize that the second pump has tripped by MCB annunciators, pump indications and Steam Generator level changes.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

Event 8 Main Turbine fails to automatically trip. In accordance with EOP-E-0, Reactor Trip or Safety Injection immediate action 2 the BOP should identify the automatic trip failure of the Main Turbine due to none of the Turbine Throttle valves indicating closed.

Verifiable actions: The BOP should manually trip the Turbine in the preferred order: Turning the MCB Turbine trip switch to trip (this action if performed will be successful). If this action is NOT performed RNO actions would be: depress the Turbine Manual pushbutton on DEH then simultaneously depress the Fast Action and GV Lower push buttons or shut the MSIVs. IF the follow up actions are performed without first manually tripping the Turbine using the trip switch it does not constitute a failure since the Turbine will be tripped but the trip would be unnecessarily delayed.

Event 9 MD AFW Pump A trips immediately upon starting.

Verifiable actions: The BOP will identify a MD AFW pump trip by annunciator ALB017-5-4, Aux Feedwater Pump A Trip or Close Ckt Trouble alarm. At this time IF the TD AFW pump has not auto started on SG low levels the BOP should determine that starting the TD AFW pump is required and open both MS-70 and MS-72 steam supply valves to the TD AFW pump to maintain AFW flow to the Steam Generators.

Event 10 Event 10: TDAFW Pump trips, (timing controlled by Lead Evaluator) while implementing EOP-ES-0.1, Reactor Trip Response. Annunciator ALB-017-7-4, Aux Feedwater Pump Turbine Trip.

Once EOP-ES-0.1 is entered the first step to Implement Function Restoration Procedures as required. A RED path will exist for EOP FR-H.1, Response To Loss of Secondary Heat Sink due to a loss of all Feedwater flow to the SGs (< 210 KPPH) and Narrow Range levels in ALL SGs < 25%. Once identified the crew should make transition to EOP-FR-H.1.

Verifiable actions: The OATC will secure ALL RCPs while the BOP continues efforts to restore AFW flow. The OATC will depressurize the RCS and block auto SI signals. The BOP will depressurize one SG to < 500 psig then establish Condensate flow to the SGs.

The scenario ends when Condensate flow is established and verified to one SG.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 1 (continued)

CRITICAL TASK JUSTIFICATION:

1. Initiate a MANUAL Turbine trip (Prior to the receipt of the Low Steam Line Pressure SI signal)

Without tripping the main Turbine adverse consequences or significant degradation in the mitigative capability of the plant will occur. Taking actions will prevent the actuation of the ECCS. If the ECCS is allowed to actuate it will fill the RCS and challenge the RCS pressure boundary by challenging the Pressurizer PORVs and code Safety valves. Also, WOG ERG-Based Critical task E-0, Manually trip the main Turbine before a severe (Orange Path) challenge develops to either the Subcriticality or the Integrity CSF or before transition to EOP-ECA-2.1, Uncontrolled Depressurization of ALL Steam Generators) whichever happens first.

2. Establish Condensate flow to the SGs before RCS bleed and feed required Failure to establish feedwater flow to any SG results in the crews having to rely upon the lower-priority action of establishing RCS bleed and feed to minimize core uncovery. This constitutes incorrect performance that fails to prevent degradation of any barrier to fission product release. The analyses presented in the ERG Background Document for FR-H.1 demonstrate that a complete loss of heat sink occurs when the SG inventories deplete (dry out). Unless some form of SG inventory is restored, the SG dryout deteriorates primary-to-secondary heat transfer, allowing core decay heat to increase the RCS temperature and pressure. The increasing RCS pressure automatically forces the pressurizer PORVs to open, which creates a small-break LOCA and simultaneously degrades the RCS fission-product barrier. As long as the RCS pressure remains high, the flow out the PORVs exceeds the ECCS flow into the RCS, which depletes RCS inventory. Eventually the core starts to uncover, degrading the core cooling CSF.

Once the core is uncovered, fuel temperatures increase rapidly until severe fuel damage occurs, unless some form of core cooling is restored. Fuel over-heating constitutes severe degradation of a fission-product barrier (fuel matrix/clad). Establishing feedwater flow into the SGs offers the most effective recovery action to restore the heat sink. The introduction of feedwater flow immediately restores SG inventory and re-establishes primary-to-secondary heat transfer, decreasing RCS pressure and cooling the core. The RCS pressure decrease then precludes the opening of the PORVs and degradation of the RCS barrier.

Note: An unanticipated critical task may be created in a scenario should an applicants action or lack of action cause an unexpected RPS or ESFAS actuation. A critical task may be assigned and graded as unsatisfactory even if corrected by another team member prior to the unanticipated RPS/ESFAS actuation. Should the applicant self-correct the action or inaction prior to the unanticipated plant response, a critical task failure should not be assigned to the applicant.

Harris 2016 NRC Exam Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC SCENARIO 1 SIMULATOR SETUP For the 2016 NRC Exam Simulator Scenario # 1 Reset to IC-161 password noinstants Go to RUN Silence and Acknowledge annunciators GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner.

Set ERFIS screens (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

SPECIAL INSTRUCTIONS Post conditions for status board from IC-19 Reactor Power 100% steady state Control Bank D at 218 steps RCS boron 1009 ppm Update the status board: "B" MDAFW Pump is OOS for motor overhaul Pump has been OOS for 12 total hours and is expected back within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Tech Spec 3.7.1.2, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO or HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, HSD following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Hang restricted access signs on MCR entry swing gates Hang CIT on B MDAFW Pump MCB switch then place protected train placards per OMM-001 Attachment 16 on "A" MDAFW Pump, MS-70 and 72, "B" ESW Pump, "B" RHR Pump and "B" CCW Pump "B" NSW pump Out Of Service for breaker repairs Repairs to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Place CIT on switch for "B" NSW and place protected train placard on "A" NSW pump switch Hang CIT on Boric Acid Transfer Pump B-SB Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 12 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior When the crew has completed their board walkdown and are ready to assume the shift then direct the Simulator Evaluator Note: Operator to place the Simulator in RUN.

After the Simulator is in RUN announce to the crew that they have the shift.

Simulator Operator: When directed by the Lead Evaluator go to Run.

SRO GP-006, Step 6.2.4.

When PRZ backup heaters are energized in manual, PK-444A1, PRZ Master Pressure Controller (a PI controller) will integrate up to a greater than normal output, opening PRZ Spray Valves to return and maintain RCS pressure at setpoint.

The result is as follows:

Procedure Note:

• PORV PCV-444B will open at a lower than expected pressure.

• ALB-009-3-2, PRESSURIZER HIGH PRESS DEVIATION CONTROL, will activate at a lower than expected pressure.

• Increased probability for exceeding Tech Spec DNB limit for RCS pressure.

OATC OP-100 section 8.15

• PLACE PK-444A, PRZ PRESS CONTROL, in Manual

• ADJUST output of PK-444A to between 40 and 45%.

OATC

• PLACE PK-444A, PRZ PRESS CONTROL, in Auto

• IMMEDIATELY PLACE desired Backup Heater Control switches to ON.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 13 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior Indicated PRNI power may increase >100% if the Turbine ramp is not started after energizing all Pressurizer Heaters.

Evaluator Note:

The crew may elect to begin boration prior to lowering turbine load.

OATC OP-107.01, Section 5.2

• DETERMINE the reactor coolant boron concentration from chemistry OR the Main Control Room status board.

• DETERMINE the magnitude of boron concentration OATC increase required.

• DETERMINE the volume of boric acid to be added using the reactivity plan associated with the IC.

FIS-113, BORIC ACID BATCH COUNTER, has a tenths Procedure Note:

position.

If the translucent covers associated with the Boric Acid and Total Makeup Batch counters FIS-113 and FIS-114, located on Procedure Caution:

the MCB, are not closed, the system will not automatically stop at the preset value.

SET FIS-113, BORIC ACID BATCH COUNTER, to obtain the OATC desired quantity.

SRO Directs boration Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 14 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior

• Boric Acid flow controller must be set between 0.2 and 6 (1 and 30 gpm.).

• Performing small borations at high flow rates may result in Procedure Note: an overboration based on equipment response times.

Boration flow should be set such that the time required to reach the desired setpoint will happen after release of the control switch.

• VERIFY the RMW CONTROL switch has been placed in the STOP position.

• VERIFY the RMW CONTROL switch green light is lit.

OATC

• SET controller 1CS-283, FK-113 BORIC ACID FLOW, for the desired flow rate.

• PLACE control switch RMW MODE SELECTOR to the BOR position.

• Boration may be manually stopped at any time by turning control switch RMW CONTROL to STOP.

Procedure Note:

• During makeup operations following an alternate dilution, approximately 10 gallons of dilution should be expected due to dilution water remaining in the primary makeup lines.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 15 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior

• START the makeup system as follows:

o TURN control switch RMW CONTROL to START momentarily.

o VERIFY the RED indicator light is LIT.

o IF expected system response is not obtained, THEN TURN control switch RMW CONTROL to STOP.

• VERIFY boration automatically terminates when the OATC desired quantity of boron has been added.

• Monitor Tavg and rod control for proper operation.

• Establish VCT pressure between 20-30 psig.

• Turn control switch RMW MODE SELECTOR to AUTO.

• START the makeup system as follows:

o TURN control switch RMW CONTROL to START momentarily.

o VERIFY the RED indicator light is LIT.

The following steps will initiate turbine load reduction IAW Evaluator Note:

GP-006.

INFORMS Load Dispatcher that a load reduction to 95% will SRO begin. (N/A, per Initial Conditions)

Routine load changes must be coordinated with the Load Procedure Note:

Dispatcher to meet system load demands.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 16 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior A failure of the VIDAR in the DEH computer has resulted in a plant trip in the past. This failure would affect operation in Operator Auto, and can be detected as follows:

• If OSI-PI is available, then VIDAR is functioning properly if the 'DEH_MEGAWATTS' point is updating.

• If OSI-PI is not available, then accessing the 'ANALOG INPUTS' screen on the Graphics Display Computer (located in the Termination Cabinet Room near the ATWS Panel) will show several points, most of which should be updating if the VIDAR Unit is functioning properly.

• If the DEH graphics computer is out of service, then Procedure Caution: VIDAR can be checked as updating on the operator panel as follows:

o Depress 'Turbine Program' display button.

o Check 'Turbine Program' display button is illuminated.

o Check 'Reference' and 'Demand' displays indicate

'0000'.

o Depress '1577'.

o Depress 'Enter'.

o If the 'Demand' display indicates '0000', then VIDAR is updating.

o If the 'Demand' display indicates '0001', then VIDAR is not updating There is no procedural guidance directing when the Evaluator Note: boration to lower power is required. The crew may elect to perform the boration prior to placing the Turbine in GO.

DIRECTS BOP to start power reduction at 4 DEH Units/Min.

SRO May direct initiation of a boration before the power reduction begins.

BOP Requests PEER check prior to manipulations of DEH Control Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 17 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior

• DEPRESS the LOAD RATE MW/MIN push-button.

• ENTER the desired rate, NOT to exceed 5 MW/MIN, in the DEMAND display. (4 DEH Units/minute)

• DEPRESS the ENTER push-button.

BOP

• DEPRESS the REF push-button.

• ENTER the desired load (per CRS) in the DEMAND display.

• DEPRESS the ENTER push-button. The HOLD push-button should illuminate.

The unloading of the unit can be stopped at any time by depressing the HOLD push-button. The HOLD lamp will Procedure Note: illuminate and the GO lamp will extinguish. The load reduction can be resumed by depressing the GO push-button. The HOLD lamp will extinguish and the GO lamp will illuminate.

• DEPRESS the GO push-button to start the load reduction.

• VERIFY the number in the REFERENCE display BOP decreases.

• VERIFY Generator load is decreasing.

• Communicate to the SRO that the Turbine is in GO.

• WHEN Turbine load is less than 95%, THEN ensure the 3A and 3B Feedwater Vents have been opened per OP-136, BOP Section 7.2.

This will be directed to a field operator.

Acknowledge directions from BOP to perform OP-136 section 7.2.

Simulator Communicator: (No actions are required by the Simulator Operator)

Wait 2-3 minutes and report completion of task to BOP Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 18 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior Ramp is completed: Power level is approximately 95%

Crew Prepares to secure the A Heater Drain Pump OP-136 OP-136, Feedwater Heaters, Vents, and Drains Section 7.0 Shutdown 7.1 Shutdown of Heater Drain Pumps 7.1.1 Initial Conditions Normally the Heater Drain Pumps are stopped when Reactor Procedure Note:

power is 40 to 45% per GP-006.

1. IF only one Heater Drain Pump is to be stopped, THEN the following conditions should be met:

a. Reactor power is less than 99% to accommodate for the loss of secondary efficiency. (YES)

BOP b. The MW feedback loop is removed from service (YES)

2. IF both Heater Drain pumps are to be stopped, THEN Maintenance has verified that PS-01MS-110 is reset to prevent a turbine runback (N/A)

OP-136 Section 7.1.2 Procedure Steps

• The intent of this section is to establish 4A (B)

Feedwater Heater level control on the Condenser Dump valve before stopping the Heater Drain Pump. This minimizes the level transient when the pump is secured.

• As the Condenser Dump valves starts to control level, Procedure Note:

the HDP discharge level control valve will start to shut and discharge flow will decrease.

• The Main Control Room operator must monitor HDP flow and provide trending information to the operator at the pneumatic alternate level controller.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 19 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior Stopping Heater Drain Pumps at power levels greater than 50% can result in oscillations in heater levels. Heater 4A (4B)

Procedure Caution:

Condenser Dump Controller may need adjustment to stabilize levels.

ERFIS group display or quick plot HDPA has been Evaluator Note:

previously created and is a plot available to the Operators

1. CREATE a plot on ERFIS to monitor Heater Drain Pump BOP discharge flow, discharge pressure and heater level.

(FHD-1255A, PHD1255A and LHD1250A)

2. ESTABLISH communications between the Main Control BOP Room and the operator at 4A pneumatic alternate level controller Simulator Acknowledge directions to establish communications with Communicator: the BOP.

Monitor the FW Heater 4A using simulator drawing FWH02 NOTE: the as-found LC-01HD-1251A(B) pneumatic Simulator Operator: controller setting is also on this display and will be asked for in step 4. Provide the settings value to the Communicator.

3. IF desired, THEN PLACE the 4A (B) Feedwater Heater Sight Glass in service by slowly opening the applicable isolation valves listed below:

• 1HD-293-LI1-2 (1HD-299-LI1-2), LG-01HD-1250A (B)

BOP Instrument Valve.

• 1HD-293-HI1-2 (1HD-299-HI1-2), LG-01HD-1250A (B)

Instrument Valve.

N/A - Not desired

4. PERFORM the following at LC-01HD-1251A (B) :

BOP a. RECORD as-found LC-01HD-1251A (B) pneumatic controller setting in the control room log.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 20 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior (As found setting of LC-01HD-1251A (B) can be found on drawing FWH02)

Simulator Communicator:

Report the as-found LC-01HD-1251A(B) pneumatic controller setting to the BOP.

Actions in Step 7.1.2.4.b cause response being monitored in Procedure Note: Step 7.1.2.4.c.

Step 7.1.2.4.b will cause HDP discharge flow to lower.

b. While monitoring Heater Drain Pump discharge flow, BOP DIRECT the local operator to slowly lower the set point on 4A (B) pneumatic alternate level controller.

Run Trigger 20 - to open the 4A FWH alternate level control Simulator Operator:

valve to lower HDP A discharge flow

c. WHEN Heater Drain Pump discharge flow is less than or BOP equal to 500 kpph, THEN STOP Heater Drain Pump A (B).

DO NOT run Trigger 21: conditionally activates when A HDP Simulator Operator:

control switch is taken to STOP.

d. DIRECT the operator at LC-01HD-1251A (B) to slowly BOP adjust 4A (B) Feedwater Heater level to approximately 2 inches.

e. RECORD as-left LC-01HD-1251A (B) pneumatic controller BOP setting in the control room log.

(As left setting of LC-01HD-1251A (B) can be found on drawing FWH02)

Simulator Communicator:

Report the as-left LC-01HD-1251A(B) pneumatic controller setting to the BOP.

5. IF necessary, THEN REPEAT Steps 7.1.2.1 through 7.1.2.4 BOP for the remaining pump. (N/A)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 21 of 67 Event

Description:

Reduce power to secure A Heater Drain Pump Time Position Applicants Actions or Behavior

6. VERIFY the 4A and 4B Feedwater Heater Sight Glasses are isolated by shutting isolation valves listed below:

• 1HD-293-HI1-2, LG-01HD-1250A Instrument Valve BOP

• 1HD-293-LI1-2, LG-01HD-1250A Instrument Valve

• 1HD-299-HI1-2, LG-01HD-1250B Instrument Valve

• 1HD-299-LI1-2, LG-01HD-1250B Instrument Valve N/A sight glasses were NOT cut in.

BOP Reports to CRS that the A Heater Drain Pump is secured The Lead Evaluator can cue Event 2 (T-ref processer Lead Evaluator: failure low) after the crew has secured the A Heater Drain Pump and the unit is again stable.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 22 of 67 Event

Description:

T-Ref Processor Failure - LOW Time Position Applicants Actions or Behavior On cue from Lead Evaluator insert Trigger 2 Simulator Operator:

T-ref processor failure low

• Uncontrolled rod motion Indications Available

• T ave - T ref MCB digital indication reads T ref at 557°F OATC RESPONDS to uncontrolled rod motion.

ENTERS and directs actions of AOP-001, MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM SRO Makes PA announcement for AOP entry Holds a crew focus brief AOP-001 Malfunction Of Rod Control And Indication System OATC PERFORMS immediate actions.

Immediate CHECK that LESS THAN TWO control rods are dropped.

Action OATC (YES)

Immediate Action OATC POSITION Rod Bank Selector Switch to MAN.

Immediate Action OATC CHECK Control Bank motion STOPPED. (YES)

SRO Conduct a FOCUS BRIEF on entry into AOP-001.

READS immediate actions and proceeds to Section 3.2.

SRO Directs BOP to place Turbine to HOLD if in GO.

BOP Places Turbine to HOLD if in GO.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 23 of 67 Event

Description:

T-Ref Processor Failure - LOW Time Position Applicants Actions or Behavior CHECK that instrument channel failure has NOT OCCURRED by observing the following:

OATC

• RCS Tavg (YES)

• RCS Tref (NO)

PERFORM the following:

• IF a power supply is lost, THEN GO TO AOP-024, Loss of Uninterruptible Power Supply. (NO)

OATC

• IF an individual instrument failed, THEN MAINTAIN manual rod control until corrective action is complete.

• IF a Power Range NI Channel failed, THEN BYPASS the failed channel using OWP-RP. (N/A)

MANUALLY OPERATE affected control bank to restore the following:

• EQUILIBRIUM power and temperature conditions OATC

• RODS above the insertion limits of Tech Spec 3.1.3.6 and PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report.

Determines Tref based on 1st Stage pressure using Curve G-4.

OATC He/she may instead use Tref just before the failure to determine the current value of Tref.

Evaluator Note: The following will be done when Tave is restored.

VERIFY proper operation of the following:

• CVCS demineralizers (YES)

OATC

• BTRS (N/A)

• REACTOR Makeup Control System (YES)

CHECK that this section was entered due to control banks SRO MOVING OUT. (NO)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 24 of 67 Event

Description:

T-Ref Processor Failure - LOW Time Position Applicants Actions or Behavior CHECK that NEITHER of the following OCCURRED:

OATC

• Unexplained RCS boration (YES)

• Unplanned RCS dilution (YES)

CHECK that an automatic Rod Control malfunction SRO OCCURRED. (NO)

MAINTAIN manual rod control until appropriate corrective action is complete.

SRO/

OATC Reviews/prepares OMM-001, Attachment 5 Equipment Problem Checklist.

Contacts support personnel for repairs.

SRO Exits AOP-001 May establish OMM-001 Att 13 limits for Tavg with Rod Control SRO/ in manual when exiting AOP-001.

OATC May discuss Equipment Problem Checklist with WCC and ask for support for the failure.

When Tavg is restored and AOP-001 exited, cue initiation Evaluators Note:

of Event 3 SG B Controlling Level Channel Failure (Low)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 25 of 67 Event

Description:

SG B Controlling Level Channel Failure (Low)

Time Position Applicants Actions or Behavior On cue from Lead Evaluator actuate Trigger 3 Simulator Operator:

SG B Controlling Level Channel Failure (Low)

ALB-14-2-1B SG B NR LVL/SP HI/LO DEV Indications Available ALB-14-5-4B STEAM GEN B LOW-LOW LVL BOP RESPONDS to alarms and ENTERS ALB-014-2-1B and 5-4B.

The APP-ALB-014-2-1B and 14-5-4B actions are similar.

IAW AD-OP-ALL-1000, the operator may take MANUAL Evaluator Note: control of a malfunctioning controller before being directed by a procedure or the SRO. Some or all of the following failure indications may be reported to the SRO.

• CONFIRM alarm using LI-484 SA, LI-485 SB, or LI-486 SA, Steam Generator B level indicators.

o Reports LI-486 reading or failed low.

o FI-486 and FI-487 are rising o Actual NR level is rising

• VERIFY Automatic Functions: NONE BOP

• PERFORM Corrective Actions:

o CHECK Steam Flow (FI-484, FI-485) AND Feed Flow (FI 486, 487) for deviation. (YES) o IF FCV-488, SG B auto level controller, is NOT sufficiently correcting level, THEN: (YES)

SWITCH to MANUAL - Lowers M/A output RESTORE level to normal (57% NR).

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 26 of 67 Event

Description:

SG B Controlling Level Channel Failure (Low)

Time Position Applicants Actions or Behavior Should provide guidance to maintain B SG level to be maintained between 52 to 62%, Trip limit of 30% Low and 73%

High (IAW OMM-001, Attachment 13)

SRO Refer to OWP-RP-06 to remove channel from service.

(See Attachment 1 at end of scenario)

Contacts I&C to have channel removed from service.

Dispatch AO to investigate Failed channel does NOT have to be removed from service Evaluators Note:

to continue the scenario.

Enters Instrumentation TS 3.3.1 Functional Unit 13, 14 Action 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other SRO channels per Specification 4.3.1.1 3.3.2 Functional Unit 5.b, 6.c Action 19 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied :

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.2.1.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 27 of 67 Event

Description:

SG B Controlling Level Channel Failure (Low)

Time Position Applicants Actions or Behavior May request an extra operator for dedicated feedwater operation Direct operator and I&C to perform OWP-RP-06 SRO Reviews/prepares OMM-001, Attachment 5 Equipment Problem Checklist.

Contacts support personnel for repairs.

The actions for OWP-RP-06 are listed in Attachment 1 in Evaluator Note:

the back of this scenario guide on page 58.

Acknowledge request and reports from SRO.

IF an extra operator is requested say that one will be sent Communicator: when available. NO one is available right now.

IF asked to report to MCR to perform OWP-RP-06 state that you will report as soon as possible.

DO NOT RUN APP for failure. Not required to continue Simulator Operator:

with scenario.

After SG level is under control, TSs have been identified and RMU control in AUTO then cue Event 4 (VCT level Channel 115 fails low)

Note: Any Tech Spec evaluation completion can be Evaluator Note:

continued after the scenario is ended.

With RMU control in AUTO an automatic VCT makeup will be generated from VCT level channel 115

(<20% causes auto makeup).

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 28 of 67 Event

Description:

VCT Level Channel 115 Fails Low Time Position Applicants Actions or Behavior On cue from the Lead Evaluator actuate Trigger 4 Simulator Operator:

VCT Level Channel failure low

• Auto Makeup initiates

• ALB-007-4-3, VCT HIGH-LOW LEVEL Indications Available

• ALB-007-5-5, COMPUTER ALARM CHEM & VOL SYSTEMS OATC RESPONDS to alarm ALB-007-4-3.

Crew may immediately secure auto makeup based on SRO AD-OP-ALL-1000 guidance since the makeup is due solely to the instrument failure.

OATC ENTERS and performs APP-ALB-007-4-3.

The SRO is likely to go directly to AOP-003, Evaluator Note: MALFUNCTION OF REACTOR MAKEUP CONTROL, while the OATC references the APP as time allows.

CONFIRM alarm using LI-115-1, Vol Control Tank Level OATC (MCB-1A2).

DETERMINES LT-115 failed LOW.

OATC May also use LI-112 (ERFIS indication) to report what actual VCT level is doing.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 29 of 67 Event

Description:

VCT Level Channel 115 Fails Low Time Position Applicants Actions or Behavior VERIFY Automatic Functions:

• AT 5% VCT level, the following occurs: (N/A) o 1CS-291, Suction from RWST (LCV-115B) opens.

o 1CS-292, Suction from RWST (LCV-115D) opens o 1CS-165, VCT Outlet/Dilution (LCV-115C) shuts OATC o 1CS-166, VCT Outlet/Dilution (LCV-115E) shuts

• AT 20% VCT level, auto makeup from the Reactor Makeup System starts. (YES)

• AT 40% VCT level, auto makeup from the Reactor Makeup System stops. (N/A)

• AT 80% VCT level, 1CS-120, VCT Level Control Vlv, fully diverts letdown flow to the RHT. (N/A)

Procedure Caution: Low VCT level is a precursor to gas binding the CSIPs.

Procedure Note: If either LT-112 or LT-115 fails high, the automatic CSIP suction swapover from the VCT to the RWST will not function if required.

IF EITHER of the following occurs:

• VCT level is less than 20% AND automatic makeup is NOT in progress SRO

• VCT level is greater than 40% AND automatic makeup is still in progress THEN GO TO AOP-003, Malfunction of Reactor Makeup Control.

ENTERS and directs actions of AOP-003, MALFUNCTION OF REACTOR MAKEUP CONTROL SRO Makes PA announcement for AOP entry Holds a crew focus brief AOP-003 Malfunction Of Reactor Makeup Control CREW CHECK instrument air available. (YES)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 30 of 67 Event

Description:

VCT Level Channel 115 Fails Low Time Position Applicants Actions or Behavior OATC CHECK BOTH LT-112 and LT-115 functioning properly. (NO)

SRO GO TO Section 3.1, LT-112 or LT-115 Malfunction.

REFER TO Attachment 1, VCT Level Control Channels SRO Operation, as necessary to assess the effects of an LT-112 or LT-115 malfunction.

An instrument malfunction may manifest itself as a slow drift rather than a full high or full low failure. Until the instrument Procedure Note:

has failed fully high or fully low, all steps should be reviewed for applicability periodically, even if not continuously applicable.

OATC CHECK that LT-115 is FAILING. (YES)

MONITOR VCT level using either of the following:

OATC

• ERFIS point LCS0112

• LI-112 (local)

OATC CHECK LT-115 FAILING LOW. (YES)

PLACE RMW CONTROL Switch in STOP.

OATC (May already have been performed.)

Normally, VCT level is maintained between 20 and 40% by Procedure Note:

auto makeup.

CONTROL VCT level as follows:

OATC

• MAINTAIN level BELOW 70%

• MAINTAIN level ABOVE 20% OR DESIRED MINIMUM OATC MAINTAIN VCT level GREATER THAN 5%.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 31 of 67 Event

Description:

VCT Level Channel 115 Fails Low Time Position Applicants Actions or Behavior Since the following procedure caution does not apply it Evaluator Note:

may not be read .

Lifting leads in the following step will simulate a low-low level Procedure Note: signal from the failed instrument. This is to allow a valid low-low level signal one instrument to initiate emergency makeup.

OATC CHECK the malfunctioning instrument FAILING LOW. (YES)

DIRECT Maintenance to investigate and repair the instrument SRO malfunction.

SRO CHECK that the instrument malfunction has been repaired.

WAIT until repairs are complete before proceeding.

Reviews/prepares OMM-001, Attachment 5 Equipment SRO Problem Checklist.

Contacts support personnel for repairs.

Respond to crew requests.

Communicator: NOTE: Do not run the APP file to lift leads for this event prior to continuing with Event 5. If questioned later report back that you are working on it.

The Lead Evaluator can cue Event 5 - C Accumulator Evaluator Note: nitrogen leak while the crew is waiting for the instrument repairs.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5 Page 32 of 67 Event

Description:

C Accumulator Nitrogen Leak Time Position Applicants Actions or Behavior On cue from Lead Evaluator actuate Trigger 5 Simulator Operator:

C Accumulator nitrogen leak Alarm ALB-01-9-1, ACCUMULATOR TANK C HIGH-LOW Indications Available PRESSURE Responds to alarm ALB-1-9-1, ACCUMULATOR TANK C OATC HIGH-LOW PRESSURE.

OATC Enters APP-ALB-01-9-1 CONFIRM alarm using SI Accumulator pressure indicators OATC PI-929 and PI-931

• No automatic actions associated with this alarm PERFORM Corrective Actions:

OATC

• IF SI Accumulator pressure has risen AND NO rise in level has occurred, THEN . . . (N/A)

N 2 through 1SI-287 is the primary source of motive power to the PRZ PORVs, with Instrument Air as backup. If 1SI-287 Procedure Note: is shut in a mode where LTOPS is required operable, and Instrument Air is not available to PORV accumulators, LTOPS must be declared inoperable.

IF SI Accumulator pressure has risen AND is accompanied by OATC a rise in level, THEN. . . (NO)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5 Page 33 of 67 Event

Description:

C Accumulator Nitrogen Leak Time Position Applicants Actions or Behavior IF SI Accumulator pressure has dropped, THEN: (YES)

• IF pressure drop is accompanied by a drop in level, THEN: (NO)

• STOP any accumulator draining in progress (NONE)

OATC

• MONITOR leakage into CNMT Sump. (NONE)

• DISPATCH an operator to CNMT to locate and isolate leakage as soon as possible.

• REFER TO OP-110, Safety Injection AND Raise Accumulator pressure.

Directs OATC to pressurize C Accumulator IAW OP-110, Safety Injection to maintain pressure within Tech Spec range SRO Reviews/prepares OMM-001, Attachment 5 Equipment Problem Checklist.

Contacts support personnel for repairs.

The following TS must be entered if Accumulator pressure Evaluator Note: lowers to less than 585 PSIG and/or when it is connected to the N 2 System.

Refer to Technical Specification 3.5.1.d Action a - With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration not within SRO limits, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5 Page 34 of 67 Event

Description:

C Accumulator Nitrogen Leak Time Position Applicants Actions or Behavior Enters OP-110, Section 8.2 - Pressurizing the SI OATC Accumulators.

Verifies Initial Conditions

• The Nitrogen System is available per OP-152.01

• Accumulator level is greater than 66%

OATC

• If the Accumulators are depressurized, the Accumulator metal temperature must be greater than 70°F before pressurization. (Contact pyrometer can be used or containment ambient temperature) [N/A]

To minimize any potential sluicing between Accumulators through leaking valves, Accumulator pressures should be Procedure Note:

approximately equal (within 4 psid between lowest and highest ERFIS indications) at the completion of this Section.

Perform the following Steps on only one Accumulator at a time.

At the MCB, verify open 1SI-287, ACCUMULATORS & PRZ PORV N2 SUPPLY.

OATC Declare the associated Accumulator inoperable per Tech Spec 3.5.1, due to being connected to Non-Safety piping (a one hour action statement in Modes 1 through 3 above 1000 psig).

To prevent exceeding the capacity of the N2 System and Procedure Note: maintain train separation for the Accumulators, only one Accumulator should be pressurized at a time.

At the MCB, open the ACCUM N2 SUPPLY/VENT for the OATC Accumulator to be pressurized: 1SI-297 for ACCUMULATOR C N2 Supply & Vent.

The Accumulator should not be pressurized to the upper Procedure Note: Technical Specification limit (665 psig) to allow for thermal expansion of the Accumulator gas during plant heatup.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5 Page 35 of 67 Event

Description:

C Accumulator Nitrogen Leak Time Position Applicants Actions or Behavior Pressurize the Accumulator to the desired pressure indicated OATC by the associated pressure indicators: PI-929, 931, ACCUMULATOR TK C PRESS.

At the MCB, shut the ACCUMULATOR N2 SUPPLY & VENT OATC valve for the Accumulator that was pressurized: 1SI-297 for ACCUMULATOR C N2 Supply & Vent.

OATC Complete Attachment 6.

The actions for OP-110, Attachment 6 are listed in Evaluator Note: Attachment 2 in the back of this scenario guide on page 63.

IF the Accumulator parameters are within the Tech Spec requirements, THEN DECLARE the Accumulator that was pressurized operable.

OATC (May not declare operable based on leak rate).

Informs SRO that the C Accumulator pressure is within Tech Spec requirements and the C Accumulator can be considered operable.

SRO Acknowledges OATC information The Lead Evaluator can cue Event 6 - EHC Pump A shaft shear while the crew is performing Attachment 6 since it Evaluator Note:

will take several minutes for DEH system pressure to lower to the alarm setpoint.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 6 Page 36 of 67 Event

Description:

EHC Pump A Shaft Shear / Standby Pump auto start failure Time Position Applicants Actions or Behavior On cue from Lead Evaluator actuate Trigger-6 EHC Pump A shaft shear with standby auto start failure Simulator Operator: NOTE: This event will take a few minutes prior to the low pressure alarm since there are accumulators in the system holding the pressure up.

Indications Available ALB-20-4-2B, EH FLUID LOW PRESS Responds to ALB-20-4-2B or indication of degrading EHC BOP pressure on PI-4221.

BOP Enters APP-ALB-20-4-2B.

BOP Confirms alarm using PI-4221.

VERIFY Automatic Functions:

BOP

• Standby DEH Pump starts at 1500 psig, as sensed by PS-01TA-4223V.

The BOP may immediately start the standby pump or wait until after reading the APP.

Evaluator Note:

• IF the standby pump is not started EHC pressure will continue to lower resulting in a Turbine / Reactor trip when system pressure reduces to < 1150 psig. The resultant Reactor trip will introduce a new critical task to the scenario. (see NOTE on Critical Task Justifications)

When dispatched to investigate, wait ~1 minute then report Communicator:

there is a shaft shear on the A EHC Pump Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 6 Page 37 of 67 Event

Description:

EHC Pump A Shaft Shear / Standby Pump auto start failure Time Position Applicants Actions or Behavior PERFORM Corrective Actions:

a. IF the Reactor is tripped, THEN GO TO EOP-E-0. (NO)

b. START the standby DEH Pump. (YES)

c. IF EH Fluid pressure drops to 1500 psig, (NO)

d. DISPATCH an operator to perform the following:

1) MONITOR DEH Pump and PCV operation.

BOP 2) VERIFY OPEN the following, a) 1EH-1, A EH Pump Suction Vlv b) 1EH-8, B EH Pump Suction Vlv c) 1EH-31, Main Hdr Press Switch Isol Vlv

3) INVESTIGATE system for leaks.

4) IF a leak is found, THEN ISOLATE the leak AND IMMEDIATELY NOTIFY Control Room.

BOP may suggest or the SRO may direct the BOP to place SRO EHC Pump A control switch to PULL-TO-LOCK.

(IF directed) locates EHC Pump A control switch and turns BOP switch to the left then pulls switch up to place in Pull To Lock.

Reviews/prepares OMM-001, Attachment 5 Equipment SRO / Problem Checklist.

OATC Contacts support personnel for repairs.

(Review the information on the following page prior to Evaluator Note:

actuation of Event 7)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 6 Page 38 of 67 Event

Description:

EHC Pump A Shaft Shear / Standby Pump auto start failure Time Position Applicants Actions or Behavior After Trigger 7 is actuated (A MFW Pump Trip) the following will occur:

• After a 2 minute time delay the B MFW pump will also trip

• The Main Turbine will NOT receive an AUTO TRIP signal from Reactor Protection and must be tripped manually by the BOP during the EOP E-0 immediate action response.

Evaluator Note:

• A MD AFW pump will also trip immediately upon starting (either from an auto or manual start signal).

When the crew is working through EOP-ES-0.1, Reactor Trip Response, the Lead Evaluator will direct the Simulator Operator to actuate Trigger 10 which will trip the TD AFW pump.

The combination of Narrow Range SG levels in ALL SGs

< 25% and Total Feed Flow to SG < 210 KPPH will cause the CSFST for Heat Sink to change to RED. The crew will immediately transition to FR-H.1 on the RED path condition.

When desired or after SRO contacts support personnel Evaluator Note:

then inform the Simulator Operator to insert Event 7.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 39 of 67 Event

Description:

Main Feedwater Pump A Trip Time Position Applicants Actions or Behavior On cue from Lead Evaluator actuate Trigger 7 Simulator Operator:

Trip of A MFW Pump ALB-016-1-4, FW Pump A/B O/C Trip -Gnd or Bkr Fail to Indications Available Close Multiple FW flow alarms BOP RESPOND to multiple alarms/indications.

BOP REPORTS MFW Pump A tripped.

AOP-010 Feedwater Malfunctions Identifies entry conditions met for AOP-010, Feedwater Crew Malfunctions.

Immediate Action BOP CHECK ANY Main Feedwater Pump TRIPPED. (YES)

Immediate CHECK initial Reactor power less than 90%. (NO)

Action OATC TRIP the Reactor and GO TO EOP E-0 Immediate Action OATC INITIATES a MANUAL Reactor Trip.

IF contacted by MCR to investigate the causes of the A and later the B MFW pump trip report that both breakers have tripped on overcurrent. There are no signs of damage at the pumps.

Communicator:

WHEN / IF WCC is contacted then report that Electrical Maintenance is investigating the problems with the breakers any repairs will be made as quickly as possible.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 9 Page 40 of 67 Event

Description:

B MD AFW Pump Trip Time Position Applicants Actions or Behavior EOP-E-0 Reactor Trip or Safety Injection SRO ENTERS EOP-E-0 Verify Reactor Trip:

Reactor Trip Confirmation OATC

• Reactor Trip AND Bypass BKRs - OPEN (YES)

• Rod Bottom lights LIT (YES)

• Neutron flux dropping (YES)

Check Turbine Trip - ALL THROTTLE VALVES SHUT

• All turbine throttle valves - SHUT (NO)

RNO - Manually trip turbine from MCB

• Locates Turbine Manual Trip switch and TRIPS Turbine Check for any of the following:

Critical BOP

• All turbine throttle valves - SHUT (YES)

Task #1 (TSLB-2 OR DEH PANEL B)

OR All turbine governor valves - SHUT (DEH PANEL B)

Critical to manually trip Turbine from MCB prior to the receipt of the Low Steam Line Pressure SI signal Perform the following:

BOP a. AC emergency buses - AT LEAST ONE ENERGIZED (YES - Both 1A-SA and 1B-SB from Off-Site)

OATC SI - ACTUATED (BOTH TRAINS) (NO)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 9 Page 41 of 67 Event

Description:

B MD AFW Pump Trip Time Position Applicants Actions or Behavior SI actuation - REQUIRED (NO)

OATC The B Main Feedwater Pump will trip 2 minutes after the A Main Feedwater Pump trip occurred. The crew should identify the trip by the following annunciator:

ALB-016-1-4, FW Pump A/B O/C Trip-Gnd or Dkr Fail to Evaluator Note: Close ALB-016-2-2, Loss of BOTH Main FW Pumps AND the B MD AFW pump will trip when it starts ALB-017-6-4, Aux Feedwater Pump B Trip or Close Ckt Trouble IF contacted to investigate the cause of the B AFW pump trip report the breaker is tripped on overcurrent. No signs of damage at the pumps.

WHEN / IF WCC is contacted report that Electrical Communicator:

Maintenance is investigating the breaker and that repairs will be made as quickly as possible.

IF asked about the A MD AFW pump status report that it is still waiting on parts to complete the motor overhaul.

EOP-Reactor Trip Response ES-0.1 Conduct a FOCUS BRIEF on entry into EOP-ES-0.1.

SRO GO TO EOP-ES-0.1, Reactor Trip Response Assigns EOP-ES-0.1 Foldout Criteria SRO IMPLEMENT Function Restoration Procedures as required.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 9 Page 42 of 67 Event

Description:

B MD AFW Pump Trip Time Position Applicants Actions or Behavior SRO Informs SM to EVALUATE EAL Matrix CHECK RCS Temperature:

BOP a. Check RCPs - ANY RUNNING (YES)

b. CHECK SG Blowdown isolation valves - SHUT. (YES)

SG (MLB-1A-SA) (MLB-1B-SB)

A 1BD-11 1BD-1 B 1BD-30 1BD-20 C 1BD-49 1BD-39 STABILIZE AND maintain temperature between 555°F AND 559°F using Table 1.

(RCS Temp trend is > 557°F and rising - middle column)

BOP Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 9 Page 43 of 67 Event

Description:

B MD AFW Pump Trip Time Position Applicants Actions or Behavior Crew CHECK Feed System Status:

OATC

• RCS Temperature - LESS THAN 564°F. (YES)

BOP

• VERIFY feed reg valves - SHUT. (YES)

• CHECK feed flow to SGs - GREATER THAN 210 KPPH.

BOP (YES - may report: Only the Turbine Driven AFW Pump is running)

CHECK Control Rod Status:

OATC

• CHECK DRPI - AVAILABLE. (YES)

• VERIFY all control rods - FULL INSERTED. (YES)

OATC CHECK PRZ level - GREATER THAN 17%. (YES)

Lead Evaluator: Cue Simulator Operator to insert Trigger 10 Trip of the Turbine Driven AFW pump On cue from Lead Evaluator actuate Trigger 10 Simulator Operator:

Trip of the TD AFW pump CREW Contacts AOs to investigate failures During the remainder of the scenario any communications for a request to restore MFW or AFW -

Communicator Maintenance is looking at the situation and will make repairs as soon as they can.

When ANY pump is available the WCC will contact the MCR.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 9 Page 44 of 67 Event

Description:

B MD AFW Pump Trip Time Position Applicants Actions or Behavior

• Identifies TD AFW Pump has tripped Crew Identifies RED PATH on CSFST for Heat Sink when:

• NR Level in ALL SGs < 25%

• Total Feed Flow to SGs < 210 KPPH Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 45 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior EOP-RESPONSE TO LOSS OF SECONDARY HEAT SINK FR-H.1 TRANSITIONS to EOP-FR-H.1 after verification of RED Path.

SRO Conduct a FOCUS BRIEF on entry into EOP-FR-H.1.

SRO Reads Caution prior to step 1 The following procedure caution does not apply and Evaluator Note:

therefore may be summarized.

CAUTION

• This procedure should NOT be performed if total feed flow capability of 210 KPPH is available AND total feed flow has been reduced due to operator action as directed by the EOPs. (The following EOPs direct feed flow to be reduced below 210 KPPH:

ECA-2.1, "UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS" FR-S.1, "RESPONSE TO NUCLEAR POWER GENERATION/ATWS" FR-P.1, "RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK" FR-P.2, "RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK" FR-Z.1, "RESPONSE TO HIGH CONTAINMENT PRESSURE")

• Feed flow should NOT be established to any faulted SG while a non-faulted SG is available.

PERFORM the following:

SRO Initiate monitoring of Critical Safety Function Status Trees.

Evaluate EAL Matrix - contacts SM to EVALUATE EAL Matrix (Refer to PEP-110)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 46 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior CHECK Secondary Heat Sink Requirements:

• RCS pressure - > ANY NON-FAULTED SG PRESS (YES)

OATC

• RCS temperature - GREATER THAN 350°F [330°F]. (YES)

• STOP any running RHR pumps. (YES)

Check If Bleed And Feed Is Required:

SRO SG WR levels - ANY TWO LESS THAN 15% [30%] (NO)

GO TO Step 4 NOTE: Foldout applies SRO SRO Assigns EOP-FR-H.1 Foldout Criteria Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 47 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior CHECK SG Blowdown and SG Sample Valves shut (YES)

BOP ESTABLISH AFW Flow to at least ONE SG:

• OBSERVE MCB indications to determine cause of AFW failure:

o CST level (NO) o MDAFW pump power supplies (YES) o TDAFW pump steam supply valves (YES) o TDAFW pump speed controller (NO)

BOP/SRO o TDAFW pump control power (NO) o AFW valve alignment (NO)

• TRY to restore AFW flow at the MCB.

(Refer to EOP-FR-H.1 Attachment 1 for guidance of rate of feed flow.)

(Refer to OP-137, Auxiliary Feedwater System, for guidance regarding AFW pump operations, precautions and limitations and valve operation.)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 48 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Check IF AFW Flow Established:

• TOTAL feed flow to SGs - GREATER THAN 210 KPPH.

BOP (NO)

GO TO Step 6.c PERFORM the following:

BOP/SRO

• CONTINUE attempts to restore AFW flow at the MCB and locally.

SRO OBSERVE NOTE prior to Step 7 AND continue with Step 7.

After stopping all RCPs and placing steam dump in the steam pressure mode, RCS pressure and temperature will increase Procedure Note: as natural circulation is established. A large loop T prior to PRZ PORV opening confirms natural circulation. This must be considered while evaluating bleed and feed criteria.

STOP Heat Input from RCP Operations:

OATC

• Stops ALL RCPs CHECK all of the following to determine if steam can be dumped to condenser:

• CHECK any intact SG MSIV - OPEN. (YES)

BOP

• CHECK Condenser Available (C-9) light (BPLB 3-3) - LIT.

(YES)

• STEAM dump control system - AVAILABLE. (YES)

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 49 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Evaluators Note: The following three substeps may have already been performed per Table 4.

• PLACE Steam Dump pressure controller in Manual AND decrease output to 0%.

• PLACE Steam Dump mode select switch in STEAM BOP PRESS.

• ADJUST Steam Dump controller setpoint to 84%

(1092 PSIG) AND place in AUTO.

The crew should NOT attempt to start either MFW Pump until the reasons for the original trips are known and Evaluator Note: corrected.

The crew may answer POWER to at least ONE main FW pump - AVAILABLE as YES but they will still end up being directed to continue OATC Check SI actuated (NO) GO TO Step 10 ESTABLISH Main FW Flow to at least ONE SG:

• CHECK condensate system - IN SERVICE. (YES)

• SUPPORT condition for FW startup - AVAILABLE.

BOP (YES)

• POWER to at least ONE main FW pump - AVAILABLE.

(YES - There is power but both MFW Pump breakers are tripped)

SRO WHEN support conditions met, THEN do Step s 10.c and 10.d.

SRO OBSERVE CAUTION prior to Step 12 and GO TO Step 12.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 50 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Following block of automatic SI actuation, manual SI actuation may be required if conditions degrade. (Examples of degraded Procedure Caution:

conditions are the inability to maintain or restore PRZ level, RVLIS indication or RCS subcooling.)

• After the low steam pressure SI signal is blocked, main steam line isolation will occur if the high steam pressure rate setpoint is exceeded.

Procedure Note:

• The Main FW pump discharge valve control switches must be held in the OPEN position to open the valves with the main FW pumps stopped.

DEPRESSURIZE RCS and block Low Steam Pressure SI:

Open the following valves while continuing with this procedure:

• Low pressure FW heater bypass valves:

o 1CE-330 BOP o 1CE-359

• High pressure FW heater bypass valve:

o 1FW-110

• Main FW pump discharge valves:

o 1FW-29 o 1FW-60

• CHECK SI - IN SERVICE. (NO)

OATC o RNO - go to step 12f DEPRESSURIZE RCS to between 1900 PSIG AND 1950 PSIG

• CHECK letdown - IN SERVICE (YES)

• DEPRESSURIZE using auxiliary spray (refer to OP-107)

OATC BLOCK SI Signals:

• Low PRZ pressure

• Low steam pressure MAINTAIN pressure less than 1950 PSIG Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 51 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Auxiliary Spray IAW OP-107 Section 8.13:

VERIFY SHUT Normal Pressurizer Spray valves:

• 1RC-107, PRZ SPRAY LOOP A PCV 444 C

• 1RC-103, PRZ SPRAY LOOP B PCV 444 D OPEN 1CS-487, PRESSURIZER AUX SPRAY Monitor ERFIS point TRC0450, PZR SURGE LINE TEMP, AND MINIMIZE swings greater than 100oF OATC VERIFY SHUT 1CS-480, ALTERNATE CHARGING PLACE FK-122, CHARGING FLOW 1CS-231, in Manual Maintain Pressurize pressure as steady as possible, using any of the following, as needed:

• ADJUST FK-122 as needed to control Pressurizer Pressure

• Energize and control PRZ heaters

• OPEN or SHUT 1CS-492, NORMAL CHARGING.

RCS pressure will need to be monitored or it will continue to decrease with AUX spray until noticed. This may also Evaluators Note:

result in letdown isolation. If VCT level drops <5%, CSIP suction will swap the RWST.

• Depressurizing only one SG minimizes the likelihood of reaching the bleed and feed criteria (due to lowering SG level) AND the likelihood of the appearance of degraded plant conditions that might require manual SI actuation.

Procedure Notes:

• The preferred SG to depressurize is the intact SG with the highest indicated wide range level.

• A second SG may be depressurized if condensate flow cannot be established to the first SG depressurized.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 52 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior DEPRESSURIZE One SG To Less Than 500 PSIG AND ESTABLISH Condensate Flow:

• IDENTIFY the SG to be depressurized.

• SHUT the following valves for the SGs that are NOT to be depressurized.

• MSIVs (A & B, 1MS-80 & 1MS-82)

BOP

• MSIV bypass valves

• SG main Steam drain isolation before MSIV (A & B, 1MS-231 & 1MS-266)

• DUMP steam at maximum rate to depressurize identified to SG to 500 PSIG using any of the following (listed in order of preference):

• Condenser steam dump ESTABLISH condensate flow using Attachment 3.

Alignment actions will commence while the SG is being Evaluators Note:

depressurized.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 53 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior EOP-FR-H.1 Attachment 3

• This attachment provides instructions for establishing condensate flow to one SG to restore secondary heat sink.

It may also be used as a reference for establishing condensate flow to SGs while implementing other EOPs.

• After the low steam pressure SI blocked, main steam line Procedure Note: isolation will occur if the high steam pressure rate setpoint is exceeded.

• If an action or its contingency in this attachment can NOT be accomplished, the operator should return to the step in effect, while continuing efforts to establish condensate flow.

CHECK Primary and Secondary Conditions To Allow Establishing Condensate Flow:

• CHECK low steam SG pressure SI - BLOCKED (YES)

BOP

• CHECK SG pressure for SG to which condensate flow is to be established - LESS THAN 500 (NO)

• GO To Step 2.

The preferred SG to depressurize is the intact SG with the Procedure Note:

highest indicated wide range level.

Main Steam isolation may actuate during this action. This actuation will shut the MSIVs and the Steam Dumps will Evaluators Note:

no longer function. If so, the depressurization should be continued using the respective SG PORV.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 54 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Depressurize One SG To Less Than 500 PSIG:

• Identify the SG to be depressurized.

• Shut the following valves for the SGs that are NOT to be depressurized.

o MSIVs o MSIV bypass valves o SG main steam drain isolations before MSIV:

SG A: 1MS-231 SG B: 1MS-266 SG C: 1MS-301

• Dump steam at maximum rate to depressurize identified SG BOP to 500 PSIG using any of the following (listed in order of preference):

o Condenser steam dump o SG PORVs o Locally operate SG PORVs using OP-126, "MAIN STEAM, EXTRACTION STEAM, AND STEAM DUMP SYSTEMS", Section 8.2.

o TDAFW pump CHECK Condensate System Status:

• At least one condensate - RUNNING (YES)

• At least one condensate booster pump - RUNNING (YES)

The MAX rate depressurization may cause the MSIVs to auto close on Steam Pressure Rate. IF this occurs the Evaluator Note:

crew should continue the depressurization using the SG PORVs.

The Main FW pump discharge valve control switches must be Procedure Note: held in the OPEN position to open the valves with the Main FW pumps stopped.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 55 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Verify the following valves - OPEN:

• Low pressure FW heater bypass valves:

BOP (1CE-330/1CE-359)

• High pressure FW heater bypass valves: (1FW-110)

• Main FW pump discharge valves: (1FW-29/1FW-60)

OATC RESET SI. (Not active)

Manually Realign Safeguards Equipment Following A Loss Of CREW Offsite Power. (Refer to E-0 Attachment 6.) (NA)

RESET FW Isolation. (NOT ACTIVE)

PLACE Feed Reg Bypass Controllers In Manual AND Set Output To Zero.

RESET AND open main FW isolation valve(s): (All open already)

BOP

• 1FW-159 (A SG)

• 1FW-277 (B SG)

• 1FW-217 (C SG)

SHUT Main FW Pump Recirc Valves: (1FW-8/1FW-39)

(Already SHUT)

Condensate Booster Pumps will trip on high discharge Procedure Caution: pressure of 625 psig (180 second time delay). This will result in delayed heat sink recovery.

PLACE Condensate Booster Pump Controllers In Manual AND BOP Control Discharge Pressure At 600 PSIG.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 56 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Local checks for flow noise may be used to confirm the Procedure Note:

presence of flow.

Dispatch Aux Operator to FRV Bypass valves to listen for flow Crew noise then report once flow noise is heard.

Simulator When directed to go to desired FRV Bypass valve, wait ~1 Communicator: minute, then report that you are standing by as requested.

To monitor for flow on simulator observe flow indicators in the CFW drawing or open the monitored parameter file -

Plant Status Monitor CFW and check the status of flow Simulator Operator:

using:

• Line 21: wcfw479(1) FRV Bypass Valve A flow

• Line 22: wcfw479(2) FRV Bypass Valve B flow

• Line 23: wcfw479(3) FRV Bypass Valve C flow ESTABLISH Feed Flow to SG(s):

BOP (Refer to Attachment 1 while performing actions that restore feed flow.)

ESTABLISH feed flow using the feed reg bypass valves from the MCB.

Critical Task #2 BOP Feed flow to at least one SG -ESTABLISHED Critical to establish Feedwater flow into at least one SG before RCS feed and bleed is required Simulator Report back to MCR when flow is observed through the Communicator: FRV Bypass valves that are being monitored.

BOP Acknowledges FRV Bypass flow noise and updates crew Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 10 Page 57 of 67 TD AFW Pump Trip Event

Description:

LOSS OF HEAT SINK Time Position Applicants Actions or Behavior Terminate the scenario after feed flow has been established and Wide Range level increase or Feed Water Flow can be identified in at least one SG.

Lead Evaluator:

Instruct the Simulator Operator to place the Simulator in FREEZE and announce I have the shift, remain in the Simulator and dont discuss the scenario. There may be follow up questions asked prior to your release.

Simulator Operator: When directed by the Lead Evaluator go to FREEZE.

Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OWP-RP-06 Reactor Protection B SG Level Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OWP-RP-06 Reactor Protection B SG Level Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OWP-RP-06 Reactor Protection B SG Level Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OWP-RP-06 Reactor Protection B SG Level Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OWP-RP-06 Reactor Protection B SG Level Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OP-110 Attachment 6 Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 OP-110 Attachment 6 Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 EOP-FR-H.1 Attachment 1 Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Operator Action Form ES-D-2 EOP-FR-H.1 Attachment 1 Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # N/A Page 67 of 67 Event

Description:

Scenario Guide Revision Summary Rev. 0 Initial Development Rev. 1 NRC D-1 Outline comments incorporated Rev. 2 Operation validation comments incorporated Rev. 3 NRC 60 day submittal comments incorporated Rev. 4 NRC Prep Week comments incorporated Rev. Final Approved for administration by NRC Region II Harris 2016 NRC Scenario 1 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 Facility: Harris Nuclear Plant Scenario No.: 2 Op Test No.: 05000400/2016301 Examiners: Operators: SRO:

OATC:

BOP:

Initial Conditions: IC-32, 28% power, MOL

• The unit is operating at 28% power

• The following equipment is under clearance o B MD AFW Pump o B NSW pump o Boric Acid Transfer Pump B-SB

• Plant is at 28% power, MOL with start up in progress. GP-005 step 127.

• Turbine is in HOLD with a rate of 2 DEH units per minute set Turnover:

• Raise Turbine ramp rate to 4 DEH units per minute then continue raise load to 100%.

• Plant risk condition is YELLOW due to startup.

• Control AFW flow to minimize RCS cooldown Critical Tasks:

• Close MSIVs for SG A and C Event Malf. No. Event Type* Event Description No.

1 N - BOP N/A Power escalation to 100% (GP-005)

R - RO I - RO/SRO 2 NIS07E Power Range N-44 fails HI (AOP-001)

TS - SRO C - BOP/SRO 3 XD1I121 TS - SRO Containment Fan Cooler Fan (AH-2 A-SA) Trips 4 C - RO/SRO CVC29A CSIP A Shaft Shear (AOP-018)

TS - SRO 5 eps12 C - BOP/SRO Total Loss of Cooling Banks on the UAT 1A (AOP-039) 6 MSS11 M - All Main Steam Line Break outside of Containment (E-0 to E-2) 7 ZDSQ2:52A I - RO/SRO A RHR Pump fails to start from sequencer MSS05A MSIVs fail to close and can't be closed from control room (ECA-2.1) 8 MSS05B C - BOP/SRO Valves close when Instrument Air is isolated locally (return to E-2)

MSS05C ZRPK616A SG B AFW isolation valve fails to close on Feedwater Isolation 9 I - BOP/SRO ZRPK616B signal

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 Scenario Summary:

The unit is currently operating at ~28% power, MOL. Normal startup is in progress IAW with GP-005, Power Operation (Mode 2 to Mode 1) section 5.0 step 127 following a Reactor trip 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ago. The Reactor went critical 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> ago. Ramp load rate is set at 2 DEH units per minute. DEH control panel load is set at 960, and the Turbine is in hold. The load dispatcher requests that when the Turbine ramp is resumed that the ramp rate be increased to 4 DEH units per minute.

The following equipment is under clearance: B MDAFW Pump, B NSW Pump and B Boric Acid Pump

• B MDAFW Pump is under clearance for pump packing repairs. The pump has been inoperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and will be restored to operable status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Tech Spec 3.7.1.2 LCO Action a and Tech Spec 3.3.3.5.b Action c applies. 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO or HSB within the next 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, HSD following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Equipment is under clearance continued:

B MDAFW Pump - (continued) - Tech Spec 3.3.3.5.b Action c B NSW Pump under clearance for shaft inspection. The pump has been under clearance for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Inspection and return to service are expected to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Equipment is under clearance continued:

Boric Acid Transfer Pump B-SB is under clearance due to breaker blown control power fuses. Has been under clearance for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The problem with the breaker has been repaired and the clearance will be removed later this shift. Tech Spec 3.3.3.5.b which is a 7 day LCO and 3.1.2.2 applies (3.1.2.2 is for tracking only). OWP-CS-05 has been completed.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 1 The crew performs a power escalation IAW GP-005. The previous crew were increasing Turbine load at 2 DEH units per minute. The load dispatcher has requested that when the power increase continues that the ramp rate is increased to 4 DEH units per minute.

Verifiable actions: For this reactivity manipulation, it is expected that the SRO will conduct a reactivity brief. The RO will dilute the RCS as necessary per the reactivity plan and monitor auto operation of rod control to raise Reactor power. The BOP will operate the DEH Main Turbine controls to first increase the Turbine ramp rate from 2 DEH units to 4 DEH units per minute then ensure the controls are set correctly prior to ramping the Turbine up to full power.

Event 2 Power Range Channel N-44 fails high. After the crew has placed the B train of Containment Fan Coolers in service PR channel N-44 will fail high. This will cause rods to start stepping in at max speed. The crew should identify that a Power Range channel failure is the cause for the automatic rod motion and enter AOP-001, Malfunction of Rod Control and Indication System.

Verifiable actions: The OATC will perform the immediate actions of AOP-001 by verifying that <2 rods are dropped (no rods have dropped), place Rod Control in MANUAL and then verify no rod motion. With concurrence from the SRO the OATC will restore Tave to pre-failure conditions by withdrawing the rods in manual.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 2 (continued)

Tech Spec Evaluation: Tech. Spec 3.3.1 for any impact due to the failure a Power Range instrument.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 2 (continued)

Tech Spec Evaluation: Tech. Spec 3.3.1 (continued)

Event 3 Containment Fan Cooler Fan AH-2 A-SA trips. Resulting in annunciator ALB 027-7-2, CONTAINMENT FAN COOLERS AH-2 LOW FLOW - O/L actuating.

Verifiable actions: The BOP will establish the B train of Containment Fan Coolers IAW OP-169, Containment Cooling and Ventilation.

Tech Spec Evaluation: Tech Spec 3.6.2.3 Action a, Containment Systems -

Containment Cooling System.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 3 (continued)

Event 4 CSIP A shaft shear will be inserted after the Unit Aux Transformer corrective actions have been completed. The failure will result in multiple alarms on ALB-006 and ALB-007 associated with the loss of charging flow. The crew will stop the A CSIP then enter AOP-018, RCP Abnormal Conditions in order to address the loss of seal injection.

Verifiable actions: The RO will STOP the A CSIP and isolate letdown by shutting letdown orifices 1CS-7 and 1CS-8. AOP-018 immediate actions for NO running CSIP is to isolate Letdown. The ASI (Alternate Seal Injection) pump will start during this event (2 minute and 45 seconds after the seal injection low flow occurs). The RO will have to manually align valves to prepare to start the B CSIP then start the B CSIP. The RO will then open HC-186.1, RCP Seal WTR INJ flow valve and direct an AO to STOP the ASI pump.

Prior to restoring letdown IAW OP-107, Chemical and Volume Control System the Major event will be inserted.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 4 Tech Spec Evaluation: Tech Spec 3.1.2.4 At least two charging/safety injection pumps shall be OPERABLE.

Tech Spec Evaluation: Tech Spec 3.5.2 Two independent ECCS subsystems SHALL be OPERABLE with each comprised of: One OPERABLE charging/safety injection pumps.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 5 Total Loss of Cooling Banks on the UAT 1A. This failure will require the crew to implement AOP-039, Startup and Unit Auxiliary Transformer Trouble entry (no immediate actions). The BOP should review annunciator ALB-022-3-1, Unit Aux Xfer-A Trouble response and dispatch an operator to locally investigate the cause of the alarm.

The report from the AO will be that all transformer cooling is lost. The crew should recognize this meets the entry conditions of AOP-039, Startup And Unit Auxiliary Transformer Trouble. The crew should monitor EFRIS to determine the UAT 1A temperature is approaching the limit that will require the transformer to be unloaded within 30 minutes.

Verifiable actions: The CRS should direct the BOP to transfer the house loads from the UAT 1A to the SUT 1A using OP-156.02, AC Electrical Distribution. The BOP will transfer loads and the SRO will prepare OMM-001, Operations Administrative Requirements, Attachment 5 Equipment Problem Checklist for the failure.

Event 6 - MAJOR - Main Steam Line Break outside of Containment Main Steam Line Break outside of Containment Verifiable Actions: The crew will trip the Reactor, initiate SI and enter EOP E-0, Reactor Trip or Safety Injection.

Major Event: The crew will transition from E-0 to E-2 to EOP-ECA-2.1, Uncontrolled Depressurization of All Steam Generators. ECA-2.1 will also direct shutting air to RAB 261. ECA-2.1 will direct that feed flow is lowered to 12.5 KPPH to each SG. This will cause a RED PATH for FR-H.1 and require the crew to transition into the Loss Of Heat Sink procedure. Since the low flow (< 210 KPPH) was initiated by the operators and a flow rate of > 210 KPPH is available the crew will transition back into ECA-2.1. After returning into ECA-2.1 a cue will be provided to the Simulator Operator to shut the A and C MSIVs. This action will simulate that the air isolation was successful and 2 of the 3 MSIVs shut. Based on ECA-2.1 foldout back to E-2 IF any SG pressure rises at any time, THEN GO TO E-2, Step 1.

With 2 MSIVs shut the crew will observe a pressure rise in the 2 SGs and return to E-2.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

Event 7 The B RHR pump will trip immediately after it starts and cannot be restarted. The A RHR pump will fail to start automatically from the A Sequencer.

Verifiable Actions: The RO will start the A RHR pump while the crew is implementing E-0 (step 7 - Verify RHR Pumps - ALL RUNNING).

Event 8 Automatic Main Steam Isolation does not occur.

Verifiable Actions: The BOP will attempt to perform a manual Main Steam Isolation by actuating the Main Steam Line Isolation switch. The MSIVs will NOT close from this actuation. The BOP will then attempt to close the MSIVs using the individual switches on the MCB. Again, the MSIVs will fail to close. After the crew continues in E-0 they will transition to EOP E-2, Faulted Steam Generator Isolation, they dispatch an AO to locally shut the instrument air supply to RAB 261 (this will isolate air to the MSIVs in an effort to shut them when auto and manual isolation does not work from the MCB). This will not initially be successful.

Event 9 The AFW Auto Isolation for the B SG will not occur.

Verifiable Actions: The BOP will manually isolate AFW flow to the B Steam Generator from the MDAFW Pump and the TDAFW Pump by shutting the associated isolation valves.

The scenario will be terminated at the lead Evaluators discretion following transition from EOP-ECA-2.1 back to EOP-E-2.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 2 (continued)

CRITICAL TASK JUSTIFICATION:

1. Control AFW flow to minimize RCS cooldown - prior to a severe (ORANGE path) challenge develops to the Integrity Critical Safety Function The crew must reduce and control the AFW flow rate to a maximum of 12.5 KPPH when directed by EOP ECA-2.1 when all SGs are faulted.

Failure to control the AFW flow rate to the SGs leads to an unnecessary and avoidable severe challenge to the integrity CSF. Also, failure to perform the critical task increases the challenges to the subcriticality CSF beyond those irreparably introduced by the plant conditions.

If AFW flow rate is not controlled, the cooldown transient for Reactor vessel inlet temperature will result in an ORANGE path challenge to the integrity CSF, especially after RCPs are tripped. Although the performance standard for this task is tied to the integrity CSF, the challenge to other CSFs is exacerbated by failure to perform the critical task. The other affected CSF would be subcriticality (since the steam break is located outside Containment the Containment CSF would not be challenged).

Failure to control the AFW flow rate, means that the blowdown from all SGs continues at a higher rate than it would if the crew performs the critical task. This continuation of the blowdown at a higher-than-necessary rate significantly worsens the power excursion. It constitutes a challenge to the subcriticality CSF beyond that irreparably introduced by the plant conditions.

2. Manually isolate MSIVs on SG A and C prior to a severe (ORANGE path) challenge develops to the Integrity Critical Safety Function NOTE: This critical task will be accomplished by first attempting to manually actuate a Main Steam Line Actuation when the automatic actuation failed then attempting to manually shut the MSIV via MCB switches. The MSIVs will not shut via the switches.

Directions will be provided to an Auxiliary Operator to isolate the Instrument Air to the MSIVs. IF the Aux Operator is dispatched the A and C MSIV will shut and the task will be successful.

Failure to close the MSIVs causes challenges to CSFs such an omission constitutes a failure by the crew to demonstrate (the ability to) recognize a failure of an automatic actuation of an ESF system or component. They should take one or more actions that would prevent a challenge to plant safety. Uncontrolled depressurization of all SGs causes an excessive rate of RCS cooldown, well beyond the conditions typically analyzed in the FSAR. The excessive cooldown rate creates large thermal stresses in the reactor pressure vessel and causes rapid insertion of a large amount of positive reactivity.

Note: An unanticipated critical task may be created in a scenario should an applicants action or lack of action cause an unexpected RPS or ESFAS actuation. A critical task may be assigned and graded as unsatisfactory even if corrected by another team member prior to the unanticipated RPS/ESFAS actuation. Should the applicant self-correct the action or inaction prior to the unanticipated plant response, a critical task failure should not be assigned to the applicant.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 2 SIMULATOR SETUP For the 2016 NRC Exam Simulator Scenario # 2 Reset to IC-162 password noinstants Go to RUN Silence and Acknowledge annunciators GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner.

Set ERFIS screens to normal at power (The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

SPECIAL INSTRUCTIONS Post conditions for status board from IC-32 and from Scenario 2 Reactor Power 28% steady state Control Bank D at 133 steps RCS boron 1527 ppm Provide a Reactivity Plan to candidates for raising power to >100%

Provide a copy of GP-005, Power Operation (Mode 2 to Mode 1) signed off up to and including step 127 section 5.0.

Update the status board: "B" MDAFW Pump is OOS for motor overhaul Pump has been OOS for 12 total hours and is expected back within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Tech Spec 3.7.1.2, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO or HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, HSD following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Hang restricted access signs on MCR entry swing gates Hang CIT on B MDAFW Pump MCB switch then place protected train placards per OMM-001 Attachment 16 on "A" MDAFW Pump, MS-70 and 72, "B" ESW Pump, "B" RHR Pump and "B" CCW Pump "B" NSW pump Out Of Service for breaker repairs Repairs to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Place CIT on switch for "B" NSW and place protected train placard on "A" NSW pump switch Hang CIT on Boric Acid Transfer Pump B-SB Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 14 of 79 Event

Description:

Raise Power Time Position Applicants Actions or Behavior The crew has been directed to raise power to 100% using GP-005, Power Operation. Step 127 directs a load increase to 90% Reactor power. The DEH load rate is set to 2 DEH Units / Min and demand is set to 960 DEH Units.

When the crew has completed their board walk down and LEAD EVALUATOR:

are ready to take the shift inform the Simulator Operator to place the Simulator in Run. When the Simulator is in run announce:

CREW UPDATE - (SROs Name) Your crew has the shift.

END OF UPDATE Simulator Operator: When directed by the Lead Evaluator go to Run.

SRO Provides direction per GP-005, starting at Step 127 CONTINUE raise Turbine Load at 4 DEH Units / Min (set at 2):

Change Turbine Ramp Rate by depressing the Turbine DEH control panel Load Rate MW/MIN button and observes:

REFERECE displays MW 0002 Depresses 4 - DEMAND displays MW 0004 Depresses ENTER - REFERENCE displays MW 0004 Depresses REF pushbutton and display returns to initial BOP readings of current MW and Demand MW Depresses the GO pushbutton and continues the load increase to 90% Reactor power Monitors turbine and feedwater system response Notifies SRO that the Turbine ramp has started SRO Acknowledges Turbine ramp has begun Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 15 of 79 Event

Description:

Raise Power Time Position Applicants Actions or Behavior The crew may elect to start a dilution before the power Evaluator Note:

change is initiated.

OP-107.01 Section 5.4 RCS Temperature Adjustment (ALT DIL)

FIS-114 may be set for one gallon less than desired. A Procedure Note: pressure transient caused by 1CS-151 shutting results in FIS-114 normally indicating one gallon more than actual flow but two gallons more would be unexpected.

If the translucent covers associated with the Boric Acid and Procedure Caution: Total Makeup Batch counters FIS-113 and FIS-114, located on the MCB, are not closed, the system will not automatically stop at the preset value.

• SETS FIS-114, TOTAL MAKEUP WTR BATCH COUNTER, to obtain the desired quantity.

• VERIFY the RMW CONTROL switch has been placed in the STOP position.

• VERIFY the RMW CONTROL switch green light is lit.

• IF the current potentiometer setpoint of controller 1CS-151, FK-114 RWMU FLOW, needs to be changed to obtain OATC makeup flow, THEN PERFORM the following:

o RECORD the current potentiometer setpoint of controller 1CS-151, FK-114 RWMU FLOW, in Section 5.4.3.

o SET controller 1CS-151, FK-114 RWMU FLOW, for the desired flow rate.

• PLACE the control switch RMW MODE SELECTOR to the ALT DIL position.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 16 of 79 Event

Description:

Raise Power Time Position Applicants Actions or Behavior Alternate Dilution may be manually stopped at any time by Procedure Note:

turning the control switch RMW CONTROL to STOP.

• START the makeup system as follows:

o TURN control switch RMW CONTROL to START momentarily.

o VERIFY the red indicator light is lit.

o IF expected system response is not obtained, THEN TURN control switch RMW CONTROL to STOP.

• VERIFY dilution automatically terminates when the desired quantity has been added.

• IF controller 1CS-151, FK-114 RWMU FLOW, potentiometer was changed in Step 5.4.2.5, THEN PERFORM the following:

o REPOSITION controller FK-114 to the position recorded OATC:

in Section 5.4.3.

o INDEPENDENTLY VERIFY FK-114 potentiometer position of Step 5.4.2.9.a is correct.

• MONITOR Tavg and rod control for proper operation.

• ESTABLISH VCT pressure between 20 - 30 psig.

• TURN control switch RMW MODE SELECTOR to AUTO.

• START the makeup system as follows:

o TURN control switch RMW CONTROL to START momentarily.

o VERIFY the red indicator light is lit.

o IF expected system response is not obtained, THEN TURN control switch RMW CONTROL to STOP.

CREW Continues Load Increase IAW GP-005 Step 128 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 17 of 79 Event

Description:

Raise Power Time Position Applicants Actions or Behavior NOTE: Per EC 74907 and EC 74914, AMSAC is set to arm at 35% (268.5 psig Turbine First Stage Pressure) based upon Siemens prediction of High Pressure Turbine First Stage Pressure versus Load curve. Early monitoring of the AMSAC Armed light will aid in collection of plant data at its Setpoint.

Procedure Note:

NOTE: The respective Main Feed Pump recirculating valve 1FW-8 or 1FW-39 should shut when Main Feed Pump suction flow reaches 4300 KPPH (ERFIS point FCE2210A or FCE2210B).

At 33% Main Turbine load, PERFORM the following:

• At AMSAC Control Panel, DEPRESS the SYSTEM RESET BOP button.

o Dispatches Auxiliary Operator to AMSAC Control Panel to depress the System Reset button.

Simulator Acknowledge directions to go to the AMSAC Control Panel Communicator: and DEPRESS the SYSTEM RESET button.

There isnt anything that needs to be done on the Simulator for the AMSAC system reset.

Simulator Communicator: Wait one minute and report back:

The AMSAC Control Panel SYSTEM RESET button has been depressed.

• CHECK SG LEVEL ATWS PANEL TROUBLE annunciator clear on ALB-17/1-1. (YES)

• PLACE the SG LVL ATWS PANEL BYPASS switch to NORMAL.

BOP o Locates the SG LVL ATWS PANEL BYPASS switch and places switch to NORMAL

• VERIFY SG LEVEL ATWS PANEL BYPASS annunciator clear on ALB-17/2-1 (YES - clears)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 18 of 79 Event

Description:

Raise Power Time Position Applicants Actions or Behavior When power has been increased ~ 5%. With acknowledgement from the other Evaluators, continue with the scenario.

Evaluator Note: Cue Simulator Operator to insert Trigger 2:

Event 2, Power Range NIS Channel 44 failure HIGH (AOP-001, Malfunction of Rod Control and Indication System)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 19 of 79 Event

Description:

Power Range NIS Channel 44 Fails High Time Position Applicants Actions or Behavior On cue from the Lead Evaluator actuate Trigger 2:

Simulator Operator:

Power Range NIS Channel 44 failure HIGH Indications Available

• Uncontrolled rod motion/bistable trips.

OATC RESPONDS to alarms/uncontrolled rod motion.

ENTERS and directs actions of AOP-001, MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM.

SRO Makes PA announcement for AOP entry Holds a crew focus brief AOP-001 Malfunction of Rod Control and Indication System.

OATC PERFORMS AOP-001 Immediate Actions.

Rods cannot be withdrawn until AOP-001 actions have Evaluator Note: been implemented to clear the overpower rod stop.

OWP-RP-26 provides the same actions as AOP-001 to clear the overpower rod stop.

Immediate CHECK that LESS THAN TWO control rods are dropped.

OATC Action (YES)

Immediate OATC POSITION Rod Bank Selector Switch to MAN.

Action Immediate OATC CHECK Control Bank motion STOPPED. (YES)

Action SRO PROCEEDS to Section 3.2.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 20 of 79 Event

Description:

Power Range NIS Channel 44 Fails High Time Position Applicants Actions or Behavior CHECK ALL of the following Rod Control System inputs -

NORMAL:

• RCS Tavg (YES)

OATC

• RCS Tref (YES)

• POWER Range NI channels (NO, NI-44 Failed)

• TURBINE first stage pressure (YES)

RNO Actions:

PERFORM the following:

• IF a power supply is lost, THEN GO TO AOP-024, Loss of Uninterruptible Power Supply. (NO)

SRO

• IF an individual instrument failed, THEN MAINTAIN manual rod control until corrective action is complete.

(YES)

• IF a Power Range NI Channel failed, THEN PLACE the affected NI Rod Stop Bypass switch to BYPASS at the Detector Current Comparator Drawer. (YES)

Proceeds to the Detector Current Comparator Drawer and BOP places NI-44 Rod Stop Bypass switch to BYPASS

• Reports completion of task to the SRO.

Manually OPERATE affected control bank to restore the following:

• Equilibrium power and temperature conditions OATC

• Rods above the insertion limits of Tech Spec 3.1.3.6 and PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report.

• Withdraws Control Bank D to restore T ave with T ref .

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 21 of 79 Event

Description:

Power Range NIS Channel 44 Fails High Time Position Applicants Actions or Behavior VERIFY proper operation of the following: (YES)

• CVCS demineralizers OATC

• BTRS

• Reactor Makeup Control System CHECK that this section was entered due to control banks SRO MOVING OUT. (NO)

GO TO Step 6.

CHECK that NEITHER of the following OCCURRED: (NO)

SRO

• Unexplained RCS Boration

• Unplanned RCS dilution Procedure Note: If control rod motion is not due to instrument malfunction, CVCS or RWMU malfunction, or unexplained boration or dilution, the only remaining explanation not yet explored is a malfunction of the automatic circuitry in the Rod Control System.

CHECK that spurious rod motion is due to malfunction of the SRO Automatic Control Rod System. (NO)

GO TO Step 9.

SRO EXIT this procedure.

• Refer to OWP-RP-26 to remove channel from service.

• Direct operator and I&C to perform OWP-RP-26

• Reviews/prepares OMM-001, Attachment 5 Equipment SRO Problem Checklist for the failure of NI-44

• Contacts WCC for assistance / generation of Work Request Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 22 of 79 Event

Description:

Power Range NIS Channel 44 Fails High Time Position Applicants Actions or Behavior Any Tech Spec evaluation may be completed with a follow-up question after the scenario.

Evaluator Note: NOTE: P-10 functional unit 19.b is an input to 19.c. Actions for functional unit 19.b are address by performing the actions for 19.c.

Enters Instrumentation TS 3.3.1 Functional Unit 2, 3, and 4 ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1. and

c. Either, THERMAL POWER is restricted to less than or SRO equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or,. the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.3.3.1 Reference the below T.S. but it will not apply for this conditions because 3 instruments is the Minimum Number required 3.3.1 Functional Unit 19 b, c, and d.

ACTION 7 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

Acknowledge request and reports from SRO.

Simulator IF asked to report to MCR to perform OWP-RP-26 state that Communicator:

you will report as soon as possible.

It is not required to implement the OWP prior to continuing Simulator Operator:

with the scenario.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 23 of 79 Event

Description:

Power Range NIS Channel 44 Fails High Time Position Applicants Actions or Behavior Note: Any Tech Spec evaluation may be completed with a follow-up question after the scenario.

Note: I&C field activities are not required to be completed before continuing with the next event. The actions for OWP-RP-26 are listed in Attachment 1 in the back of this scenario guide on page 64.

Lead Evaluator:

Note: It is not required for T ave to match T ref or Rod Control to be placed in Automatic before continuing with the next event.

After Control Bank D have been withdrawn to restore T ave with T ref , cue Simulator Operator to insert Trigger 3 Event 3, Containment Fan Cooler Trips (AH-2 A-SA)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 24 of 79 Event

Description:

Containment Fan Cooler Fan (AH-2 A-SA) Trips Time Position Applicants Actions or Behavior On cue from the Lead Evaluator actuate Trigger 3:

Simulator Operator:

Containment Fan Cooler Trips (AH-2 A-SA)

ALB-027-7-2 CONTAIMENT FAN COOLERS AH-2 LOW FLOW-Indications O/L Available:

ALB-001-6-5 ESF SYS TRN A BYPASSED OR INOPERABLE APP The actions from the APP-ALB-027-7-2 are below but crew will ALB-027 also have actions from APP-ALB-001 to address as well

• RESPONDS to alarm on ALB-027-7-2 and ALB-001-6-5 BOP

• Refers to annunciator response CONFIRM alarm using:

• AH-2 fans running indication

• AH-2 fan trouble indication BOP

• Damper position indication VERIFY Automatic Functions:

• Fans trip on overload. (YES)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 25 of 79 Event

Description:

Containment Fan Cooler Fan (AH-2 A-SA) Trips Time Position Applicants Actions or Behavior PERFORM Corrective Actions:

• CHECK the fan status indication.

• IF the running fan has tripped, THEN START standby containment fan per OP-169, Containment Cooling and Ventilation.

BOP

• CHECK proper damper alignment per OP-169, Containment Cooling and Ventilation.

• DISPATCH an operator to check the status of the following breakers:

o 1A22-SA-2A, AH-2 (1A-SA) CNMT Fan Cooler When dispatched as an AO to investigate fan breakers, Simulator approximately 2 to 3 minutes later report there is an Communicator: overcurrent trip condition on breaker 1A22-SA-2A, AH-2 (1A-SA) CNMT Fan Cooler.

• Direct BOP to start the B train of containment fan coolers using OP-169, Containment Cooling and Ventilation

• May direct selecting AH-2 (1B-SA) as the lead fan.

• T/S 3.6.2.3 action (a) is applicable based on the initial trip of SRO AH-2 (1A-SA). (7-day LCO)

• If AH-2 (1B-SA) is selected as the lead fan, then the T/S LCO is no longer in effect.

• If AH-2 (1B-SA) is not selected as the lead fan, then T/S 3.6.2.3 action (a) will remain in effect. (7-day LCO)

Evaluator Note: The OATC may need to borate while holding power.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 26 of 79 Event

Description:

Containment Fan Cooler Fan (AH-2 A-SA) Trips Time Position Applicants Actions or Behavior

• Verifies Initial Conditions met in OP-169 section 5.1.

• Uses CNMT Purge as necessary to control CNMT pressure.

BOP

• Place the control switches for both fans in each Containment Fan Cooler unit AH-2 A-SA (AH-1 B-SB) and AH-3 A-SA (AH-4 B-SB) to LO-SPD.

After any fan cooler is started in low speed, the fan should be allowed to come up to speed for approximately 15 seconds before shifting to fast speed. This reduces the starting current required for high speed operation.

Procedure Note: The following switch sequence must be performed without delay, one fan at a time, to prevent fan coast down before being started in fast speed. This sequence is functionally related (obtain a single result in close sequence or time), allowing signoff to be delayed until running in HI-SPD.

For each of the fans started in Step 5.1.2.2, START the fans in HI-SPD, as follows:

AH-2 A-SA (AH-1 A-SB)

• PLACE AH-2 A-SA (AH-1 A-SB) control switch to STOP.

• PLACE AH-2 A-SA (AH-1 A-SB) control switch to HI-SPD.

AH-2 B-SA (AH-1 B-SB)

BOP

• PLACE AH-2 B-SA (AH-1 B-SB) control switch to STOP.

• PLACE AH-2 B-SA (AH-1 B-SB) control switch to HI-SPD.

AH-3 A-SA (AH-4 A-SB)

• PLACE AH-3 A-SA (AH-4 A-SB) control switch to STOP.

• PLACE AH-3 A-SA (AH-4 A-SB) control switch to HI-SPD.

AH-3 B-SA (AH-4 B-SB)

• PLACE AH-3 B-SA (AH-4 B-SB) control switch to STOP.

• PLACE AH-3 B-SA (AH-4 B-SB) control switch to HI-SPD.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 27 of 79 Event

Description:

Containment Fan Cooler Fan (AH-2 A-SA) Trips Time Position Applicants Actions or Behavior Proceed to OP-169 section 7.1 to secure A train fans.

PLACE the control switch for each of the following fans to be removed BOP from service to STOP:

• AH-2 A-SA (AH-1 A-SB)

• AH-2 B-SA (AH-1 B-SB)

• AH-3 A-SA (AH-4 A-SB)

• AH-3 B-SA (AH-4 B-SB)

Evaluate Tech Spec 3.6.2.3.a

• T/S 3.6.2.3 action (a) is applicable based on the initial trip of AH-2 (1A-SA). (7-day LCO)

• If AH-2 (1B-SA) is not selected as the lead fan, then T/S 3.6.2.3 action (a) will remain in effect. (7-day LCO)

SRO

• Implements OWP-CV-02, Containment Ventilation, for AH-2 Fan. Instructs OATC or BOP to perform actions (energize ESF Bypass Panel A window 4-1 by depressing the associated window)

• Initiates Equipment Problem Checklist and contacts WCC for assistance OACT Implements OWP-CV-02 by depressing window 4-1 on ESF BOP Bypass Panel A After B Train Containment Fan Coolers have been placed in service, cue the Simulator Operator to insert Trigger 4, CSIP Evaluator Note: A Shaft Shear Event 4, CSIP A Shaft Shear Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 28 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior On cue from Lead Evaluator actuate Trigger 4:

Simulator Operator:

CSIP A shaft shear ALB-06-1-1 CHARGING PUMP DISCHARGE HEADER HIGH-Indications LOW FLOW Available:

ALB-08-2-1 RCP SEAL WATER INJECTION LOW FLOW

• RESPONDS to alarms on ALB-06-1-1 and ALB-08-2-1.

• REPORTS CSIP A shaft shear (From MCB indications of OATC no flow and pump still running with abnormal amps)

• Takes MCB switch for A CSIP to STOP and reports to CRS that A CSIP is secured ENTERS and directs actions of AOP-001, MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM.

SRO Makes PA announcement for AOP entry Holds a crew focus brief AOP-018 RCP Abnormal Conditions PERFORMS immediate actions.

• CHECK ANY CSIP RUNNING. (YES but shaft sheared.

NO if CSIP A was preemptively secured when shaft shear was identified)

Immediate OATC

• ISOLATE letdown by verifying the following valves SHUT:

Action o 1CS-7, 45 GPM Letdown Orifice A o 1CS-8, 60 GPM Letdown Orifice B o 1CS-9, 60 GPM Letdown Orifice C Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 29 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior ENTERS AOP-018, RCP Abnormal Conditions.

Makes plant PA announcement for AOP entry SRO No immediate actions (if the OATC did not secure CSIP A)

Conducts a Focus brief REFER to PEP-110, Emergency Classification and Protective SRO Action Recommendations, AND ENTER the EAL Matrix.

The crew should dispatch AOs to investigate.

Simulator IF dispatched, wait 1-2 minutes then report that the shaft is broken on the A CSIP. Report as TB operator (if Communicator:

dispatched) that there are no apparent problems at the breaker for A CSIP.

Minimum allowable flow for a CSIP is 60 gpm which is provided by normal miniflow during normal operation and alternate Procedure Note:

miniflow during safety injection. Maintaining CSIP flow greater than or equal to 60 gpm also satisfies this requirement.

EVALUATE plant conditions AND GO TO the appropriate section:

SRO MALFUNCTION SECTION PAGE Loss of CCW and/or Seal Injection to 3.1 5 RCPs CHECK ALB-5-1-2A, RCP Thermal Bar HDR High Flow, alarm OATC CLEAR. (YES)

CHECK ALL RCPs operating within the limits of Attachment 1.

SRO (YES)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 30 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior

• CHECK ALL RCPs RUNNING. (YES)

• CHECK the following NORMAL for ALL RCPs:

OATC o CCW flow (YES) o Seal Injection flow (NO)

SRO RESTORE using the applicable attachment:

MALFUNCTION ATTACHMENT Loss of Seal Injection flow from Attachment 4 (Page 28)

CSIPs ONLY

• CHECK at least one CSIP RUNNING. (NO)

• DISPATCH an operator to monitor operation of the ASI System.

• ADJUST charging flow as follows:

o PLACE controller FK-122.1, CHARGING FLOW, in MANUAL AND SHUT.

o VERIFY OPEN 1CS-235 SB, CHARGING LINE ISOLATION.

o VERIFY OPEN 1CS-238 SA, CHARGING LINE ISOLATION.

o CHECK RCS pressure GREATER THAN 1400 PSIG (YES) o SET FK-122.1 DEMAND position to 30%.

OATC

• SHUT HC-186.1, RCP Seal WTR INJ Flow.

• VERIFY a suction path for the standby CSIP by performing the following:

o VERIFY CSIP suction flowpath from VCT as follows:

VERIFY greater than 5% level is established in VCT.

(YES)

VERIFY the following valves are OPEN:

• LCV-115C, VCT Outlet (1CS-165) (YES)

• LCV-115E, VCT Outlet (1CS-166) (YES)

GO TO Step 19.

• MAINTAIN CCW HX outlet temperature less than 105°F.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 31 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior The ASI System will add negative reactivity to the RCS. If the Procedure Note: standby CSIP can NOT be started, a plant load reduction must be initiated to accommodate the boration.

• START the standby CSIP. (CSIP B)

• CHECK charging header pressure greater than 2200 psig on PI-121. (YES)

• OPEN HC-186.1, RCP Seal WTR INJ Flow.

OATC

• DIRECT the operator monitoring the ASI System to STOP the ASI Pump by placing CS-210.1, ASI PUMP MOTOR CONTROL SWITCH, in STOP. (At the ASI System Control Panel)

Acknowledge request to place ASI Pump Motor Control Simulator Switch, CS-210.1, in STOP. Wait 1 minute, then perform manipulation, and report that the ASI Pump Motor Control Communicator:

Switch has been placed in STOP and the ASI Pump is Secured.

DO NOT restore Seal Injection to an RCP that has lost all Procedure Caution:

seal cooling for 4 minutes.

• ADJUST HC-186.1, RCP Seal WTR INJ Flow, to establish seal injection flow as necessary to maintain the following:

o LESS than 31 gpm total flow to all RCPs o BETWEEN 8 and 13 gpm to all RCPs OATC

• DIRECT the operator monitoring the ASI System to PLACE CS-210.1, ASI PUMP MOTOR CONTROL SWITCH, in AUTO. (At the ASI System Control Panel)

Simulator Acknowledge request to place ASI Pump Motor Control Communicator: Switch, CS-210.1, in AUTO. Perform manipulation, and report the ASI Pump Motor Control Switch has been placed in AUTO.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 32 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior START CSIP room ventilation per OP-172, Reactor Auxiliary BOP Building HVAC System. (AH-9B)

• Starts AH-9B RESTORE Charging and Letdown flow per OP-107, Chemical OATC and Volume Control System.

The steps for evaluating restoration of letdown begin on Evaluator Note:

page 38.

• MONITOR Tavg to Tref. (ASI injection has added negative reactivity)

• INITIATE action to determine and correct the cause of the loss of the CSIP.

SRO o Completes an Equipment Problem Checklist and contacts WCC for assistance.

o Directs AO to remove control power from A CSIP IF directed - Remove control power from the A CSIP.

Simulator Operator: Use remote function CVC047 to open knife switch for control power to the A CSIP.

Simulator Communicator: Report back after control power has been removed.

CHECK seal injection flow from CSIPs has been established OATC between 8 and 13 gpm to all RCPs.

This step will not be completed before next event is Evaluator Note:

initiated.

WHEN seal injection flow from CSIPs has been established between 8 and 13 gpm, THEN PERFORM OST-1126, Reactor SRO Coolant Pump Seals Controlled Leakage Evaluation Monthly Interval Modes 1-4.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 33 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior ENTERS TS:

• 3.1.2.4, CSIPs, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore at least 2 CSIPs, or be in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SRO

• 3.5.2, ECCS Subsystems, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the inoperable subsystem to operable status, or be in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OP-107 Chemical and Volume Control System OP-107, Section 5.4 - Initiating Normal Letdown Verifies

• Initial Conditions:

o Charging flow established OATC o PRZ Level > 17%

o 1CS-7, 1CS-8, 1CS-9 (Letdown Orifice Isolation valves) SHUT If Charging flow was stopped or greatly reduced prior to letdown being secured, there is a possibility that the Letdown Procedure Caution: line contains voids due to insufficient cooling. This is a precursor to water hammer, and should be evaluated prior to initiating letdown flow.

VERIFY 1CC-337, TK-144 LTDN TEMPERATURE, controller is:

• in AUTO AND OATC

• set for 110 to 120 F (4.0 to 4.7 on potentiometer) normal operation PK-145.1 LTDN PRESSURE, 1CS-38, may have to be Procedure Note:

adjusted to control at lower pressures.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 34 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior

• VERIFY 1CS-38 Controller, PK-145.1 LTDN PRESSURE, in MAN with output set at 50%.

• VERIFY open the following Letdown Isolation Valves:

OATC

• 1CS-2, LETDOWN ISOLATION LCV-459

• 1CS-1, LETDOWN ISOLATION LCV-460

• VERIFY open 1CS-11, LETDOWN ISOLATION.

The following table gives the minimum charging flow required to keep the regenerative heat exchanger temperature below the high temperature alarm when letdown is established:

Letdown Flow Minimum Charging Flow (to be established) necessary when letdown is established Procedure Note: 45 gpm 20 gpm 60 gpm 26 gpm 105 gpm 46 gpm 120 gpm 53 gpm If PRZ level is above the programmed level setpoint, charging flow should be adjusted to a point above the minimum required Procedure Note:

to prevent regen heat exchanger high temperature alarm but low enough to reduce PRZ level.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 35 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior

• ADJUST controller 1CS-231, FK-122.1 CHARGING FLOW, as required to:

o Maintain normal pressurizer level program o Keep regenerative heat exchanger temperature below the high temperature alarm when the desired letdown orifice is placed in service.

• OPEN an Orifice Isolation Valve (1CS-7, 1CS-8, 1CS-9) for the orifice to be placed in service.

OATC

• ADJUST 1CS-38 position by adjusting PK-145.1 output as necessary to control LP LTDN Pressure (PI-145.1) at 340 to 360 psig, to prevent lifting the LP Letdown Relief.

• WHEN Letdown pressure has stabilized at 340 to 360 psig on PI-145.1, LP LTDN PRESS, THEN PERFORM the following:

o ADJUST PK-145.1 LTDN PRESSURE setpoint to 58%

o PLACE the controller in AUTO.

• VERIFY PK-145.1 LTDN PRESSURE Controller maintains Letdown pressure stable at 340 to 360 psig.

• OPEN additional orifice isolation valves (1CS-7, 1CS-8, 1CS-9) as required.

• ADJUST charging flow as necessary to:

OATC o Prevent high temperature alarm (per table above) o Maintain pressurizer programmed level.

• PLACE PRZ level controller, LK-459F, in AUTO, as follows:

o PLACE PRZ level controller, LK-459F, in MAN to cancel any integrated signal.

o RECORD FI-122A.1, CHARGING FLOW. _____GPM Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 36 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior OATC DETERMINE PRZ level controller, LK-459F setpoint by one of the two methods. (N/A Step not performed)

DETERMINE LK-459F based on the table below:

LTDN Flow Charging LK-459F Setpoint Flow (approx.. value) 45 gpm 27 gpm *3%

60 gpm 42 gpm *8%

105 gpm 87 gpm *34%

120 gpm 102 gpm *46%

• Approximate values based on NOT/NOP Calculate PRZ level controller, LK-459F setpoint.

LK-459F setpoint = (Desired Charging Flow / 150 gpm)2 x 100%

OATC

• ADJUST PRZ level controller, LK-459F, to the calculated setpoint.

• PLACE PRZ level controller, LK-459F, in AUTO.

• WHEN the following occurs:

o Program pressurizer level is matching the current pressurizer level AND o Letdown and seal return are balanced with seal injection flow and charging flow.

THEN place controller 1CS-231, FK-122.1 CHARGING FLOW, in AUTO.

• COMPLETE Section 5.4.3.

The SRO may address OWP-CS, CHEMICAL AND VOLUME CONTROL SYSTEM. This OWP verifies status light box Evaluator Note:

verification when CSIP A is tagged out for maintenance and is not needed to be implemented in this scenario.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 37 of 79 Event

Description:

A CSIP Shaft Shear Time Position Applicants Actions or Behavior Contacts WCC for EIR and support. Requests that either SRO control power is removed or breaker is racked out on A CSIP.

IF the crew did not have the control power removed or the breaker racked out on A CSIP, when the SI signal occurs later in the scenario the A CSIP will restart.

Evaluator Note: After Letdown is restored cue Simulator Operator to insert Trigger 5, Total Loss of Cooling Banks on the UAT 1A Transformer Event 5, Total Loss of Cooling Banks on the UAT 1A Transformer Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 38 of 79 Event

Description:

Total Loss of Cooling Banks on the UAT 1A Transformer Time Position Applicants Actions or Behavior On cue from Lead Evaluator insert Trigger 5 Simulator Operator:

Total Loss of Cooling Banks on the UAT 1A Transformer Indications Available ALB-022-3-1, UNIT AUX XFMR-A TROUBLE APP- UNIT AUX XFMR-A TROUBLE ALB-022

• This alarm is common for any local alarm at UAT 1A.

Procedure Note:

• If this annunciator is locked in, consideration should be given for compensatory actions.

Ground fault indication on both a 480V bus and the 6.9KV bus feeding it indicate transformer degradation. This could lead to catastrophic failure. Actions up to and including a reactor trip Procedure Caution: may be required in preparation for loss of bus resulting from transformer de-energization. If the transformer is confirmed to be grounded action should be taken to immediately isolate the grounded transformer.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 39 of 79 Event

Description:

Total Loss of Cooling Banks on the UAT 1A Transformer Time Position Applicants Actions or Behavior CONFIRM alarm using:

(see Alarm / Device / Setpoint in APP ALB-022-3-1)

VERIFY Automatic Functions:

(NO)

• If UAT 1A Lockout Fault Pressure Trip occurs:

o Generator Lockout occurs BOP o Auto transfer to SUT 1A occurs o UAT 1A Cooling Pumps and Fans will stop (To enable automatic control, both 86/G1A and 86/G1B Generator Lockout relays must be reset at the MCR Generator Relay Panels).

PERFORM Corrective Actions:

• IF the loss of UAT 1A results in a loss of (NO)

BOP Emergency Bus 1A-SA, THEN GO TO AOP-025, Loss of One Emergency AC Bus (6.9KV) or Loss of One Emergency DC Bus (125V).

A ground makes the electrical system unreliable; therefore, a Procedure Caution: high priority should be placed on locating and isolating the ground.

• DISPATCH an operator to 286 RAB Swgr Room to check the following relays for grounds:

o Aux Bus 1A-3, UAT 1A to Aux Bus 1A, (NO)

BOP 59/UTAX relay contact status o Aux Bus 1D-1, UAT 1A to Aux Bus 1D, (NO) 59/UTAY relay contact status CREW Dispatches an AO to check the following relays for grounds Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 40 of 79 Event

Description:

Total Loss of Cooling Banks on the UAT 1A Transformer Time Position Applicants Actions or Behavior Acknowledge request and wait 2 minutes then report back Simulator using the telephone NO grounds present on Aux Bus 1A Communicator:

or 1D

• DISPATCH an operator to UAT-1A Local Panel BOP Alarm to check for alarms.

Acknowledge request and wait 3 minutes then report back Simulator using the radio The High Winding Temperature Communicator:

Annunciator is in and No cooling fans are running.

• IF UAT 1A local alarms exist, THEN GO TO (YES)

BOP AOP-039, Startup and Unit Auxiliary Transformer Trouble.

ENTERS and directs actions of AOP-039, STARTUP AND UNIT AUXILIARY TRANSFORMER TROUBLE SRO Makes PA announcement for AOP entry Holds a crew focus brief AOP-039 STARTUP AND UNIT AUXILIARY TRANSFORMER TROUBLE Procedure Note: This procedure contains no immediate actions.

DISPATCH an operator to the alarming transformer with the applicable

Attachment:

BOP

• Attachment 2, Unit Auxiliary Transformer 1A or 1B Trouble Local Actions DISPATCH an operator to perform Attachment 2 for the 1A BOP UAT Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 41 of 79 Event

Description:

Total Loss of Cooling Banks on the UAT 1A Transformer Time Position Applicants Actions or Behavior Acknowledge request and wait 3 minutes then report back Simulator using the radio The Cooling Control toggle switch is in Communicator: MANUAL per AOP-039 Attachment 2 and ALL cooling fans are NOT running.

• Monitor applicable transformer parameters:

o Should monitor listed electrical indicators and temperatures for UAT A.

CREW

• NOTIFY the following personnel of any problems with Startup or Unit Auxiliary Transformers:

• Responsible Engineer

• Load Dispatcher (System Operator)

• Plant/Transmission Activities Coordinator (PTAC)

SRO GO TO the applicable Section:

Section Page 3.2, Unit Auxiliary Transformer Trouble 16 SRO Unit Auxiliary Transformer Trouble, Section 3.2 CHECK alarming UAT supplying associated 6900V BOP (YES)

Aux Buses.

The following actions are taken in response to reports Procedure Note: received from the operator performing Attachment 2, Unit Auxiliary Transformer 1A or 1B Trouble Local Actions.

SRO GO TO the applicable Step:

Section Step Page UAT - Total Loss of Cooling Banks 3 17 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 42 of 79 Event

Description:

Total Loss of Cooling Banks on the UAT 1A Transformer Time Position Applicants Actions or Behavior

• Each UAT has three transformer cooling banks. Each bank consists of one oil pump and three associated cooling fans. A cooling bank is considered to be in service if the Procedure Note: pump and at least one fan are operating.

• This step may be terminated if the transformer has at least one cooling bank restored to service.

UATs are not designed to be self-cooled. If NO transformer cooling banks are operating, the transformer should be removed from service within 30 minutes (1-hour absolute Procedure Caution: maximum) of cooling loss if loaded (6-hours if unloaded) unless cooling is restored. Bubble formation in the oil reduces heat transfer and may result in transformer winding failure.

PERFORM the following for TOTAL loss of transformer cooling SRO banks:

• VERIFY the Cooling Control Switch has been BOP (YES) placed in MANUAL.

• REDUCE UAT load using ONE of the following methods:

o TRANSFER affected buses to the SUT, if available.

(Refer to OP-156.02 as necessary.)

SRO o TRANSFER to equipment with another power supply.

Directs the BOP to transfer Aux Bus 1A and Aux Bus 1D to the SUT per OP-156.02.

The actions for OP-156.02, Section 7.1 are listed in Evaluator Note: Attachment 2 in the back of this scenario guide on page 68.

• CHECK that ANY cooling banks have been SRO (NO) restored.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5 Page 43 of 79 Event

Description:

Total Loss of Cooling Banks on the UAT 1A Transformer Time Position Applicants Actions or Behavior PERFORM the following:

• COMMENCE power reduction using AOP-038, Rapid Downpower.

• GO TO Step 7 to remove transformers from service within the applicable time limits:

o 1-hour from loss of cooling (loaded)

Note: It is not required for the crew to implement AOP-038 before continuing with the next event.

After Aux Bus 1A is transferred to the SUT and the SRO Evaluator Note: communications with the Work Control Center are completed cue Simulator Operator to insert Trigger 6, Main Steam Line Break outside Containment Event 6, Main Steam Line Break outside Containment Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 44 of 79 Event

Description:

Main Steam Line Break Outside Containment Time Position Applicants Actions or Behavior On cue from Lead Evaluator insert Trigger 6 Simulator Operator:

MS Line Break Outside of CNMT

• ALB-09-3-3, PRZ CONT LOW PRESS AND HEATERS ON

• ALB-09-5-1 PRESSURIZER HIGH-LOW PRESS Indications Available

• Rising Reactor power

• RCS pressure lowering

• Charging flow rising

• SG pressures lowering Recommends Reactor Trip, Manually Trips Reactor and OATC recommends Manual MSLI (with no objection from SRO)

EOP-E-0 EOP-E-0, Reactor Trip or Safety Injection Immediate Manually trips the Reactor OATC Action PERFORM immediate actions of EOP-E-0 VERIFY Reactor Trip:

Immediate OATC Action

• Trip breakers RTA and BYA OPEN (YES)

• Trip breakers RTB and BYB OPEN (YES)

• ROD Bottom lights LIT - Zero steps LIT (YES)

• NEUTRON flux dropping (YES)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 45 of 79 Event

Description:

Main Steam Line Break Outside Containment Time Position Applicants Actions or Behavior Check Turbine Trip - ALL THROTTLE VALVES SHUT Immediate BOP Action

• ALL turbine throttle valves - SHUT (YES)

Perform The Following:

a. AC emergency buses - AT LEAST ONE ENERGEIZED Immediate BOP b. AC emergency buses - BOTH ENERGIZED Action (YES) 1A-SA and 1B-SB Buses are energized by off-site power SI - ACTUATED (BOTH TRAINS)

Immediate (YES or NO)

OATC Action SRO may direct initiation of MSLI (following recognition of steam break) while monitoring for SI Actuation Criteria SI - Required (YES)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 46 of 79 Event

Description:

Main Steam Line Break Outside Containment Time Position Applicants Actions or Behavior Perform the Following:

a. Review Foldout page

b. Evaluate EAL Matrix

• Informs Shift Manager to evaluate EAL Matrix Foldout Applies:

SRO Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 47 of 79 Event

Description:

Main Steam Line Break Outside Containment Time Position Applicants Actions or Behavior OATC RCPs should be tripped based on RCP trip criteria being met.

The B RHR pump trips when started. The A RHR pump Evaluator Note: does not AUTO start from the sequencer and must be manually started.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 7 Page 48 of 79 Event

Description:

B RHR Pump Trips and A RHR Pump fails to start Time Position Applicants Actions or Behavior

• VERIFY ALL CSIPs AND RHR pumps - RUNNING. (NO).

Event #7 OATC

• Identifies that there are no RHR pumps running

• Verifies completion of A Sequencer through load block 9

• STARTS A RHR pump Dispatch operator to investigate the cause of the trip of B BOP RHR pump.

If directed to check status of B RHR pump breaker, then Simulator after ~ 2 minutes, report that B RHR pump breaker has an Communicator:

overcurrent trip.

OATC SI flow > 200 gpm: . (YES)

RCS pressure - LESS THAN 230 PSIG. (NO)

OATC GO TO Step 12.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 8 Page 49 of 79 Event

Description:

MSIVs fail to close Time Position Applicants Actions or Behavior Event #8

• MAIN Steam isolation - ACTUATED. (YES)

(MSIVs are still open: RNO)

BOP Perform the following:

• Attempts to manually actuate Main Steam Line Isolation (to close ALL MSIVs but the MSIVs will not shut)

• MSLI may have been actuated pre-emptively after tripping the reactor if crew identified the steam leak.

Critical Critical to attempt actuation, then when failure is identified Task # 1 to contact Aux Operator to isolate Instrument Air and Vent air to attempt to locally shut the MSIVs Any SG pressure - 100 psig lower than pressure in two other BOP SGs (NO, depressurizing at approximately the same rate)

RNO: GO TO Step 16 OATC CHECK CNMT Pressure - HAS REMAINED < 10 PSIG. (YES)

BOP Verify AFW flow - AT LEAST 210 KPPH ESTABLISHED (YES)

Energize AC buses 1A1 AND 1B1 Locates and closes:

BOP

• Emergency Bus A-SA to XFMR A1 Breaker A1 A-SA

• Emergency Bus B-SA to XFMR B1 Breaker B1 A-SA Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 50 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior Directs BOP to: Verify Alignment of Components From SRO Actuation of ESFAS Signals Using Attachment 3, Safeguards Actuation Verification, While Continuing With This Procedure.

BOP Obtains Attachment 3 and performs verifications The actions for EOP-E-0 Attachment 3 are listed in Attachment 3 in the back of this scenario guide on page 71.

The RO will perform all board actions until the BOP completes Attachment 3. The BOP is permitted to properly Evaluator Note:

align plant equipment IAW EOP-E-0 Attachment 3 without SRO approval.

The Scenario Guide still identifies tasks by board position because the time frame for completion of Attachment 3 is not predictable.

First action of Attachment 3:

BOP Directs TB AO -Place Air Compressor 1A and 1B in the local control mode.

Simulator Acknowledge request to place AC 1A and 1B in the local Communicator: control mode:

When contacted to place A/B air compressors in Local Simulator Operator:

Control mode, run CAEP :\air\ACs_to_local.txt.

Simulator When CAEP is complete, report that the air compressors Communicator: are running in local control mode.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 51 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior ASSIGNS OATC to perform the following:

SRO Stabilize AND Maintain Temperature Between 555°F AND 559°F Using Table 1.

OATC Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 52 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior The only action available to control RCS temperature is to limit AFW flow but a flow reduction to < 210 kpph should Evaluator Note:

not be initiated until SG levels have recovered to at least 25% in one SG.

Check PRZ PORVs AND Spray Valves:

• CHECK PRZ PORVs - SHUT (YES)

OATC

• PRZ spray valves - SHUT. (YES)

• CHECK PRZ PORV Block Valves - AT LEAST ONE OPEN. (YES)

• ANY SG pressure - DROPPING IN AN SRO UNCONTROLLED MANNER (YES)

• ANY SG - COMPLETELY DEPRESSURIZED. (NO)

GO TO E-2, FAULTED STEAM GENERATOR ISOLATION SRO Step 1.

EOP-E-2 FAULTED STEAM GENERATOR ISOLATION SRO Enters E-2 At least one SG must be maintained available for RCS cooldown.

Procedure Caution: Any faulted SG OR secondary break should remain isolated during subsequent recovery actions unless needed for RCS cooldown.

SRO Initiate Monitoring of Critical Safety Function Status Trees BOP Verify ALL MSIVs - SHUT (NO)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 53 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior SRO directs the following actions to be taken outside the MCR:

Contacts Aux Operator to:

Perform the following:

SRO

• Locally shut instrument air supply to RAB 261: 1IA-814 (north of AH-19 1A-SA)

• Locally remove cap AND open drain valve: 1IA-1876 (located in corridor outside VCT valve gallery)

Simulator Acknowledge request:

At RAB 261: Shut instrument air supply to 1IA-814 and Communicator:

Remove cap AND open drain valve on 1IA-1876 Simulator Operator: DO NOT PERFORM ANY ACTIONS AT THIS TIME

• Verify all MSIV bypass valves - SHUT (NO)

BOP o Direct Aux Operator to Locally shut or isolate.

• Check Any SG Pressure - STABLE OR RISING (NOT FAULTED) (NO)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 54 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior EOP-Uncontrolled Depressurization Of All Steam Generators ECA-2.1 SRO GO TO ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, Step 1.

SRO Foldout applies.

The ECA-2.1 Foldout criteria that may apply is: Minimum Evaluator Note: Feed Flow: IF level in any SG is less than 25% [40%], THEN maintain a minimum of 12.5 KPPH feed flow to that SG.

CHECK MSIVs AND Bypass Valves:

• VERIFY all MSIVs - SHUT (NO)

• Perform the following: (Previously directed)

BOP

• Locally shut instrument air supply to RAB 261: 1IA-814 (north of AH-19 1A-SA)

• Locally remove cap AND open drain valve: 1IA-1876 (located in corridor outside

• VERIFY all MSIV Bypass Valves - SHUT (YES)

IF the TDAFW pump is the only available source of feed flow, Procedure Caution: THEN maintain steam supply to the TDAFW pump from one SG.

IF local actions are required, attempts to isolate all boundaries Procedure Note: of one SG should be completed prior to starting those for another SG.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 55 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior CHECK Secondary Pressure Boundary for ALL SGs:

• VERIFY SG PORVs - SHUT (YES)

• VERIFY Main FW isolation valves - SHUT (YES)

• SHUT steam supply valves to TDAFW pump:

o SG B: 1MS-70 (SHUT) o SG C: 1MS-72 (SHUT)

BOP

• VERIFY main steam drain isolations before MSIVs -

SHUT: (YES) o SG A: 1MS-231 o SG B: 1MS-266 o SG C: 1MS-301

• VERIFY SG Blowdown isolation valves - SHUT (YES)

• VERIFY MS Analyzer isolation valves - SHUT (YES)

AS SG pressure and steam flow decrease, RCS hot leg temperatures will eventually stabilize and may increase.

Procedure Note:

Adjusting feed flow and steam dump will control RCS hot leg temperatures.

CONTROL RCS Temperature:

OATC

• CHECK RCS cooldown rate - LESS THAN 100°F/HR (NO)

Lower feed flow to 12.5 KPPH to each SG.

Critical BOP Task # 2 Critical to lower AFW flow prior to a severe (ORANGE path) challenge develops to the Integrity Critical Safety Function CREW Identifies RED Path on Heat Sink and transitions to FR-H.1 EOP-Response To Loss Of Secondary Heat Sink FR-H.1 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 56 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior SRO Enters EOP-FR-H.1, Response To Loss Of Secondary Heat Sink This procedure should NOT be performed if total feed flow capability of 210 KPPH is available AND total feed flow has been reduced due to operator action as directed by the EOPs.

The following EOPs direct feed flow to be reduced below 210 KPPH:

ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM Procedure Caution: GENERATORS" FR-S.1, "RESPONSE TO NUCLEAR POWER GENERATION/ATWS" FR-P.1, "RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK" FR-P.2, "RESPONSE TO ANTICIPATED PRESSURIZED THERMAL SHOCK" FR-Z.1, "RESPONSE TO HIGH CONTAINMENT PRESSURE" Feed flow should NOT be established to any faulted SG while a non-faulted SG is available.

Reads Caution prior to step 1 and determines that FR-H.1 should not be performed SRO ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS" EXITS FR-H.1 and returns to EOP-ECA-2.1 Returns to EOP-ECA-2.1 and continues in procedure at step SRO 3.c.

OATC Check RCS hot leg temperatures - STABLE OR DROPPING (YES / NO)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 57 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior OATC Maintain RCP Seal Injection Flow Between 8 GPM And 13 GPM. (Adjusts as needed)

Procedure Note: The RCP Trip Criteria is in effect until SI is terminated.

Check RCP Trip Criteria:

a. Check RCPs - AT LEAST ONE RUNNING (NO - RCPS OATC are OFF)

GO TO Step 6 Check PRZ PORV AND Block Valves:

a. Verify power to PORV block valves - AVAILABLE (YES)

b. PRZ PORVs - SHUT (YES)

OATC c. GO TO Step 6f.

d. Check block valves - AT LEAST ONE OPEN (YES)

e. IF a PRZ PORV opens on high pressure, THEN verify it shuts after pressure decreases to less than opening setpoint.

A SG may be suspected to be ruptured if it fails to dry out following isolation of feed flow. Local checks for radiation can Procedure Note:

be used to confirm primary-to-secondary leakage.

Sampling of the RCS and SGs is directed in Step 29 Check Secondary Radiation:

Check for all of the following:

• Condenser vacuum pump effluent rad - NORMAL (YES)

BOP/SRO

• SG blowdown radiation - NORMAL (YES)

• Main steamline radiation - NORMAL (YES)

• SG activity sample - NORMAL (IF AVAILABLE) (N/A)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 58 of 79 Event

Description:

Main Steam Line Break Outside Containment (Continued)

Time Position Applicants Actions or Behavior Check is RHR Pumps Should Be Stoped OATC

• Check any RHR pumps - RUNNING (B pump - NO, pump tripped on start, A pump - YES)

B SG MSIV will not close in this scenario. After the MSIVs for A and C SGs are closed, pressures will rise and meet EOP-E-2 transition criteria.

LEAD Evaluator Wait UNTIL AFTER SI is RESET then DIRECT the Simulator Note:

Operator run Trigger 8 to close SG A and C MSIVs.

NOTE: C SG MSIV will shut 10 seconds AFTER A MSIV shuts.

NOTE: C SG MSIV will shut 10 seconds AFTER A MSIV shuts.

Simulator Operator:

When directed by the Lead Evaluator run Trigger 8 shut SG A and SG C MSIVs.

Simulator Inform the MCR that Instrument Air has been isolated and Communicator: vented to RAB 261 (E-2 Step 2 RNO is complete)

• RCS Pressure - GREATER THAN 230 PSIG (YES)

• RCS pressure - STABLE OR INCREASING (YES)

• Check RHR pump suction - ALIGNED TO RWST (YES)

OATC

• Reset SI.

• Manually Realign Safeguards Equipment Following A Loss Of Offsite Power. (Refer to E-0, Attachment 6)

• Stop RHR pumps. (YES)

CREW IDENTIFIES A and C MSIVs have shut and SG pressure rising in both SGs Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 6 Page 59 of 79 Event

Description:

Main Steam Line Break (Continued)

Time Position Applicants Actions or Behavior GO TO EOP E-2, Faulted Steam Generator Isolation in SRO accordance with ECA-2.1 FOLDOUT criteria when SG Pressure begins to rise.

At least one SG must be maintained available for RCS cooldown.

Procedure Caution: Any faulted SG OR secondary break should remain isolated during subsequent recovery actions unless needed for RCS cooldown.

Initiate Monitoring Of Critical Safety Function Status Trees.

SRO (IMPLEMENT Function Restoration Procedures as required)

Verify ALL MSIVs - SHUT: (NO - B MSIV is still OPEN)

RNO Actions have already been performed CHECK Any SG Pressure - STABLE OR RISING (NOT FAULTED) (YES)

BOP IDENTIFY Any Faulted SG:

• ANY SG pressure - DECREASING IN AN UNCONTROLLED MANNER. (YES-B)

• ANY SG - COMPLETELY DEPRESSURIZED.

(YES - B)

IF the TDAFW pump is the only available source of feed flow, Procedure Caution: THEN maintain steam supply to the TDAFW pump from one SG.

ISOLATE Faulted SG(s) (Identified In Step 5):

BOP

• VERIFY faulted SG(s) PORV - SHUT. (YES)

• VERIFY Main FW isolation valves - SHUT. (YES)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 9 Page 60 of 79 Event

Description:

SG B AFW Isolation Valve Fails to Close on FW Isolation Signal Time Position Applicants Actions or Behavior

• VERIFY MDAFW AND TDAFW pump isolation valves to faulted SG(s) - SHUT. (NO, unless closed earlier)

Event 9 BOP Closes 1AF-93, MDAFW Pump B and 1AF-143 TDAFW pump isolation valves to SG B to isolate AFW flow.

After ensuring the actions for Event 9 are completed the Lead Evaluator scenario may be terminated at any point since the crew Discretion:

has re-entered EOP-E-2 Shut faulted SG(s) to steam supply valve to TDAFW pump -

SHUT.

• SG B: 1MS-70 (Previously Shut)

• VERIFY MS drain isolation(s) before MSIVs 1MS-266

- SHUT (YES)

BOP

• VERIFY SG BD isolation valves 1BD-30 and 1BD SHUT (YES)

• VERIFY MS analyzer isolation valves - SHUT (YES)

• CHECK CST Level - GREATER THAN 10%. (YES)

A SG may be suspected to be ruptured if it fails to dry out Procedure Note: following isolation of feed flow. Local checks for radiation can be used to confirm primary-to-secondary leakage.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 9 Page 61 of 79 Event

Description:

SG B AFW Isolation Valve Fails to Close on FW Isolation Signal Time Position Applicants Actions or Behavior Any SG - ABNORMAL RADIATION - (NO)

OR UNCONTROLLED LEVEL RISE- (NO)

SRO GO TO Step 10 CHECK if SI has been terminated:

a. Check for all of the following:

• Check BIT outlet valves - SHUT OR ISOLATED (NO)

OATC GO TO Step 13 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 9 Page 62 of 79 Event

Description:

SG B AFW Isolation Valve Fails to Close on FW Isolation Signal Time Position Applicants Actions or Behavior CHECK SI Termination Criteria:

a. RCS Subcooling - GREATER THAN 10°F [40°F] - C (YES) 20°F [50°F] - M

b. LEVEL in at least one intact SG - GREATER THAN 25% [40%]. (YES/NO) OR TOTAL feed flow to SGs - GREATER THAN OATC 210 KPPH. (YES)

c. RCS pressure - STABLE OR INCREASING. (YES)

d. PRZ level - GREATER THAN 10% [30%]. (YES)

RESET SI - takes both SI reset switches to RESET and verifies on the Bypass Permissive status light panel that the SI Actuated light extinguishes and the SI Reset Auto SI-Blocked light comes on MANUALLY realign Safeguards Equipment following a loss of SRO offsite power. (Refer to E-0, "REACTOR TRIP OR SAFETY INJECTION", Attachment 6.)

Reset Phase A and Phase B Isolation Signals.

OATC

• Resets Phase A signal (Phase B was not actuated)

Restore Instrument Air and Nitrogen to Containment:

Open Instrument Air AND Nitrogen Valves to CNMT:

• 1IA-819 OATC

• 1SI-287 STOP all but ONE CSIP.

CHECK RCS pressure - STABLE OR INCREASING. (YES)

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Exam Scenario # 2 Event # 9 Page 63 of 79 Event

Description:

SG B AFW Isolation Valve Fails to Close on FW Isolation Signal Time Position Applicants Actions or Behavior Lead Evaluator: Terminate the scenario.

Inform Simulator Operator to go to FREEZE Inform crew: I have the shift. Please take a seat and dont discuss the scenario. There may be follow up questions that are asked by the examiners after we discuss your responses.

Place the Simulator in FREEZE when directed by the Lead Simulator Operator Examiner.

Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OWP-RP-26 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OWP-RP-26 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OWP-RP-26 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OWP-RP-26 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OP156.02 Section 7.1 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OP156.02 Section 7.1 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 OP156.02 Section 7.1 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # N/A Page 79 of 79 Event

Description:

Scenario Guide Revision Summary Rev. 0 Initial Development Rev. 1 NRC D-1 Outline comments incorporated Rev. 2 Operation validation comments incorporated Rev. 3 NRC 60 day submittal comments incorporated Rev. 4 NRC Prep Week comments incorporated Rev. Final Approved for administration by NRC Region II Harris 2016 NRC Scenario 2 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 Facility: Harris Nuclear Plant Scenario No.: 3 Op Test No.: 05000400/2016301 Examiners: Operators: SRO:

OATC:

BOP:

Initial Conditions: IC-8, MOL, ~4% power

• The unit is on hold until secondary chemistry is within limits with Reactor Power ~ 4%

• GP-005, Power Operation, step 97.b

• The following equipment is under clearance o B NSW pump o 1SI-3, Boron Injection Tank Outlet valve o Boric Acid Transfer Pump B-SB

• Power ascension is on hold for 1SI-3 and B BAT Pump repair completions.

Turnover:

• Start the B Condensate Booster Pump IAW OP-134

• Establish Flow from At Least One High-Head SI Pump Critical Tasks:

• Manually trip all RCPs within 10 minutes of reaching RCP Trip Criteria of

> 200 gpm SI Flow with < 1400 psig RCS Pressure Event Malf. No. Event Type* Event Description No.

1 N/A N - BOP/SRO Start the B Condensate Booster Pump (OP-134) jfb7579 AH-39 Containment Fan Coil Unit fan trip with back up auto start 2 z2715tic C - BOP/SRO failure (C RCP cooling fan) iann xn29e04 tt:144 Letdown Temperature Controller fails LD/Diversion Valve fails to 3 I - RO/SRO jtb143b bypass demineralizers C - RO/SRO Component Cooling Water system leak (AOP-014) with manual 4 ccw08a TS - SRO makeup required to maintain level I - BOP/SRO B SG PORV partially opens with the controller in automatic and 5 pt:308b TS - SRO manual RCP A rising vibration (AOP-018). Vibrations require a manual 6 rcs09a C - RO/SRO Reactor trip (E-0) , then secure A RCP and PRZ spray valve.

7 rcs18a M - ALL SBLOCA inside Containment (E-0 to E-1)

Failure of BIT outlet valve 1SI-4 to open requiring alternate high head sis017 injection flow path use 8 C - RO/SRO sis018 Establish SI flow prior to securing RCPs when EOP-E-0 foldout requires them to be secured zrpk504a Failure of automatic Main Steam Line Isolation to occur when 9 C - BOP/SRO zrpk504b Containment pressure exceeds 3 psig

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 Low power scenario - Turnover to crew is the unit startup on hold. The plant is in Mode 2 with Reactor power less than 5%. Power ascension is on hold until secondary chemistry parameters are within limits and 1SI-3 and the B Boric Acid pump are restored to service. At the conclusion of last shift the B Condensate Booster pump oil system leak repairs and PMT were completed and the clearance was removed. The pump is ready to be returned to service. When the crew takes the shift the expectation is to start the B Condensate Booster pump in accordance with OP-134, Condensate System, Section 5.6. After the pump is running they will hold power until secondary chemistry and the clearances are lifted on 1SI-3 and B BA pump. They should prepare to continue with GP-005, Power Operation, to obtain rated power conditions.

The following equipment is under clearance:

• B Normal Service Water Pump is under clearance for shaft inspection. The pump has been under clearance for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Inspection and return to service is expected to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (No Tech Specs are associated with this component)

• 1SI-3, Boron Injection Tank Outlet valve has been under clearance the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for breaker repairs. The repairs are close to completion and the valve is expected to be returned to service within the next hour. The valve is currently shut with power removed. OWP-SI-01 has been completed. Tech Specs 3.5.2 and 3.6.3 apply.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

• 1SI-3, Boron Injection Tank Outlet valve, Tech Specscontinued

• Boric Acid Transfer Pump B-SB is under clearance for the last 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to breaker blown control power fuses. The problem has been repaired and the clearance will be removed within the next hour. Tech Spec 3.3.3.5.b which is a 7 day LCO and 3.1.2.2 applies (3.1.2.2 is for tracking only). OWP-CS-05 has been completed.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

Upon turnover and assuming the shift the BOP will start the B Condensate Booster pump in accordance with OP-134, Condensate System, Section 5.6 Second Condensate Booster Pump Start up.

Event 1 Start B Condensate Booster Pumps for short term 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reliability run.

Verifiable Action: The BOP will zero out the B Condensate Booster Pump M/A station, open the discharge and recirc valve for the pump then start the B Condensate Booster Pump. After the pump is running the BOP will increase B pump speed and place the controller in Automatic.

After the B Condensate Booster pump is in operation the crew will still be on startup hold but should have evaluated and discussed raising power IAW GP-005 to prepare to place the Main Feedwater Regulating valves in service.

Event 2 Trip of AH-39 Containment Fan Coil Unit fan with back up auto start failure.

Verifiable Action: The failure will cause annunciator ALB-029 4-5 Containment Fan Coolers AH-39 Low Flow-O/L to alarm. The crew should identify that the standby fan did not auto start and start the standby fan. The SRO will complete OMM-001 Attachment 5 and request assistance from the WCC center.

Event 3 Letdown Temperature Controller fails - LD/Diversion Valve fails to bypass demineralizers. This failure will cause temperature controller TK-144 output to decrease to zero. Without cooling to the letdown heat exchanger, temperatures observed on TI-143 will increase. At 135°F annunciator ALB-007-3-2, Demin Flow Diversion High Temp will alarm.

Verifiable Action: The OATC will respond in accordance with the alarm procedure for ALB 007-3-2. The OATC should identify that the divert valve to the VCT has failed to respond and report the failure to the SRO. The OATC should manually bypass the CVCS Demineralizers with 1CS-50 (TCV-143), then take manual control of TK-144 to restore letdown temperature to normal Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

Event 3 (continued)

The SRO should provide a temperature band between 110°F to 120°F to the OATC in accordance with AD-OP-ALL-1000, Conduct of Operations, for operation of components in manual. (Temperature band guidance can be found in OP-107, Chemical Volume And Control). The CVCS Demineralizers should remain bypassed pending an evaluation for continued resin use. The SRO will complete OMM-001 Attachment 5 and request assistance from the WCC center.

Event 4 Component Cooling Water system leak - requiring AOP-014, Loss of Component Cooling Water entry and manual makeup to maintain level. A CCW leak in the running pump suction header will develop. The leak will be within CCW Surge Tank makeup capability.

Verifiable Action: The crew should identify the leak by observation of MCB indications for CCW Surge Tank level or MCB annunciators based on CCW Surge Tank low level.

The OATC will respond to the CCW Surge Tank level change and/or alarm and enter AOP-014, LOSS OF COMPONENT COOLING WATER. The OATC will maintain CCW Surge Tank level in the normal operating range by opening the Demin water make up valve 1DW-15, on the MCB. After dispatching an AO and locally isolating the leak the OATC will then start the standby B CCW pump and secure the running A CCW pump in accordance with OP-145, Component Cooling Water, then isolate the leak.

The SRO will complete OMM-001 Attachment 5 and request assistance from the WCC center.

Tech Spec Evaluation: The SRO should evaluate TS 3.7.3. TS 3.5.2 and TS 3.0.3 (both trains of ECCS INOPERABLE).

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

Event 4 (continued)

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

Event 5 Failure of the B SG PORV fails partially OPEN in AUTO and MANUAL - The B SG PORV pressure instrument will fail high causing the PORV to open. Annunciator ALB-014-8-5 Computer Alarm Steam Generators will alarm and status of the PORV position can be observed on the MCB red/green PORV indication status lights. When the operator attempts to shut the valve the it will remain open requiring local action to isolate the penetration.

Verifiable Action: The BOP should respond to indications and depress the manual pushbutton for PK-308B1 and lower the output to zero.

Tech Spec Evaluation: The SRO should evaluate TS 3.6.3, Containment Isolation Valves and PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report. IF the Tech Specs are not referred to during the scenario, then (if required) ask a follow up question at the end of the scenario dealing with the LCO.

TS 3.6.3 - Action c. Isolate the affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The redundant manual isolation valve per PLP-106, Attachment 5 is 1MS-61.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

Event 6 RCP A high vibration. During this event the A RCP vibrations will begin to increase and over 3 minutes peak at 28 mils shaft. Note: the shaft vibration instrumentation reads up to 30 mils. The crew will respond to the A RCP malfunction by either identifying rising vibrations or when ALB-010-1-5, RCP-A Trouble alarms. The crew should see the A RCP vibration probe readings are increasing. The crew should enter AOP-018, Reactor Coolant Pump Abnormal Conditions and perform the immediate actions of checking any CSIP running (YES). Vibrations will continue to increase and exceed AOP-018 Attachment 1 RCP trip criteria of 20 mils shaft.

Verifiable Action: The OATC will perform a manual Reactor trip and secure the A RCP and associated PRZ spray valve after E-0 immediate actions are completed.

The crew will then transition from EOP E-0 to ES-0.1, Reactor Trip Response. The Lead Examiner can allow the crew to stabilize the plant then have the Simulator Operator insert the next event (Small Break LOCA).

Event 7 - Major - Small Break LOCA SBLOCA inside Containment (Loop A)

The crew should recognize a rapid decrease in Pressurizer level and RCS pressure.

If the crew responds quickly to the event they may manually actuate a Safety Injection based on ES-0.1 foldout criteria of not being able to maintain Pressurizer level > 5% or RCS subcooling < 10°F. If they do not respond quickly an Automatic Safety Injection will occur. The crew will then transition from ES-0.1 back to E-0, Reactor Trip or Safety Injection. They will again carry out immediate actions of E-0.

Event 8 Failure of BIT outlet valve 1SI-4 to open requiring alternate high head injection flow path use. 1SI-4 will fail to automatically open with the Safety Injection signal and cannot be manually opened from the MCB switch. Additionally, 1SI-3 was under clearance and cannot be opened from the MCB due to control power being removed from the breaker.

In order to obtain Safety Injection flow the crew will have to use the alternate high head injection flow path as directed by E-0 RNO actions.

Verifiable Action: The OATC will OPEN alternate high head Safety Injection to cold legs valve 1SI-52 SA and then identify Safety Injection flow exceeding 200 gpm.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SCENARIO

SUMMARY

2016 NRC EXAM SCENARIO 3 (continued)

Event 8 (continued)

Establish SI flow prior to securing RCPs when EOP-E-0 foldout requires them to be secured. Shortly after entering E-0, the crew should recognize that the RCS pressure is low enough to meet Foldout Criteria for securing all RCPs but there is no flow indicated on FI-943 (normal SI flow indication). The crew will have to establish SI flow by opening the alternate high head Safety Injection to cold legs valve 1SI-52 SA.

Verifiable Action: After the OATC opens 1SI-52 SA adequate flow (> 200 gpm) will be indicated on FI-940 (alternate SI flow indication). Once flow is verified the low press and SI flow criteria will be met and the OATC will then STOP the B and C RCPs.

Event 9 Failure of automatic Main Steam Line Isolation to occur when Containment pressure exceeds 3 psig. As the Small Break LOCA continues to flow RCS to the Containment the pressure in the Containment will continue to rise. An automatic Main Steam Isolation signal is generated when Containment pressure is > 3.0 psig. The crew will have shut the MSIVs due to the cooldown encountered from securing the A RCP but the MSIV before seat drain valves (1MS-231, 1MS-266, 1MS-301) will remain OPEN. The MCB switch for manual actuation of MSLI will NOT function.

Verifiable Action: The BOP will have to manually shut from the individual MCB switches for each MSIV before seat drain valve. These valves can be verified that they are not in the correct position by review the Safeguards panel on ERFIS (after a MSLI actuation).

Event 7 (continued)

After making transitions from E-0 to E-1 to ES-1.2 the crew will determine that the RCS cool down rate exceeded 100°F/HR and will have to wait prior to reducing RCS temperature further. The scenario ends when the crew has determined that the 100°F/HR cool down rate has been exceed.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 CRITICAL TASK JUSTIFICATION:

1. Establish Flow from At Least One High-Head SI Pump. Prior to manually tripping ALL RCPs.

Failure of the crew to manually align Safety Injection flow through the alternate high head injection flow path results in a degradation of the capacity of the ECCS systems. The only available makeup water source during this event is the high pressure safety injection from the CSIPs. Until the alternate high head safety injection flow is aligned the safety margin of the plant is significantly reduced and may result in irreparable damage to the reactor core.

The acceptable results obtained in the FSAR analysis of a small-break LOCA are predicated on the assumption of minimum ECCS pumps injection. The analysis assumes that a minimum pumped ECCS flow rate, which varies with RCS pressure, is injected into the core. The flow rate values assumed for minimum pumped injection are based on operation of one each of the following ECCS pumps: Charging/SI pump, high-head SI pump and low-head SI pump.

Operation of this minimum required complement of ECCS injection pumps is consistent with the FSAR assumption that only minimum safeguards is actuated.

2. Manually trip all RCPs within 10 minutes of reaching RCP Trip Criteria of > 200 gpm SI Flow with < 1400 psig RCS Pressure Securing RCPs during a SB LOCA event will prevent depleting the RCS to a critical inventory by pumping more mass through the break than would occur if the RCP operation were ceased. (Critical inventory is defined as the amount of inventory remaining in the RCS when the break completely uncovers and the break flow changes from a mixture of liquid and steam to all steam.) The LOCA event in this scenario is a SB LOCA that requires the RCPs to be secured when E-0 foldout conditions are met. IF the crew continues to allow the RCPs to operate due to lack of establishment of SI flow of > 200 gpm then RCS inventory will continue to deplete. Manually tripping the RCPs before depletion below the critical inventory conservatively ensures that Peak Clad Temperature remains below 2200°F. This action should be accomplished within 10 minutes of RCP Trip Criteria of > 200 gpm SI Flow with < 1400 psig RCS Pressure and prior to transitioning out of EOP-E-0.

Note: An unanticipated critical task may be created in a scenario should an applicants action or lack of action cause an unexpected RPS or ESFAS actuation. A critical task may be assigned and graded as unsatisfactory even if corrected by another team member prior to the unanticipated RPS/ESFAS actuation. Should the applicant self-correct the action or inaction prior to the unanticipated plant response, a critical task failure should not be assigned to the applicant.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Scenario Outline Form ES-D-1 HARRIS 2016 NRC EXAM SCENARIO 3 SIMULATOR SETUP For the 2016 NRC Exam Simulator Scenario # 3 Reset to IC-163 password noinstants Go to RUN Silence and Acknowledge annunciators GO TO FREEZE and inform the lead examiner the Simulator is ready. DO NOT GO TO RUN until directed by the lead examiner.

(The examiner has provided to the candidate with initial conditions and the initiating cues prior to placing the simulator in RUN.)

SPECIAL INSTRUCTIONS Scenario will have ONLY the A Condensate Booster Pump running. Turnover information for starting the B Condensate Booster pump: will be that the Condensate Booster Pump B Lock-Out Relays are reset. The crew will need to start the B pump then secure the A Condensate Booster Pump. OP-134 Section 5.6.1 Initial condition

1. 3 requires Rx Power to be >5% and should be N/Ad for this start since it will be for swapping pumps.

Post conditions for status board from IC-8 with Reactivity Data for RCS boron of 1567 Mode 2 <5% Reactor power Startup on HOLD due to 1SI-3 and B BA Pump.

Note: Unit cannot enter Mode 1 until 1SI-3 and BA pump are restored.

Provide a marked up copy of GP-005 Rev 93 through Step 97 the step for >5% power is not initialed Control Bank D at 109 steps RCS boron 1567 ppm RCS press 2220 - 2250 psig all PZR heaters ON SG level maintained with "A" MFW pump and FW Reg Bypass Vlvs in Auto RCS temp 558°F, stable on Steam Dumps RCS temp band from step 51 is 555°F - 561°F Main Turbine at 1800 rpm Hang CIT on 1SI-3, Boron Injection Tank Outlet valve Place completed copy of OWP-SI-01 in OWP book Hang CIT on Boric Acid Transfer Pump B-SB Hang CIT on B NSW Pump Post Reactivity Signs (GP-005 Prerequisite #32)

Hang STAR placard on Rod Control In/Out Switch Hang STAR placard on Steam Dump controller M/A station Set CRT screen 3 to "QP POAH" Update the status board:

1SI-3 Tech Spec 3.4.2 - 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO, OOS for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> BA Transfer pump B Tech Spec 3.3.3.5.b - 7 day LCO, OOS for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 12 of 71 Event

Description:

Start B Condensate Booster Pump Time Position Applicants Actions or Behavior When the crew has completed their board walk down and are ready to take the shift inform the Simulator Operator to place the Simulator in Run. When the Simulator is in run Lead Evaluator:

announce:

CREW UPDATE - (SROs Name) Your crew has the shift.

END OF UPDATE Simulator When directed by the Lead Evaluator, ensure that the Communicator: annunciator horns are on and place the Simulator in RUN.

After the crew has taken the shift the BOP will place B Lead Evaluator:

Condensate Booster Pump in service.

Before inserting the first failure wait for the B Evaluator Note: Condensate Booster Pump start to be completed AND the BOP to return to the at the controls area.

The actions for OP-134, Section 5.6 are listed in Evaluator Note: Attachment 1 in the back of this scenario guide on page 60.

Directs BOP to start the B Condensate Booster pump in SRO accordance with OP-134 section 5.6 Performs OP-134 Reviews Sections 5.6, Starting Second Condensate /

BOP Condensate Booster Pump

• Contacts Turbine Building AO to observe swap Simulator I printed out a copy of OP-134 Sections 5.6 and have the Communicator: procedure sections in hand.

Informs AO that they are about to start B Condensate Booster BOP Pump and

• Makes PA announcement prior to starting pump Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 13 of 71 Event

Description:

Start B Condensate Booster Pump Time Position Applicants Actions or Behavior Condensate System, Section 5.6, Second Condensate OP-134 Booster Pump Start Up Step 1: PERFORM prestart checks on Condensate Booster Pump B per Attachment 6.

BOP Contacts Aux Operator to perform prestart checks per Attachment 6 I walked down the B Condensate Booster Pump and Simulator completed the prestart checks. The pump is ready to be Communicator:

started.

Step 2: VERIFY CONDENSATE BOOSTER PUMP B BOP RECIRC, 1CE-261 in MODU and shut.

• Checks 1CE-261 in MODU and shut. (YES)

There are no Condensate Booster Pump trips to protect the Procedure Caution:

pump from running without seal water.

Step 3: PLACE PK-2308 CNDST BSTR PUMP B SPEED BOP CONTROLLER to MAN and zero the demand signal.

• Checks PK-2308 in MAN with zero demand signal Step 4: VERIFY OPEN 1CE-268, CONDENSATE BOOSTER BOP PUMP B DISCHARGE.

• OPENS 1CE-268

• Computer points listed in Section 6.0 of this procedure may be monitored for information.

• When the Condensate Booster Pump control switch is Procedure Note: placed to the START position, the Aux Lube Oil Pump will start and supply the VSF Coupling with oil until oil pressure is greater than or equal to 10 psig as indicated on PI-01LO-2304B, at which time the Condensate Booster Pump starts.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 14 of 71 Event

Description:

Start B Condensate Booster Pump Time Position Applicants Actions or Behavior The amount of time the associated recirc valve, 1CE-261 is Procedure Caution: open, should be minimized due to lack of lubrication without Condensate Booster Pump running.

PLACE the control switch CONDENSATE BOOSTER PUMP B RECIRC, 1CE-261 in the OPEN position immediately prior to BOP starting Condensate Booster Pump B.

• Places control switch for B Condensate Booster Pump recirc valve 1CE-261 to OPEN Starting the second Condensate Booster Pump may cause the previously running pump controller to reject to Manual. This is due to the speed sensor on the pump being started initially providing a speed input signal that is based on electrical noise.

Procedure Note:

If the running CBP controller rejects to manual, it is permissible to return the controller to Auto once the CBP being started reaches the no-load speed. If the controller again rejects to manual, then further investigation would be required.

Step 6: START B Condensate Booster Pump.

• Places B Condensate Booster Pump start switch to BOP START

• Verifies indications that the pump has started and running as expected Simulator Report that the B Condensate Booster pump has a good Communicator: start Step 7: Locally VERIFY Condensate Booster Pump B Aux Lube Oil Pump has stopped.

• Contacts Aux Operator to verify Aux Lube Oil Pump has stopped Simulator B Condensate Booster Pump Aux Lube Oil Pump has Communicator: STOPPED Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 15 of 71 Event

Description:

Start B Condensate Booster Pump Time Position Applicants Actions or Behavior Step 8: CHECK differential pressure across the Pall Replaceable Duplex Filter, as indicated between PI-01LO-2304B1 and PI-01LO-2304B2 is less than 15 PSI (less than 9 BOP PSI when oil temperature has warmed up to normal).

• Contacts Aux Operator to check differential pressure across the filter The differential pressure across the Duplex Filter, as Simulator indicated between PI-01LO-2304B1 and PI-01LO-2304B2 is Communicator:

11 PSI. I will continue to monitor for normal response.

Step 9: N/A Step 10: SLOWLY INCREASE the demand signal on PK-2308 CNDST BSTR PUMP B SPEED CONTROLLER to match the demand signal on the previously running Condensate Booster BOP Pump Speed Controller.

• Slowly increases demand signal on PK-2308 and matches the demand signal on the A Condensate Booster Pump Speed Controller.

Step 11: WHEN the demand signals are matched, THEN PLACE PK-2308 CNDST BSTR PUMP B SPEED BOP CONTROLLER to AUTO.

• Verifies demand signals are matched and places PK-2308 in AUTO Step 12: PLACE the control switch for CONDENSATE BOOSTER PUMP B RECIRC, 1CE-261 in the MODU position.

BOP

• Places control switch for B Condesate Booster Pump recirc valve 1CE-261 to MODU position.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 16 of 71 Event

Description:

Start B Condensate Booster Pump Time Position Applicants Actions or Behavior Step 13: After 5 to 10 minutes of running, VERIFY the VSF coupling oil level is in the normal operating range.

BOP

• Contacts Aux Operator to verify the VSF coupling oil level is in the normal operating range after 5 to 10 minutes from when the pump was started.

Acknowledge request to verify the VSF coupling oil level is Simulator normal in 5 to 10 more minutes.

Communicator:

I will call you back if there is something abnormal.

When the BOP has completed start of the B Condensate Booster pump, and the CRS has been informed the B Condensate Booster Pump is running, continue with the Evaluator Cue: scenario.

Cue Simulator Operator to insert Trigger 2:

Event 2 - AH-39 fan trip with backup fan auto start failure Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 17 of 71 Event

Description:

AH-39 Containment Fan Coil Unit Fan trip Time Position Applicants Actions or Behavior On cue from the Lead Evaluator actuate Trigger 2:

AH-39 Containment Fan Coil Unit Fan trip Simulator Operator:

(Note: there is approximately 20 second delay from the initiation of the trigger to actuation of fan trip and alarm)

• ALB-029-4-5 CONTAINMENT FAN COOLERS AH-39 LOW Indications Available FLOW - O/L

• Increasing C RCP stator winding temperatures BOP RESPONDS to alarms and ENTERS APP-ALB-029-4-5

• CONFIRM alarm using:

o AH-39 fans running indication (NO) o Damper position indication (YES)

• VERIFY Automatic Functions:

o Running fan trips (YES) o Backup fan starts (NO) (BOP starts the standby fan when directed by SRO, may utilize OP-169 section 5.2 or the BOP APP for guidance)

• PERFORM Corrective Actions:

o CHECK standby fan STARTS AND lead fan STOPS.

o DISPATCH an operator to check status of the following breakers:

1D1-1A, AH-39 (1A-NNS) CNMT Fan Cooler 1E1-7C, AH-39 (1B-NNS) CNMT Fan Cooler Directs BOP to start standby Air Handler (this may take place prior to getting the report of the breaker condition)

SRO NOTE: The BOP may start the standby fan and then state to the SRO that he/she is starting the fan. In this case the SRO will just concur with this action.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 18 of 71 Event

Description:

AH-39 Containment Fan Coil Unit Fan trip Time Position Applicants Actions or Behavior After approximately 1 minute from being dispatched to check the breaker for 1D1-1A, AH-39 (1A-NNS) CNMT Fan Simulator cooler breaker, report that:

Communicator: The indications on the Static Trip Unit show that an Overload Condition occurred for AH-39 A fan. There are no abnormalities on the AH-39B breaker.

o IF any breaker has tripped on OVERLOAD or SHORT CIRCUIT as indicated on the Static Trip Unit, THEN PERFORM the following: (Directs AO to perform based on report from communicator)

BOP DEPRESS the breaker Alarm Reset.

RACK OUT the breaker using OP-156.02, AC Electrical Distribution.

VERIFY cause of the over current trip is determined prior to returning the breaker to service.

Simulator Acknowledge request to perform directed actions Communicator: at 1D1-1A Rack out breaker 1D1-1A for AH-39 and clear alarm

• Activate Trg 15 Trigger 15 will clear the alarm then 30 seconds later it will Simulator Operator: override the switch to STOP and turn off the RED and GREEN MCB switch lights.

Have communicator report back 30 seconds after running the trigger.

Reviews/prepares OMM-001, Attachment 5 Equipment SRO Problem Checklist for the failure of AH-39.

Contacts WCC and EMs for assistance with repairs.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 19 of 71 Event

Description:

AH-39 Containment Fan Coil Unit Fan trip Time Position Applicants Actions or Behavior When the BOP has completed start of the AH-39 (1B-NNS)

CNMT Fan cooler, and the CRS has been informed the AH-39 (1B-NNS) CNMT Fan cooler is running, continue Evaluator Cue: with the scenario.

Cue Simulator Operator to insert Trigger 3:

Event 3 - Letdown Temperature Controller fails to bypass demineralizers Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 3 Page 20 of 71 Event

Description:

LD Temp Controller fails with failure to bypass Demineralizers Time Position Applicants Actions or Behavior Event 3 - Letdown Temperature Controller fails to bypass demineralizers Evaluator Cue:

When breaker racking and assistance communications are completed, cue Simulator Operator to insert Trigger 3.

On cue from the Lead Evaluator actuate Trigger 3:

Simulator Operator: Letdown Temperature Controller fails LD/Diversion Valve fails to bypass demineralizers

• ALB-007-3-2, DEMIN FLOW DIVERSION HIGH TEMP

• ALB-007-5-5, COMPUTER ALARM CHEM & VOL SYSTEMS Indications Available:

• TK-144 output - decreases to 0

• TI-144.1 HX Out Temp - decreases to 0

• TI-143 temperature increasing If the crew catches this failure early and temperature does not increase above 135°F then they may NOT identify that 1CS-50 is failed since there will be no reason for the valve to change position.

Evaluator Note: Changes in Letdown temperature can have an effect on the demineralizers resins. During high input temperature a boron release can occur (effects similar to a boration) and during low input temperatures a boron absorption can occur (effects similar to a dilution).

RO Responds to alarm and enters APP-ALB-007-3-2.

• CONFIRM alarm using:

o TI-143, LP Letdown Temperature.

o Reports TI-143 reading or trending high.

RO

• VERIFY Automatic Functions:

o Manually positions 1CS-50, Letdown to VCT/Demin, to divert flow to the VCT.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 3 Page 21 of 71 Event

Description:

LD Temp Controller fails with failure to bypass Demineralizers Time Position Applicants Actions or Behavior Evaluator Note: CRS may direct manual control before APP is implemented.

• PERFORM Corrective actions:

o VERIFY that 1CS-50 diverts flow to the VCT, bypassing the BTRS and Purification Demineralizers.

o PERFORM the following as needed to lower letdown temperature:

RO VERIFY proper charging flow is established. (YES)

LOWER letdown flow. (N/A - CCW Problem) o IF CCW flow to the Letdown Heat Exchanger appears low, THEN:

TAKE manual control of TK-144.

OPEN 1CC-337, to raise CCW flow.

• Provide a temperature band IAW OMM-001 for operation of components in manual. OP-107 page 31 with TK-144 controller in auto directions is to maintain temperature from 110 - 120°F. (NOTE this is not the only procedure that provides temperature guidance)

SRO

• Notify Health Physics that the demineralizers will remain bypassed

• Reviews/prepares OMM-001, Attachment 5 Equipment Problem Checklist

• Contacts Work Control and/or System Engineer for assistance.

If contacted as WCC, System Engineer, Health Physics or Simulator Chemistry: Maintain flow bypassing the demineralizers Communicator:

until a resin damage assessment is completed.

After crew has restored CCW flow to the Letdown Heat Exchanger, cue Simulator Operator to insert Trigger 4.

Evaluator Cue: NOTE: there is a 2 minute delay prior to the CCW alarm actuating.

Event 4 - Component Cooling Water (CCW) system leak Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 22 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior On cue from the Lead Evaluator actuate Trigger 4:

Simulator Component Cooling Water system leak Communicator: (Note: After the trigger is inserted it will take ~ 2 minutes for the CCW alarm to occurs)

• ALB-005-8-5, COMPUTER ALARM CCW SYSTEM Indications Available

• ALB-005-6-1, CCW SURGE TANK HIGH-LOW LEVEL

• ALB-026-2-1, GROSS FAILED FUEL DET TROUBLE The crew may enter AOP-014, LOSS OF COMPONENT Evaluator Note:

COOLING WATER, when the first alarm is confirmed.

RESPONDS to alarm ALB-005-8-5, COMPUTER ALARM CCW RO SYSTEM, and ALB-005-6-1, CCW SURGE TANK HIGH-LOW LEVEL.

BOP REPORTS CCW Surge Tank level alarm on alarm screen.

Actions from the APP are below but crew will most likely APP perform a direct entry into AOP-014.

ALB-005 GO to page 23 if AOP-014 is entered The CCW Surge Tank baffle plate separates Side A and Side B Procedure Note:

up to the 38% level.

CONFIRM alarm using:

RO

• LI-670A.1, CCW Surge Tank Level (Side A)

• LI-676A.1, CCW Surge Tank Level (Side B)

MAINTAIN CCW Surge Tank level per SRO direction using RO 1DW-15, CCW Make Up.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 23 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior VERIFY Automatic Functions:

• CCW flow to the GFFD and Primary Sample Panel will isolate on a low CCW Surge Tank level (40%).

RO (Level should remain > 40%)

• CCW Holdup Tank Transfer Pump and the CCW Drain Tank Transfer Pump will trip on a high CCW Surge Tank level (75%). (N/A)

PERFORM Corrective Actions:

• IF surge tank level is high AND rising. (N/A)

• IF radiation activity level is increasing, THEN GO TO AOP-016, Excessive Primary Plant Leakage. (NO)

RO

• IF the alarm is due to plant heatup, THEN DRAIN the surge tank to normal level. (NO)

• IF surge tank level is low, THEN GO TO AOP-014, Loss of Component Cooling Water. (YES)

ENTERS and directs actions of AOP-014, LOSS OF COMPONENT COOLING WATER.

SRO Makes PA announcement for AOP entry Holds a crew focus brief AOP-014 Loss Of Component Cooling Water Procedure Note:

• This procedure contains no immediate actions.

• Loss of CCW may require implementation of the SHNPP Emergency Plan.

REFER TO PEP-110, Emergency Classification And Protective SRO Action Recommendations, AND ENTER the EAL Matrix .

EVALUATE plant conditions AND GO TO the appropriate SRO section:

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 24 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior

• Determines section 3.2 is appropriate CONTACTS AO to check RAB for CCW leaks.

BOP/RO (This action is not procedurally directed but should happen during the course of implementing this AOP.)

Acknowledge request.

Simulator Communicator: Wait 1 minute then report a leak in the suction header strainer flange between 1CC-27 and CCW Pump A.

The crew should begin to trace out where the leakage is and what to do to isolate the leakage using the MCR Simplified Flow Diagrams (SFDs). They should identify Evaluator Note: that the leak is isolated by shutting 1CC-27. They will see that isolating the leak will require them to secure the A CCW pump and start the standby pump. The crew may also shut 1CC-36 to isolate the discharge of the pump.

Identifies leak location on SFDs and determines method to isolate the leakage Crew

• Secure A CCW pump and start B CCW pump

• Shut 1CC-27

• Shut 1CC-36 Simulator Acknowledge crew directions to shut 1CC-27 and/or 1CC-Communicator: 36.

Close 1CC-27 and/or 1CC-36 and delete MF CCW08A Simulator Communicator: Then have Communicator report the valves closed approximately 1 minute after completion.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 25 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior Simulator If a leak report is requested from MCR to RW Operator:

Communicator: Rad Waste reports a rise in RAB floor drain in-leakage.

After leak is isolated by shutting 1CC-27 the crew may also shut 1CC-36, IF the control power to the A CCW pump is Evaluators Note: not removed from the pump any auto start of the A CCW pump will cause the pump to run without a suction or discharge path The path through the procedure may be different for each crew since it depends on when the leak location is known Evaluators Note: and how certain questions are answered. However, each crew should initiate makeup, swap running pumps, isolate the leak, and address the Tech Spec.

The GFFD and RCS sample panel will isolate on low CCW Procedure Note:

Surge Tank level of less than or equal to 40%.

MAINTAIN CCW Surge Tank level between 45% and 75%

RO using 1DW-15, CCW Make Up.

An affected CCW Pump is one to which any of the following apply:

• Less than 4% level indicated on the CCW Surge Tank Procedure Note:

• Exhibits abnormal flow

• IF non-essential header isolation valves are open, less than 4% level indicated on either CCW Surge Tank affects both CCW Pumps.

CHECK BOTH of the following conditions exist:

• ALL CCW Surge Tank level indicators are greater than 4%

RO (YES)

• CCW Pump flow indication is NORMAL (YES)

SRO CHECK EITHER RHR Train in Shutdown Cooling Mode. (NO)

RO/SRO CHECK RCS temperature greater than 200°F. (YES)

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 26 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior RO CHECK CCW Surge tank level is > 40% (YES)

CHECK that CCW loads from the Non-Essential header require RO/SRO isolation by ANY of the following: (NO)

RO/SRO CHECK CCW lost to ANY operating RHR Train: (NO)

The steps highlighted below may not be performed IF the Evaluator Note: crew starts the 2nd CCW pump and has isolated the leak before reaching these steps.

Operation of RCPs for greater than 10 minutes without CCW Procedure Caution: cooling to the motor oil coolers may result in RCP bearing damage.

CHECK CCW expected to be lost for greater than 10 minutes.

SRO (NO)

Reactor Makeup Water Tank contains potentially tritiated water.

Making up to the CCW System from the Reactor Makeup Procedure Caution: Water Tank could result in CCW System contamination.

Operation of the system while it is contaminated requires an evaluation per 10CFR50.59.

RO CHECK CCW Surge Tank level STABLE OR RISING. (YES)

• If the leak location is known, non-applicable steps (Steps 15 through 21) are not required to be performed.

• If the leak location is not known, the CRS may direct performance of Steps 15 through 21 in any order. Steps 22 Procedure Note: and 23 directing diagnostic and walkdown may be useful in determining leak location and may be performed prior to or in parallel with Steps 15 through 21.

• Elevated leakage may be indicated by higher indicated levels, higher level controller setpoints, annunciators, Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 27 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior evolutions in progress, notification by personnel, Chemistry sample results or other means.

• RCDT in-leakage is indicated by elevated level controller output.

Simulator If a leak report is requested from MCR to RW Operator:

Communicator: Rad Waste reports a rise in RAB floor drain in-leakage.

From the note above since the leak location is known from the Aux Operator report steps 15 - 21 are NOT required to be SRO performed. Attachment 3 may be referenced but is not required to be implemented.

If walkdown has NOT commenced acknowledge request.

Simulator Communicator: Wait 1 minute then report a leak in the suction header strainer flange between 1CC-27 and CCW Pump A.

PERFORM a walkdown of CCW piping looking for leaks.

SRO

• Walkdown was performed and leak location identified and isolated Leakage in excess of 15.8 gph per train (unanticipated makeup Procedure Note: greater than twice per shift) could exceed surge tank makeup capacity under design basis conditions.

WHEN the leak is LOCATED, THEN PERFORM the following:

• CHECK that CCW System leakage can be isolated.

(YES)

• INITIATE corrective actions to restore system to service.

• DIRECT Chemistry to sample CCW for proper corrosion inhibitor concentration. - Contacts Chemistry Simulator Acknowledge request for CCW sample Communicator:

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 28 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior RESTORE CCW flow to the following, as needed:

• Sample Heat Exchangers SRO

• GFFD

• Excess Letdown

• RCDT Heat Exchangers The SRO will complete OMM-001 Attachment 5 and request assistance from the WCC center to repair system leakage.

The SRO should evaluate TS 3.7.3, 3.5.2 and 3.0.3.

TS 3.7.3 Action:

With only 1 CCW pump flow path OPERABLE, restore at least two flow paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in CSD within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SRO TS 3.5.2:

Both trains of ECCS INOPERABLE with NO applicable action statement apply TS 3.0.3.

TS 3.0.3 Action:

When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

HSB within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, HSD within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and CSD within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

DOCUMENT component manipulations per AD-OP-ALL-0204, SRO Plant Status Control.

DIRECTS RO to start the B CCW pump and stop the A CCW SRO Pump per OP-145.

If requested to remove control power from A CCW Pump Simulator Operator: either use the CCW drawing or find Remote Function:

CCW075 CP_OFF OP-145 Component Cooling Water Section 5.2 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 29 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior VERIFIES Initial Conditions and contacts Aux Operator to RO perform pre-start checks on the B CCW pump Simulator When contacted by RO - B CCW pump prestart checks Communicator: are completed and the pump is ready to be started.

NOTE: Starting the second pump could cause P fluctuations across REM-01CC-3501ASA (BSB) which may shut solenoid valves 1CC-23 and 1CC-40.

NOTE: Starting the second pump may cause flow oscillations Procedure Note:

which could shut 1CC-252. Re-opening of 1CC-252 should not be attempted until the second pump is secured.

NOTE: APP-ALB-005 Windows 1-3, 2-1, and 3-2 are expected alarms when starting the second CCW Pump.

With one CCW pump running and the standby pump capable of an automatic start, ensure a minimum flow rate of 7850 gpm exists as indicated on FI-652.1 (FI-653.1). If both CCW pumps Procedure Caution: are running OR the CCW trains are separated, a minimum of 3850 gpm per pump is required. This lower flowrate should only be allowed for short durations to accomplish pump swapping or system realignment.

Makes PA announcement that B CCW pump is about to be started. Stand clear of the pump and breaker.

Step 1: At the MCB, START CCW Pump Train B-SB.

RO

• Locates MCB start switch for B CCW pump and starts pump

• Verifies that indications are normal for the started pump.

Simulator Inform RO that B CCW pump has a good start.

Communicator:

Step 2: VERIFY flow is greater than or equal to 3850 gpm on RO FI-653.1 and FI-652.1.

RO Step 3: VERIFY OPEN, 1CC-23 and 1CC-40, REM 3501 A Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 30 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior CCW Inlet Solenoid Valve and REM 3501 B CCW Inlet Solenoid Valve respectively.

• Contacts Aux Operator to verify Acknowledge request and report back in 1 minute Simulator Communicator: 1CC-23 and 1CC-40, REM 3501 A CCW Inlet Solenoid Valve and REM 3501 B CCW Inlet Solenoid Valves are OPEN PERFORM one of the following:

RO

• SECURE a second CCW Pump using Section 7.1.

OP-145 Component Cooling Water Section 7.1 RO VERIFIES Initial Conditions.

The following Steps are written assuming shutdown of Procedure Note: Train B-SB CCW pump. If shutting down Train A-SA CCW pump, use components in parenthesis.

Step 1: VERIFY OPEN, the following valves:

• 1CC-99, CCW HEAT EXCHANGER A TO NONESSENTIAL SUP (YES)

• 1CC-113, CCW HEAT EXCHANGER B TO RO NONESSENTIAL SUP (YES)

• 1CC-127, CCW NONESSENTIAL RETURN TO HEADER B (YES)

• 1CC-128, CCW NONESSENTIAL RETURN TO HEADER A (YES)

Step 2: VERIFY SHUT, 1CC-147 and 1CC-167, CCW FROM RO RHR HEAT EXCHANGER B-SB AND A-SA (YES)

Procedure Note: If pressure falls below 52 psig, the CCW pump will restart.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 31 of 71 Event

Description:

CCW System Leak Time Position Applicants Actions or Behavior Failure of equipment to secure in the following step will result in the associated EDG being inoperable. Tech Spec Procedure Caution:

3.8.1.1 is applicable until the breaker for the affected load is opened.

Step 3: At the MCB, PLACE the control switch for CCW Pump Train A-SA to STOP AND HOLD until system pressure RO stabilizes above 52 psig.

• Stops A CCW pump (may have been completed previously)

Step 4: VERIFY the following for Train A:

• FLOW stops using FI-653.1 (FI-652.1)

RO

• PRESSURE remains greater than 75 psig suing PI-650 (PI-649).

Step 5: CHECK Train B flow rate between 10,000 and 11,000 RO gpm on MCB indicator FI-663.1. (YES)

After the crew has completed AOP-014 actions to this point, cue the Simulator Operator to insert Trigger 5, B Evaluator Cue: SG PORV opens Event 5, B SG PORV fails partially OPEN in AUTO and MANUAL Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 32 of 71 Event

Description:

B SG PORV fails partially OPEN in AUTO and MANUAL Time Position Applicants Actions or Behavior On cue from Lead Evaluator insert Trigger 5 Simulator Operator:

SG B PORV, 1MS-60 fails Open in Auto and Manual This event is a Steam Generator partially failed open. This will require the field operators to locally isolate the valve Evaluator Note: by its block valve, 1MS-61. The SRO should evaluate Tech Specs 3.3.3.5, Remote Shutdown System, and 3.6.3, Containment Isolation Valves.

Available

• ALB-014-8-5, Computer Alarm Steam Generators Indications:

• ERFIS Point ID ZMS1254A 1MS-58 SG A PORV Not Shut Actions from the APP are below but crew will most likely APP perform actions IAW AD-OP-ALL-1000, Section 5.5.13.

ALB-014 GO to page 33 for AD-OP-ALL-1000 guidance.

ERFIS alarms will not re-flash on the Annunciator Panel when elevating from a Warning to Alarm on the same point. The only indication that the alarm has changed state will be a color Procedure Note:

change from yellow to red on the alarm screen. New alarm points that come in subsequently will initiate a single re-flash of the Annunciator window and follow the same process.

CONFIRM alarm using one or more of the following ERFIS points (GD ALB 14):

BOP Point ID Description Alarm Reset

• ZMS1255A 1MS-60 SG B PORV NOT SHUT SHUT If PBD8410 is determined to be the alarm input, Flash Tank Relief Valve operation may have occurred. Continuous Calorimetric results may be unacceptable due to non-conservative program inputs.

Procedure Note: If the alarm input is a SG PORV, and that SG PORV fails to fully reset after operation, entry into the Emergency plan would be required.

A SG Blowdown line break may require entry into the Emergency Plan.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 33 of 71 Event

Description:

B SG PORV fails partially OPEN in AUTO and MANUAL Time Position Applicants Actions or Behavior CHECK instrumentation on MCB associated with the alarming point.

• DEPRESS Manual Pushbutton for PK-308B1 to take manual control of B SG PORV BOP

• LOWER output for PK-308B1 to Shut B SG PORV 1MS-60 o Output responds as desired, however 1MS-60 remains open DISPATCH an operator to check local indications associated with the alarming point.

IF contacted to investigate PORV failure acknowledge the Simulator request and report back after 2 -3 minutes that Steam is Communicator: coming from the SG PORV exhaust manifold and the valve open 5 to 10 %.

IF any SG PORV fails to fully reset after operation, THEN REFER to PEP-110, Emergency Classification and Protective BOP Action Recommendations.

• NOTIFIES SM to refer to the EAL Matrix The SRO should evaluate TS 3.3.3.5 and TS 3.6.3.

TS 3.6.3 Action:

SRO Isolate the affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The redundant manual isolation valve per PLP-106, Attachment 5 is 1MS-61.

If requested to Shut 1MS-61 insert Trigger 21 and direct the Simulator Operator: Simulator Communicator to report 1MS-61 is shut after 1 minute.

AD-OP-BOP IDENTIFIES B SG PORV is partially OPEN via status lights ALL-1000 Takes actions in accordance with AD-OP-ALL-1000 when an automatic controller malfunctions DEPRESS Manual Pushbutton for PK-308B1 to take manual BOP control of B SG PORV LOWER output for PK-308B1 to SHUT B SG PORV 1MS-60

• Output responds as desired, however 1MS-60 remains open Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 34 of 71 Event

Description:

B SG PORV fails partially OPEN in AUTO and MANUAL Time Position Applicants Actions or Behavior BOP DISPATCH operator to locally isolate B SG PORV IF contacted to investigate PORV failure acknowledge the Simulator request and report back after 2 -3 minutes that Steam is Communicator: coming from the SG PORV exhaust manifold and the valve open 5 to 10 %.

The SRO will complete OMM-001 Attachment 5 Equipment Problem Checklist for the failure of SG B PORV The SRO should evaluate TS 3.3.3.5 and TS 3.6.3.

SRO TS 3.6.3 Action:

Isolate the affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The redundant manual isolation valve per PLP-106, Attachment 5 is 1MS-61.

Contacts WCC and support personnel for repairs.

If requested to Shut 1MS-61 insert Trigger 21 and direct the Simulator Operator: Simulator Communicator to report 1MS-61 is shut after 1 minute.

After the crew has completed has stabilized the plant, cue Lead Evaluator:

Event 6, RCP A rising vibrations Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 35 of 71 Event

Description:

A RCP High Vibrations Time Position Applicants Actions or Behavior Simulator On cue from the Lead Evaluator insert Trigger 6:

Communicator: A RCP high vibration

• ALB-010-1-5, RCP-A TROUBLE Indications Available:

• A RCP vibration monitors rising and red high vibration lights lit RO Responds to alarm ALB-010-1-5.

Crew may review ALB-010-1-5 but will likely go directly to Evaluator Note:

AOP-018 when high vibration is recognized.

ENTERS and directs actions of AOP-018, Reactor Coolant Pump Abnormal Operations.

AOP-018 SRO Makes PA announcement for AOP entry Holds a crew focus brief RO Perform AOP-018 Immediate Action Immediate Action Check any CSIP running. (YES)

SRO Inform SM to refer to PEP-110 and enter the EAL Matrix.

EVALUATE plant conditions AND GO TO the appropriate SRO section:

MALFUNCTION SECTION PAGE High Reactor Coolant Pump Vibration 3.2 8 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 36 of 71 Event

Description:

A RCP High Vibrations Time Position Applicants Actions or Behavior The answer to the following question may be YES at this time but the limit will be exceeded in short order. This is a Evaluator Note: continuous action step that should be implemented when the limit is exceeded. The scenario guide is therefore written as if the limit is exceeded when the step is read.

Check all RCPs operating within limits of Att 1. (NO)

SRO/RO When answer is NO (not operating w/limits) follow below:

SRO CHECK the Reactor is TRIPPED. (NO)

TRIP the Reactor AND GO TO EOP-E-0. (Perform Steps 4 SRO through 7 as time permits.)

Directs RO to manually trip the Reactor.

Steps through immediate actions of EOP-E-0 with crew SRO Makes PA announcement for EOP entry Holds a crew focus brief Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 37 of 71 Event

Description:

A RCP High Vibrations Time Position Applicants Actions or Behavior EOP-E-0 Reactor Trip Or Safety Injection Verifies Reactor is Tripped (YES)

Immediate RO Action Verifies Turbine is Tripped - All throttle valves shut (YES)

Immediate BOP Action Immediate Verify Power To AC Emergency Buses (YES)

BOP Action AC emergency buses - BOTH energized Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 38 of 71 Event

Description:

A RCP High Vibrations Time Position Applicants Actions or Behavior Safety Injection Activated (NO)

RNO action:

Perform the following:

a) Check Safety Injection - REQUIRED (NO)

Immediate RO Action b) IF Safety Injection actuation is NOT required, THEN GO TO ES-0.1, "REACTOR TRIP RESPONSE", Step 1.

Directs RO/BOP to secure the A RCP and continue with SRO AOP-018 steps 4-7 STOPS A RCP and places PK-444C.1 to manual then shuts RO/BOP valve with demand at 0%

SRO Transitions to ES-0.1, "REACTOR TRIP RESPONSE", Step 1.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 39 of 71 Event

Description:

A RCP High Vibrations Time Position Applicants Actions or Behavior EOP-Reactor Trip Response ES-0.1 Procedure Note: Foldout applies SRO Assigns foldout items of ES-0.1.

Evaluator Aide:

SRO Initiate Monitoring Of Critical Safety Function Status Trees.

SRO Evaluate EAL Matrix.

Check RCS Temperature:

a. Check RCPs - ANY RUNNING (YES - B and C SRO / b. Check SG blowdown isolation valves shut (NO)

BOP Shut SG blowdown FCVs:

• 1BD-18 (FCV-8405A) (SHUTS)

• 1BD-37 (FCV-8405B) (SHUTS)

• 1BD-56 (FCV-8405C) (SHUTS)

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6 Page 40 of 71 Event

Description:

A RCP High Vibrations Time Position Applicants Actions or Behavior Since the A was secured RCS temperature will continue Evaluator Note: to drop. The crew will most likely shut the MSIVs here.

After the MSIVs are shut RCS temperature will recover.

Stabilize AND Maintain Temperature Between 555°F AND 559°F using Table 1.

BOP Informs CRS of cooldown then shuts ALL MSIVs While the crew is stabilizing the plant after the MSIVs are shut AND the crew sees that RCS temperature is stable or Evaluator Note:

increasing then insert Event 7 Small Break LOCA inside Containment Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 41 of 71 Event

Description:

Small Break LOCA Time Position Applicants Actions or Behavior Simulator On cue from the Lead Evaluator insert Trigger 7:

Communicator: Small Break LOCA Identifies changing Primary plant conditions and recommends SI based on fold out criteria of EOP-ES-0.1 for rapidly RO / BOP degrading Subcooling approaching the setpoint and Pressurizer level will not being able to be maintained > 5%

SRO Directs RO to actuate Safety Injection RO Manually actuates Safety Injection Re-enters E performs a crew alignment brief then has crew verify:

• Reactor Trip (YES)

• Turbine Trip (YES)

EOP-E-0 SRO

• AC emergency buses energized (YES)

• Safety Injection - Actuated (Both Trains) (YES)

SRO Assigns foldout items of E-0.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 42 of 71 Event

Description:

Small Break LOCA Time Position Applicants Actions or Behavior Evaluator Aide:

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 7 Page 43 of 71 Event

Description:

Small Break LOCA Time Position Applicants Actions or Behavior Identifies Containment Adverse Conditions CREW Containment Pressure > 3 psig SRO Evaluate EAL Matrix.

Verifiy CSIPs - all running (YES)

RO A and B running Verify RHR Pumps - all running (YES)

RO A and B running RO Safety Injection flow > 200 gpm (NO)

Perform the following:

a) Verify high head safety injection alignment:

(1) CSIP suction from RWST valves - OPEN 1CS-291 (LCV-115B) (YES) 1CS-292 (LCV-115D) (YES)

(2) VCT outlet valves - SHUT 1CS-165 (LCV-115C) (YES) 1CS-166 (LCV-115E) (YES)

RO (3) Charging line isolation valves - SHUT 1CS-235 1CS-238 (4) BIT outlet valves - OPEN 1SI-3 (NO- under clearance) 1SI-4 (NO - unknown why)

ATTEMPTS TO OPEN 1SI-4 (valve will NOT open)

Informs SRO 1SI-4 will not OPEN Directs RO actions when high head safety injection flow path can NOT be aligned. Establish any other high head injection SRO flow path (listed in order of preference):

• Directs RO to OPEN 1SI-52 SA Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 8, 9 Page 44 of 71 Event

Description:

SI-4 failure, RCP B and C manual Trip Time Position Applicants Actions or Behavior Locates MCB controls for 1SI-52 SA and turns on control power and takes switch to OPEN Event 8 Informs SRO that 1SI-52 SA is OPEN.

RO Critical to establish SI flow of > 200 gpm using alternate Critical high head safety injection to cold legs prior to securing Task # 1 RCPs Identifies that Safety Injection flow is now exceeding 200 gpm RCS pressure may be < 1400 psig by this point in the scenario. It may not be yet depending on the crews progression through the scenario. When the crew Evaluators Note:

identifies that SI flow is > 200 gpm and RCS pressure is

< 1400 psig they will secure RCPs IAW E-0 RCP trip criteria.

Identifies that RCP trip criteria is met based on SI flow > 200 Event 9 gpm with RCS pressure < 1400 psig and informs the SRO that RCP trip criteria is met and secures both RCP B and RCP C RO Critical Critical to secure RCPs within 10 minutes of reaching RCP Task # 2 Trip criteria of SI flow > 200 gpm with RCS pressure < 1400 psig prior to exiting E-0 SRO RCS Pressure - LESS THAN 230 PSIG (NO)

Main Steam Line Isolation - ACTUATED

• NO - automatic MSLI is failed SRO Directs crew to actuate MSLI Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 45 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Actuates MSLI Verifies MSIVs and Bypass Valves are SHUT Event 10 RO / BOP The crew should identify that the MSIV before seat drain valves 1MS-231, 1MS-266 and 1MS-301 have failed to shut and SHUT each valve.

Any SG pressure - 100 PSIG LOWER THAN PRESSURE IN BOP TWO OTHER SGs (NO)

When Containment Spray is actuated a Phase B actuation signal will also be generated. Depending on the crews pace through the procedures they may reach this Evaluator Note:

point with RCPs still in operation and RCS pressure above the E-0 fold out criteria for tripping RCPs. IF the crew has not secured RCPs at this point they will now.

Check CNMT Pressure - HAS REMAINED LESS THAN 10 PSIG (YES / NO - it will exceed 10 psig)

RO

• Verify CNMT spray - ACTUATED (YES)

• Stop all RCPs o Locates MCB switches for RCPs and STOPS B and C RCP (if not already done)

If SG Levels are high, RNO will be to establish greater than 40% level in all SGs.

Evaluator Note:

High SG levels may be expected due to low power/steam demand at time of trip.

BOP Verify AFW flow - AT LEAST 210 KPPH ESTABLISHED (YES)

Sequencer Load Block 9 (Manual Loading Permissive) -

BOP ACTUATED (BOTH TRAINS) (YES)

BOP Energize AC buses 1A1 AND 1B1 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 46 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior The actions for EOP-E-0 Attachment 3 are listed in Attachment 2 in the back of this scenario guide on page 63.

The RO will perform all board actions until the BOP completes Attachment 3. The BOP is permitted to properly Evaluator Note:

align plant equipment IAW EOP-E-0 Attachment 3 without SRO approval.

The Scenario Guide still identifies tasks by board position because the time frame for completion of Attachment 3 is not predictable.

Verify Alignment Of Components From Actuation Of ESFAS BOP Signals Using Attachment 3, "Safeguards Actuation Verification", While Continuing With This Procedure.

Directs AO to place 1A and 1B Air Compressor in the local BOP control mode per EOP-E-0 Attachment 3 step 22 Acknowledge the request to place 1A and 1B Air Simulator Compressor in the local control mode per E-0 Attachment Communicator:

3 step 22 When directed to place the 1A and 1B Air Compressor in Simulator the local control mode:

Communicator: Run APP\air\acs_to_local When the APP for 1A and 1B Air Compressor has Simulator completed running call the MCR and inform them that the Communicator:

air compressors are running in local control.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 47 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Directs AO to locally unlock AND turn ON the breakers for the CSIP suction AND discharge cross-connect valves, referring to E-0, Attachment 3, step 23.

BOP Acknowledge request to unlock and turn on the breakers Simulator for the CSIP suction and discharge cross-connect valves Communicator:

E-0, Attachment 3, step 23.

When requested to unlock and turn on CSIP suction and discharge cross-connect valves: Run APP\cvc\E-0 Att 3 Simulator CSIP suct & disch valve power Communicator:

When the APP has completed running inform MCR that E-0, Attachment 3, step 23 is complete.

RCPs are secure therefore WR CL temperatures should be used when checking RCS temperature. RCS temp trend Evaluator Note:

will be < 557° and dropping - control FF, maintain total FF

> 210 KPPH until SG level > 40% (all MSIVs are shut)

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 48 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Stabilize AND Maintain Temperature Between 555°F AND 559°F Using Table 1.

RO PRZ PORVs - SHUT (YES)

RO PRZ Spray Valves - SHUT (YES - RCPs are secured)

PRZ PORV Block Valves - AT LEAST ONE OPEN (YES)

• Any SG pressure - DROPPING IN AN UNCONTROLLED MANNER OR COMPLETELY DEPRESSURIZED (NO)

• Any SG - ABNORMAL RADIATION (NO)

SRO OR UNCONTROLLED LEVEL RISE (NO)

• CNMT Pressure - NORMAL NO - GO TO E-1, "LOSS OF REACTOR OR SECONDARY COOLANT", Step 1.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 49 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior EOP-E-1 Loss Of Reactor or Secondary Coolant Procedure Note: Foldout applies Performs focus brief with crew SRO Assigns foldout items of EOP- E-1.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 50 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Evaluator Aide: EOP-E-1 Foldout Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 51 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior SRO Initiate Monitoring Of Critical Safety Function Status Trees.

Maintain RCP Seal Injection Flow Between 8 GPM And 13 RO GPM.

• Check Intact SG Levels: Any level - > 25% [40%]

(YES / NO depends on monitoring and control - should be YES IF NO then Maintain total FF > 210 KKPH until level > 40% in BOP at least 1 intact SG)

• Control feed flow to maintain all intact levels between 25%

And 50% [40% And 50%].

• Any level - Rising in an uncontrolled manner (NO)

Check PRZ PORV AND Block Valves:

• Verify AC buses 1A1 AND 1B1 - ENERGIZED (YES)

RO

• Check PRZ PORVs - SHUT (YES)

• Check block valves - AT LEAST ONE OPEN (YES)

• IF a PRZ PORV opens on high pressure, THEN verify it shuts after pressure drops to less than opening setpoint.

Check SI Termination Criteria:

RO RCS subcooling - > 10°F [40°F] - C 20°F [50°F] - M (NO)

Check CNMT Spray Status:

SRO

• Check any CNMT spray pump - RUNNING (YES)

• Consult plant operations staff to determine if CNMT spray should be placed in standby.

Simulator IF contacted for CNMT spray pump evaluation tell CRS that Communicator: at this time leave the CNMT spray pumps running.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 52 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Check Source Range Detector Status:

Intermediate range flux - < 5x10-11 AMPS (YES)

RO Verify source range detectors - ENERGIZED (YES)

Transfer nuclear recorder to source range scale.

(Transfers NR-45 to source range scale)

Check RHR Pump Status:

Check RHR pump suction - ALIGNED TO RWST RWST Suction OPEN RO

• RHR A: 1SI-322 (YES)

• RHR B: 1SI-322 (YES)

RCS Pressure - > 230 psig (YES)

RCS pressure - STABLE OR RISING (NO)

Check RCS And SG Pressures:

BOP Check for both of the following:

/ RO All SG pressures - Stable or Rising (YES)

RCS pressure - Stable or Dropping (YES)

Establish CCW Flow To The RHR Heat Exchangers:

• Verify both CCW pumps - RUNNING (No only B CCW)

RO

• Open 1CC-167

• Verify CCW flow to the RHR heat exchanger Check EDG Status: Check AC emergency buses 1A-SA AND 1B-SB - ENERGIZED BY OFFSITE POWER (YES)

Check bus voltages BOP Check breakers 105 and 125 CLOSED (YES)

Check any EDG - RUNNING UNLOADED (YES)

Evaluator Note: EOP-FR-P.1 may be entered. Implementation will not be required with RCS pressure < 230psig and RHR HX Flows

> 1000gpm.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 53 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior RO Reset SI Manually realign safeguards equipment following a loss of BOP offsite power Shutdown any unloaded EDGs using OP-155 section 7 Simulator Acknowledge the request, state that you are heading out to Communicator: the EDGs and will call back when you are there.

Initiate Evaluation Of Plant Status:

• RHR system - CAPABLE OF COLD LEG RECIRCULATION (YES)

• Check auxiliary AND radwaste processing building radiation

- NORMAL (YES)

SRO Check RCS Status:

Check for both of the following:

• RCS pressure - LESS THAN 230 PSIG (NO)

• Any RHR HX header flow - GREATER THAN 1000 GPM (NO)

GO TO ES-1.2, "POST LOCA COOLDOWN AND DEPRESSURIZATION", Step 1.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 54 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior EOP-POST LOCA COOLDOWN AND DEPRESSURIZATION ES-1.2 Implements EOP-ES-1.2 SRO Performs crew alignment brief

• Reset SI

• Manually Realign Safeguards Equipment Following A Loss Of Offsite Power.

• Reset Phase A AND Phase B Isolation Signals.

RO

• Open Instrument Air AND Nitrogen To CNMT:

o 1IA-819 o 1SI-287 Monitor AC Buses:

• Check AC emergency buses 1A-SA AND 1B-SB -

ENERGIZED BY OFFSITE POWER (YES)

• Check bus voltages BOP

• Check breakers 105 and 125 CLOSED (YES)

• Check all non-emergency AC buses - ENERGIZED (YES)

PRZ heaters should NOT be energized until PRZ water level Procedure Caution indicates greater than minimum recommended by plant operations staff to ensure heaters are covered.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 55 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Secure PRZ Heaters:

• Place backup heaters in the OFF position.

• Verify control heaters - OFF (YES)

• Consult Plant Ops Staff for recommended minimum PRZ water level to keep heaters covered Check RHR Pump Status: (OFF)

RO

• Aligned to RWST (YES)

• RCS Pressure > 230 psig (YES)

• RCS Pressure - stable or increasing (NO)

• Check RHR pump suction aligned to RWST (YES)

• Stop RHR pumps At some point during the implementation of EOP-ES-1.2 the break will clear and the Safety Injection flow filling the RCS with cold RWST water will cause pressure and temperature reduction.

Soon afterward the pressure will decrease to < 650 psig Evaluator Note: allowing the Safety Injection Accumulators to inject into the RCS. The injection will cause further temperature and pressure reductions. The critical safety function status tree for RCS integrity will begin to toggle from Green to Yellow to Orange to Red. Eventually RCS Integrity will remain RED and the crew will transition to EOP-FR-P.1

• Check Intact SG Levels: Any level - GREATER THAN 25%

[40%] (YES)

BOP

• Control feed flow to maintain all intact levels between 25%

and 50% [40% and 50%].

After the low steam pressure SI signal is blocked, main Procedure Note: steamline isolation will occur if the high steam pressure rate setpoint is exceeded.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 56 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior

• Check PRZ Pressure: Pressure - LESS THAN 2000 PSIG (YES) o Block low steam pressure SI RO o Initiate RCS Cooldown To Cold Shutdown: Maintain cooldown rate in RCS cold legs - LESS THAN 100°F/HR EOP Response to Imminent Pressurized Thermal Shock FR-P.1 SRO Implements EOP-FR-P.1 Performs crew alignment brief Foldout applies Assigns RO and BOP foldout actions SRO

• RO - None

• BOP - AFW Supply Switchover criteria, Cold Leg Recirculation Switchover criteria Evaluator Aide: EOP-FR-P.1 Foldout Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 57 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Check RCS Pressure:

Check for both of the following:

• RCS pressure - LESS THAN 230 PSIG RO (NO / YES - it will be soon)

• Any RHR HX header flow - > 1000 GPM RO restarts RHR pumps when RCS pressure < RHR shutoff head - EOP-ES-1.2 foldout action item Check RCS Cold Leg Temperature Trend:

RO

• Check RCS Cold Leg Temperatures - STABLE OR RISING (NO)

Procedure Note: A faulted SG is any SG that is depressurizing in an uncontrolled manner or is completely depressurized.

Stop RCS Cooldown:

Verify SG PORVs - SHUT (YES)

BOP Verify condenser steam dump valves - SHUT (YES)

Check RHR system - IN SHUTDOWN COOLING MODE (NO)

Any non-faulted SG level - > 25% [40%] (YES)

Control feed flow to non-faulted SG(s) to stop RCS cooldown.

IF the TDAFW pump is the only available source of feed flow, Procedure Caution: THEN maintain steam supply to the TDAFW pump from one SG.

BOP Minimize RCS Cooldown From Faulted SG(s):

Check any SG - FAULTED (NO)

Check PRZ PORV Block Valves:

RO

• Verify power to block valves - AVAILABLE (YES)

• Check block valves - AT LEAST ONE OPEN (YES)

Procedure Note: IF PRZ PORV opens on high pressure, Step 6 should be repeated after pressure drops to less than PORV setpoint.

Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 58 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Check PRZ PORVs:

Check all of the following:

• Check LTOPS control switches - IN NORMAL (NOT RO BLOCKED) (NO - BLOCKED)

• Check PRZ pressure - < 2335 psig (YES)

• Verify PRZ PORVs - SHUT (YES)

RO Check SI Flow - > 200 gpm (YES)

Check SI Termination Criteria:

SRO Check for both of the following:

RCS subcooling - > 60°F [90°F] - C (NO)

Procedure Caution: Following a complete loss of normal seal cooling, the affected RCP(s) should NOT be started prior to a status evaluation.

Check If An RCP Should Be Started:

SRO RCS subcooling - GREATER THAN 10°F [40°F] - C (NO)

Go to step 32 Following an excessive cooldown, reactor vessel stress must Procedure Caution: be relieved to enhance and maintain vessel integrity. Do NOT perform any actions that raise pressure OR cause an RCS cooldown until the soak is complete.

Even if a soak period is required, steam may be released from Procedure Note: intact SGs with pressure higher than the saturation pressure for lowest cold leg temperature.

Determine RCS Soak Requirements:

RCS cooldown rate - > 100°F in any 60 min period Perform one hour RCS soak:

SRO

• Maintain RCS temperature stable.

• Maintain RCS pressure stable.

• Perform actions of other procedures that do NOT cause an RCS cooldown OR raise pressure.

Examiners Note: END OF SCENARIO Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Operator Action Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 10 Page 59 of 71 Event

Description:

MSLI failure Time Position Applicants Actions or Behavior Direct the Simulator Operator to place the Simulator to FREEZE Lead Evaluator Announce CREW UPDATE - The Exam Team has the shift. Inform the crew to remain seated at their desk and to not discuss the scenario.

Simulator Operator When directed by the Lead Evaluator go to FREEZE Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 OP-134, Section 5.6 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 OP-134, Section 5.6 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 OP-134, Section 5.6 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 EOP-E-0 Attachment 3 Harris 2016 NRC EXAM Scenario 3 Rev. 2

Appendix D Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # N/A Page 71 of 71 Event

Description:

Scenario Guide Revision Summary Rev. 0 Initial Development Rev. 1 NRC D-1 Outline comments incorporated Rev. 2 Operation validation comments incorporated Rev. 3 NRC 60 day submittal comments incorporated Rev. 4 NRC Prep Week comments incorporated Rev. Final Approved for administration by NRC Region II Harris 2016 NRC EXAM Scenario 3 Rev. 2

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 001/NEW/C/A///EOP-ECA-1.1 ATT 1//011EG2.1.25/

Given the following plant conditions:

- At time 0704, the plant is operating at 100% power

- 'A' RHR pump is under clearance Time 0706 PZR level has lowered to 56%, Containment pressure and radiation readings are rapidly rising 0707 The OATC attempts to manually trip the Reactor but neither Reactor Trip switch opens the Reactor Trip breakers 0710 Containment pressure is 26 psig and rising 0715 The Turbine Building AO manually opened the 'A' and 'B' MG Set Output Breakers and all rods insert into the Reactor 0728 The CRS transitions to EOP-E-1, Loss of Reactor or Secondary Coolant 0733 RHR Pump 'B' trips on overcurrent 0736 The CRS transitions to EOP-ECA-1.1, Loss of Emergency Coolant Recirculation 0744 The CRS is at step 19.c, determine minimum SI flow from Attachment 1 to establish the minimum SI flow needed.

Which ONE of the following (1) represents the minimum SI flow REQUIRED in EOP-ECA-1.1 Attachment 1 AND (2) the reason for calculating this minimum SI Flow?

(Reference Provided)

A. (1) 400 gpm (2) to ensure the existence of an adequate Reactor Vessel inventory such that core cooling is ensured B. (1) 400 gpm (2) to match decay heat in order to further decrease SI pump flow and delay RWST depletion.

C. (1) 425 gpm (2) to ensure the existence of an adequate Reactor Vessel inventory such that core cooling is ensured.

D. (1) 425 gpm (2) to match decay heat in order to further decrease SI pump flow and delay RWST depletion.

Thursday, May 19, 2016 5:04:41 PM 203

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: At 0707 a valid Reactor Trip signal was generated but the Reactor Trip did not occur until 0715. The time that the CRS should use for "Time After Reactor Trip" is 29 minutes (0744 to 0715) and NOT the time that a Reactor Trip signal was generated but was ineffective in opening the Reactor Trip breakers. Using EOP ECA-1.1 Attachment 1 the time after a Reactor Trip of 29 minutes falles between the 25 to 30 minute area which will require a minimum SI flow of 425 gallons. IF the CRS used time from Reactor Trip signal (0707) a lower flow rate of 400 gallons would be used and would NOT satisfy the minimum SI flow rate required needed to remove decay heat.

The WOG description for EOP ECA-1.1 when determining the minium SI flow states that the reason for calculating this minimum SI flow is to match decay heat in order to further decrease SI pump flow and delay RWST depletion.

A. Incorrect. The first part is plausible if the candiate uses the time that a Reactor Trip signal was generated which would yeild a time of 37 minutes after a Reactor Trip and a minimum SI flow from Attachment 1 of 400 gpm. The second part is plausible since this is backgound document information for a Large Break LOCA but this reason is for RVLIS SI termination criteria.

B. Incorrect. The first part is plausible see A(1). The second part is correct.

C. Incorrect. The first part is correct. The second part is plausible see A(2).

D. Correct.

Thursday, May 19, 2016 5:04:41 PM 204

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000011 Large Break LOCA / 3 011EG2.1.25; Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10 / 43.5 / 45.12 )

Importance Rating: RO 3.9 SRO 4.2 Technical

Reference:

EOP-ECA-1.1 Attachment 1, Rev. 0, Page 52 WOG Background document for EOP-ECA-1.1 References to be provided: EOP-ECA-1.1, Attachment 1 Learning Objective: EOP-LP-2.3 Objective 5.d Question Origin: New Comments: None Tier/Group: T1G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item.

Thursday, May 19, 2016 5:04:41 PM 205

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 002/BANK/FUNDAMENTAL//T.S. 3.9.8.1/NONE//025AG2.2.25/

Given the following plant conditions:

- Plant is in Mode 6

- Refueling Cavity Level is at 23 6

- Both trains of RHR are in service for Shutdown Cooling

- 'B' EDG is under clearance for scheduled maintenance Subsequently:

- A Loss of Offsite Power occurs

- 'A' EDG starts and the 'A' Sequencer reaches Load Block 9 Which ONE of the following completes the statements below?

The MINIMUM action required to comply with Technical Specification 3.9.8.1 -

Refueling Operations: Residual Heat Removal And Coolant Circulation - High Water Level is to start the 'A' RHR pump (1) .

The basis of the LCO is to ensure that sufficient cooling capacity is available to maintain the RCS below (2) .

A. (1) ONLY (2) 140°F B. (1) ONLY (2) 200°F C. (1) AND restore power to the 'B' RHR Pump (2) 140°F D. (1) AND restore power to the 'B' RHR Pump (2) 200°F Thursday, May 19, 2016 5:04:41 PM 206

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: To comply with Technical Specification 3.9.8.1 LCO the 'A' RHR Pump must be started. The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the Reactor vessel below 140°F as required during the REFUELING MODE. and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible because 200°F is the transition to Mode 4 where additional concerns arise.

C. Incorrect. The first part is plausible if the candidate misapplies the LCO mode of applicability and determines the actions for Mode 5 or lower cavity level were applicable which would require two RHR pump to be operable and one in operation. The second part is correct.

D. Incorrect. The first part is plausible if the candidate misapplies the LCO mode of applicability and determines the actions for Mode 5 or lower cavity level were applicable which would require two RHR pump to be operable and one in operation. The second part is plausible because 200°F is the transition to Mode 4 where additional concerns arise.

Thursday, May 19, 2016 5:04:41 PM 207

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000025 Loss of RHR System / 4 025AG2.2.25; Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

(CFR: 41.5 / 41.7 / 43.2)

Importance Rating: RO 3.2 SRO 4.2 Technical

Reference:

Tech Spec 3.9.8.1 pg 3/4 9-9 (page 339)

Tech Spec Bases 3/4.9.8 pg B 3/4 9-2 (page 96)

References to be provided: None Learning Objective: Lesson Plan RHR System, Objective 9.f Question origin: Bank Comments: None SRO justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know the Technical Specification Bases of the limiting conditions of operation during refueling operations.

Requires knowledge of Technical Specification Bases that are not system knowledge.

Thursday, May 19, 2016 5:04:41 PM 208

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 003/NEW/C/A//AOP-014/NONE/EARLY/026AA2.02/SAT Given the following plant conditions:

- The unit is operating at 95% power beginning of life

- RCS boron concentration is 1391 ppm

- Tavg is 585.5°F

- 'A' RHR Pump is being run for quarterly surveillance test on recirc to the RWST Subsequently:

- Tavg is 586.1°F

- An automatic diversion of the VCT to the RHT is in progress

- CCW Surge Tank level is 15% and lowering

- 1DW-15, CCW Makeup, is open in accordance with AOP-014, Loss of Component Cooling Water Which ONE of the following identifies (1) the location of the leak AND (2) the procedure direction(s) required to for this event?

A. (1) 'A' RHR Heat Exchanger (2) Check RAB/Containment Sumps for rising level B. (1) 'A' RHR Heat Exchanger (2) Direct Chemistry to sample the 'A' RHR Heat Exchanger for corrosion inhibitors C. (1) Seal Water Return Heat Exchanger (2) Locally isolate the CCW side of the Seal Water Return Heat Exchanger D. (1) Seal Water Return Heat Exchanger (2) Locally bypass and isolate the Seal Water side of the Seal Water Return Heat Exchanger Thursday, May 19, 2016 5:04:41 PM 209

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: With a CCW System leak in progress the candidate must evaluate the response of the RCS Temperature combined with the rise in VCT level to determine that a dilution event has occurred. Once the determination that a dilution event has occurred the candidate must compare the heat exchanger operating characteristics and determine that a leak in the seal water return heat exchanger will result in the transfer of non-borated CCW water into the borated seal water system.

Finally the candidate must recall the mitigating action directed by AOP-014 to determine the proper sequence of actions required to isolated the non-borated water source.

A. Incorrect. The first part is plausible since the 'A' RHR system is in service for surveillance testing; however this is incorrect because the 'A' RHR system is aligned to the RWST vice the RCS which is physically isolated by check valves when the unit is on-line. The second part is plausible since these would be correct actions if the RHR system pressure is less than the CCW system pressure; however this is incorrect because the RHR to CCW system pump pressure differential would result in leakage into the CCW system causing CCW surge tank level to rise.

B. Incorrect. The first part is plausible since the 'A' RHR system is in service for surveillance testing; however this is incorrect because the 'A' RHR system is aligned to the RWST vice the RCS which is physically isolated by check valves when the unit is on-line. The second part is plausible see A2.

C. Incorrect. The first part is correct. The second part is plausible since the candidate has determined the source of the leak from the CCW system is the seal water return HX and isolating the CCW side of the HX will isolate the leak and stop the inadvertent dilution; however this is incorrect because the seal return is required to be bypassed firsted to ensure a flowpath is maintiained while seal return is in service.

D. Correct.

Thursday, May 19, 2016 5:04:41 PM 210

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000026 Loss of Component Cooling Water / 8 026AA2.02; Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The cause of possible CCW loss (CFR: 43.5 / 45.13)

Importance Rating: RO 2.9 SRO 3.6 Technical

Reference:

AOP-014, Section 3.2 Step 16, Rev. 37, Page 23, 24 References to be provided: None Learning Objective: AOP-LP-3.14, objective 3 Question Origin: New Comments: None Tier/Group: T1G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of diagnostic steps and decision points in the abnormal operating procedures (AOP) that actions to mitigate an event using specific procedure sections.

Thursday, May 19, 2016 5:04:41 PM 211

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 004/NEW/C/A//EOP-ES-0.2/NONE//056AA2.76/

Given the following conditions:

- The unit is operating at 100% power

- 'A' Reactor Water Makeup Pump is in operation Subsequently a loss of Offsite Power occurs

- RCS cooldown to < 200°F will be required Which ONE of the following completes the statements below concerning the operation of the Reactor Water Makeup Pumps?

The 'A' RW Makeup Pump (1) .

To prevent an inadvertant RCS dilution event the standby Reactor Makeup Water Pump breaker must be opened and placed under clearance in accordance with (2) prior to reducing RCS temperature below 200°F.

A. (1) re-starts automatically during sequencer operation (2) EOP-ES-0.2, Natural Circulation Cooldown B. (1) re-starts automatically during sequencer operation (2) GP-007, Normal Plant Cooldown Mode 3 to Mode 5 C. (1) must be manually re-started (2) EOP-ES-0.2, Natural Circulation Cooldown D. (1) must be manually re-started (2) GP-007, Normal Plant Cooldown Mode 3 to Mode 5 Thursday, May 19, 2016 5:04:41 PM 212

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The Reactor Makeup Water (RMWU) Pumps are powered by 1A24 and 1B24 which is energized from safety busses. The 480V loads from the emergency bus is manually energized after the sequencer completes running. The RWMU pumps will operate the same as they did prior to the LOOP once power is restored to their 480V bus.

EOP-ES-02, Step 25 requires that only one RMWU Pump is operable and the standby pump breaker is opened and placed under clearance for inadvertent dilution prevention.

A. Incorrect. The first part is plausible since the VCT is the normal suction source of a Charging pump that is sequenced on in load block 1 for all 3 sequencer programs. During a loss of power to a safety bus the sequencer runs in program "A". There isn't a Safety Injection during Program "A" and the normal suction source to the Charging pumps is the VCT. VCT makeup is provided by the RMWU pump and BA pump. It is plausible to have both the BA pump and RMWU pumps sequenced on to provide makeup to the VCT. But neither pump has power until after a manual 480V breaker for emergency loads is closed. The second part is correct.

B. Incorrect. The first part is plausible see A(1). The second part is plausible since the RCPs do not have power and a Natural Circ Cooldown would be in progress to reduce RCS temperature to < 200°F and GP-007 provides guidance to place the RMWU pumps under clearance during a normal shutdown. However in accordance with EOP ES-0.2 step 25 the procedure will place the RMWU pumps under clearance to prevent an inadvertant dilution event and refer to GP-007 for additional action.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible see B(2).

000056 Loss of Off-site Power / 6 056AA2.76; Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Reactor makeup water pump (running)

(CFR: 43.5 / 45.13)

Importance Rating: RO 2.6 SRO 2.6 Technical

Reference:

EOP-ES-0.2 Step 25, Rev. 0, Page 42 OMM-004, Rev 38, Page 62 Thursday, May 19, 2016 5:04:41 PM 213

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal References to be provided: None Learning Objective: Student Text SEQ, Objective 4 GP-LP-3.7, Objective 1 Question Origin: New Comments: Discuss K/A match with Mike based on RMWU Pump not running post LOOP and SRO match.

On 3-23-2016 Mike Donithan concurred with HNP development of this question with SRO procedure selection and RWMU pump power restoration as acceptable for SRO level of knowledge and K/A match.

Phonecon 3/23: I somewhat misunderstood HNPs position on this K/A: they have the framework of a question that sounds like it will work. I committed that if the question as-submitted follows that plan then I will deem it an acceptable K/A match.

So keep K/A 056AA2.76 Tier/Group: T1G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item. Additionally the candidate must evaluate diagnostic steps and decision points in the normal operating procedures (OP) that actions to mitigate an event using specific procedure sections.

Thursday, May 19, 2016 5:04:41 PM 214

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 005/NEW/C/A//EOP-USERS GUIDE/NONE/EARLY/065AG2.4.8/SAT Given the following plant conditions:

- The plant is operating at 100% power

- At 1015, ALB-002-8-1, Instrument Air Low Pressure, alarms and the crew enters AOP-017, Loss of Instrument Air Subsequently the following indications are observed:

Time IA Pressure SG Levels 1016 73 psig 57%

1017 65 psig 54%

1018 58 psig 41%

1019 37 psig 28%

Which ONE of the following identifies (1) the FIRST time the Reactor is REQUIRED to be tripped in accordance with AOP-017 AND (2) the appropriate plant procedure(s) to be implemented?

Procedure Titles:

EOP-E-0, Reactor Trip Or Safety Injection AOP-017, Loss Of Instrument Air A. (1) 1018 (2) ONLY EOP-E-0 B. (1) 1018 (2) EOP-E-0 AND AOP-017 C. (1) 1019 (2) ONLY EOP-E-0 D. (1) 1019 (2) EOP-E-0 AND AOP-017 Thursday, May 19, 2016 5:04:41 PM 215

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with AOP-017, the Operator is required to maintain levels greater than 30% and Main Feedwater flow to ALL Steam Generators.

FW regulating valves receive a shut signal when instrument air pressure fails to 60 psig. At 1040 the instrument air header pressure of 58 psig therefore the operator is no longer able to maintain Main Feedwater flow to all SG's. Without MFW to the S/G's AOP-017 directs the operator to insert a manual Reactor trip and perform EOP-E-0 while continuing with AOP-017. The operator is required to evaluate the event and determine that a loss of instrument air does not result in an actuation of the Safety Injection system. EOP-ES-0.1, Reactor Trip Response is the appropriate EOP to transition into after entering EOP-E-0 while continuing with AOP-017.

A. Incorrect. The first part is correct. The second part is plausible since without an actuation of the Safety Injection system the correct procedure to implement is EOP-ES-0.1; however this in not correct since EOP-ES-0.1 alone does not address the loss of instrument air system and AOP-017 directs the CRS to continue with the implementation of the AOP while entering the EOP network in response to the Reactor Trip. Additionally the EOP-User's Guide allow the implementation of AOP's that enhance plant control.

B. Correct.

C. Incorrect. The first part is plausible since the Reactor is required to be tripped once SG levels cannot be maintained above 30%; however this is incorrect since it is not the FIRST time that a Reactor trip is required. The second part is plausible since without an actuation of the Safety Injection system the correct procedure to implement is EOP-ES-0.1; however this in not correct since EOP-ES-0.1 alone does not address the loss of instrument air system and AOP-017 directs the CRS to continue with the implementation of the AOP while entering the EOP network in response to the Reactor Trip. Additionally the EOP-User's Guide allow the implementation of AOP's that enhance plant control.

D. Incorrect. The first part is plausible since the Reactor is required to be tripped once SG levels cannot be maintained above 30%; however this is incorrect since it is not the FIRST time that a reactor trip is required. The second part is correct.

Thursday, May 19, 2016 5:04:41 PM 216

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000065 Loss of Instrument Air / 8 065AG2.4.8; Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: RO 3.8 SRO 4.5 Technical

Reference:

AOP-017, Section 3.0, Step 1, Rev 40, Page 4 EOP-User's Guide, Section 5.0, Step 5.1.2, Rev 45, Page 12 References to be provided: None Learning Objective: AOP-LP-3.17, Objective 3 Question Origin: New Comments: None Tier/Group: T1G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of diagnostic steps and decision points in the abnormal operating procedures (AOP) and the EOP-User's Guide user rules of procedure implentation that addresses the actions to mitigate an event using EOP and AOP procedure sections.

Thursday, May 19, 2016 5:04:41 PM 217

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 006/PREVIOUS/C/A//EOP-E-0/NONE//WE04EA2.1/

Given the following plant conditions:

- A Reactor Trip and Safety Injection has occured

- EOP-E-0, Reactor Trip Or Safety Injection, is being implemented and SI has been reset The current conditions are as follows:

- RCS Pressure is 1500 psig

- PRZ level is off scale low

- Subcooling is 3°F

- Containment pressure 0.2 psig

- RM-1RR-3597, RHR Pump 1B, is in HIGH alarm

- MLB-4A-SA-6-3 and MLB-4B-SB-6-3, RAB Equip C/D Sump Alert Lvl, status lights are lit

- SG levels are: A = 23%, B = 24%, C = 15%

- Total AFW flow has been reduced to 215 KPPH Which ONE of the following procedures will be implemented when exiting EOP-E-0?

A. EOP-ES-1.1, SI Termination B. EOP-ECA-1.2, LOCA Outside Containment C. EOP-FR-H.1, Response to Loss of Secondary Heat Sink D. EOP-ES-1.2, Post LOCA Cooldown and Depressurization Thursday, May 19, 2016 5:04:41 PM 218

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The transition to ECA-1.2 is correct. The Radiation Monitor in alarm and sump level alert lights indicate that the leak is in the B RHR Pump Room.

LOCA outside containment. Transition to ECA-1.2 would occur at step 60 of E-0.

A. Incorrect. Plausible if the candidate misundertands SI termination criteria and determines that positive subcooling satisfies the criteria (+ 3°F and improving), however this is incorrect because PZR level must also be above 10% and the current indication is offscale low which does not meet the requirement for SI termination.

B. Correct.

C. Incorrect. Plausible since S/G levels are all less than 25%, which meet FR-H.1 entry conditions (Containment conditions normal), however this is not correct because total feed flow must also be less than 210 KPPH.

D. Incorrect. Plausible since this is the procedure that would be implemented for the question conditions if Auxiliary Building radiation levels were normal, however this is not correct because the RHR B pump room rad monitor is in High alarm and both channels of the RAB equipment drain sump are in Alert Alarm indicating that the sump level is rising.

Thursday, May 19, 2016 5:04:42 PM 219

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal W/E04 LOCA Outside Containment / 3 WE04EA2.1; Ability to determine and interpret the following as they apply to the (LOCA Outside Containment): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CFR: 43.5 / 45.13)

Importance Rating: RO 3.4 SRO 4.3 Technical

Reference:

EOP-E-0, Step 60, Rev 4, Page 48 References to be provided: None Learning Objective: EOP-LP-2.3/3.3 Objective 1.d Question Origin: Previous 2014 NRC SRO Exam 81 randomly selected Comments: None Tier/Group: T1/G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.

Thursday, May 19, 2016 5:04:42 PM 220

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 007/NEW/C/A//EOP-E-0 AND TS3.3.1/TS 3.3.1//028AG2.1.32/

Given the following plant conditions:

- At 0800 A unit startup is in progress in accordance with GP-005, Power Operation (Mode 2 to Mode 1)

- At 0900 The OATC is placing Rod Control into Automatic PRZ level transmitter LT-459A fails high Which ONE of the following identifies (1) the actions required, IF any, in accordance with Technical Specification 3.3.1, RPS Instrumentation AND (2) the bases for this Functional Unit?

(Reference provided)

A. (1) Action per T.S. 3.3.1 is NOT required since 2 channels are still available (2) Protects downstream piping against water damage due to PRZ flooding.

B. (1) Action per T.S. 3.3.1 is NOT required since 2 channels are still available (2) Prevent water relief of liquid coolant through the PRZ safety valves.

C. (1) The inoperable channel must be placed in the tripped condition prior to 1500 (2) Protects downstream piping against water damage due to PRZ flooding.

D. (1) The inoperable channel must be placed in the tripped condition prior to 1500.

(2) Prevent water relief of liquid coolant through the PRZ safety valves.

Thursday, May 19, 2016 5:04:42 PM 221

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Technical Specification 3.3.1 Functional Unit 11, Pressurizer Water Level - High (Above P-7) requires action 6 to be completed for an inoperable channel. During a unit startup in accordance with GP-005, Power Operation (Mode 2 to Mode 1), step 121, at 15% turbine load Rod Control is transferred from Manual to Automatic. Therefore if the OATC is placing Rod Control in Automatic the Reactor must be > 10% power (P-10 initiates P-7). The High PRZ water level trip setpoint provides sufficient margin such that the undesirable condition of discharging liquid coolant through the safety valves is avoided.

A. Incorrect. The first part is plausible since this would be correct if P-7 was not present but since turbine power must be > 15% to place Rod Control in Automatic P-13 (Turbine > 10% power) has been exceeded. With either P-10 (Reactor power > 10%) or P-13 met P-7 which would be enabled which would re-instate the 92% PRZ High level Reactor trip. The second part is plausible since a PZR high level does result in the potential for damage to the piping downstream of the PZR PORV's and Safeties; however this is incorrect because this is the basis for the RPS P-14, High-High SG Level turbine trip.

B. Incorrect. The first part is correct. The second part is plausible see A(2).

C. Incorrect. The first part is plausible see A(1). The second part is correct.

D. Correct.

000028 Pressurizer Level Malfunction / 2 028AG2.1.32; Ability to explain and apply system limits and precautions.

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: RO 3.8 SRO 4.0 Technical

Reference:

Tech Spec 3.3.1 Table 3.3-1, Reactor Trip System Instrumentation Function Unit 11. Pressurizer Water Level--High (Above P-7)

FSAR 7.2.2.3.4, Reactor Trip System - Pressurizer Water Level, Page 7.2.2-13 (Page 6349)

References to be provided: T.S. 3.3.1 Table 3.3-1 Functional Units (7-11)

Page 3/4 3-2 (Page 124)

T.S. 3.3.1 Table 3.3-1 Page 3/4 3-7 (Page 129)

Learning Objective: Student Text RPS, Objective 7 Student Text PZRLC, Objective 11 Thursday, May 19, 2016 5:04:42 PM 222

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Question Origin: New Comments: HNP: Unable to write a SRO question to K/A 028AG2.4.34. Requested new K/A.

Phonecon 3/23: HNP couldnt come up with an AOP-004 scenario, so selected a new K/A, keeping APE 028 and randomly selecting from the 43 Generic items in ES-401 D.1.b:

New K/A 028AG2.1.32: Pressurizer Level Control Malfunction - Ability to explain and apply system limits and precautions.

Tier/Group: T1G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must be knowledgable of application of Technical Specification required actions and basis for the system limits. Requires knowledge of Technical Specification that is greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions, information below the LCO line and is not solely based on knowing the Safety Limits.

Thursday, May 19, 2016 5:04:42 PM 223

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 008/BANK/C/A//AOP-004/NONE//068AA2.06/

Given the following plant conditions:

- The unit has experienced a loss of Control Room habitability and control has been established at the ACP

- Normal operating No-Load temperature and pressure has been established Subsequently the following trends are noted:

- RCS temperature is 557°F and rising

- RCS pressure is 2255 psig and rising

- PRZ level is 26% and rising As CRS at the ACP which ONE of the following actions would you direct to control the RCS pressure rise?

Procedure Titles:

AOP-004, Remote Shutdown AOP-019, Malfunction of RCS Pressure Control A. OPEN one PRZ PORV to manually control pressure in accordance with AOP-004.

B. OPEN PRZ spray valves to manually restore pressure in accordance with AOP-019.

C. Dispatch an operator to open the breakers for C and D PRZ heater groups to control pressure in accordance with AOP-004.

D. Dispatch an operator to open the breakers for A and B PRZ heater groups and restore pressure in accordance with AOP-019.

Thursday, May 19, 2016 5:04:42 PM 224

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: While control of the PRZ pressure control functions are transferred to the ACP in accordance with AOP-004 only the A and B bank of heaters are automatically controlled from the ACP. The first action required in accordance with AOP-004 is to remove control power fuses and trip PRZ heaters C and D locally in the associated switchgear room. By removing the fuses for the heaters the rise in PRZ pressure will be terminated.

A. Incorrect. Plausible since RCS pressure is rising and the PRZ PORVs provide a relief path to lower pressure in the event that pressure rises to 2335 psig; however this is incorrect since the RCS pressure is 2255 psig the action to relieve pressure via the PRZ PORVs is not required .

B. Inorrect. Plausible since other procedures could be referenced while controlling the plant from the ACP. ACP controls of the PRZ heaters have been established and implementing procedure AOP-004 followup actions provide the necessary information to safely shut down the plant.

However, AOP-019 is the procedure normally entered to mitigate PRZ pressure control malfunctions and the action described is directed by AOP-019.

C. Correct.

D. Incorrect. Plausible same as answer 'B' reasoning.

Thursday, May 19, 2016 5:04:42 PM 225

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000068 (BW/A06) Control Room Evac. / 8 068AA2.06; Ability to determine and interpret the following as they apply to the Control Room Evacuation: RCS pressure (CFR: 43.5 / 45.13)

Importance Rating: RO 4.1 SRO 4.3 Technical

Reference:

AOP-004 Section 3.2 step 16.a, Rev 67, page 59 References to be provided: None Learning Objective: AOP-LP-3.4, Objective 4 Question Origin: Bank Comments: None Tier/Group: T1G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of diagnostic steps and decision points in the abnormal operating procedures (AOP) that actions to mitigate an event using specific procedure sections.

Thursday, May 19, 2016 5:04:42 PM 226

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 009/NEW/C/A//T.S. 3.4.8/NONE//076AG2.4.47/

Given the following plant conditions:

- The plant tripped form 100% power and is now stable Post trip Chemistry RCS Dose Equivalent I-131 sample results are as follows:

TIME ACTIVITY (Ci/gm) 0900 0.6 0915 0.9 0930 1.2 0945 1.4 1000 1.7 Which ONE of the following completes the statements below?

The FIRST time that Technical Specification 3.4.8, Reactor Coolant System: Specific Activity, action statement is required to be entered is at (1) .

The basis of Technical Specification LCO 3.4.8 action to reduce RCS Tavg below 500°F is to (2) .

A. (1) 0930 (2) ensure that the 1-hour dose at the SITE BOUNDARY will not exceed a small fraction of the 10 CFR Part 100 dose guideline limits in the event of a SGTR B. (1) 0930 (2) prevent a release of activity should a SGTR occur by preventing the SG atmospheric reliefs from automatically lifting C. (1) 1000 (2) ensure that the 1-hour dose at the SITE BOUNDARY will not exceed a small fraction of the 10 CFR Part 100 dose guideline limits in the event of a SGTR D. (1) 1000 (2) prevent a release of activity should a SGTR occur by preventing the SG atmospheric reliefs from automatically lifting Thursday, May 19, 2016 5:04:42 PM 227

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: TS 3.4.8 requires verification of DOSE EQUIVALENT I-131 specific activity less than or equal to 1.0 uCi/gm. As stated in the stem, at some time between 0915 and 0930, Dose Equivalent I-131 exceeded the 1.0 uCi/gm limit.

Therefore, 0930 is the first time at which LCO 3.4.8 is NOT met.

The Tech Spec basis for second part of the question states: Reducing Tavg to less than 500°F prevents the release of activity should a steam generator tube rupture occur, since the saturation pressure of the Reactor Coolant is below the lift pressure of the atmospheric steam relief valves.

A. Incorrect. The first part is correct. The second part is plausible since this is the Technical Specification basis for RCS specific activity but this is for 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose not a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> dose and does not take into account the 500° RCS temperature. TS 3.4.8 Specific Activity Bases: The limitations on the specific activity of the Reactor Coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with anassumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm.

B. Correct.

C. Incorrect. The first part is plausible since the plant has experienced a transient and per Tech Spec 3.4.7 "RCS Chemistry" transient limits for Chloride and Floride are < 1.50 (ppm). Therefore, it is reasonable for a candidate to have a misconception about these limits through confusion. This misconception would result in the first time that RCS Dose Equivalent I-131 being out of spec was at 1000.

The second part is plausible see A(2).

D. Incorrect. The first part is plausible see C(1). The second part is correct.

Thursday, May 19, 2016 5:04:42 PM 228

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 000076 High Reactor Coolant Activity / 9 076AG2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

(CFR: 41.10 / 43.5 / 45.12)

Importance Rating: RO 4.2 SRO 4.2 Technical

Reference:

Tech Spec 3.4.8 and Tech Spec Basis References to be provided: None Learning Objective: AOP-LP-3.32 Objectives 3 and 4 Question Origin: New Comments: HNP was unable to write a question to the SRO level for the original K/A 000076 High Reactor Coolant Activity 076AG2.4.2; Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. There are NO EOP entry conditions in HNP procedures associated with High Reactor Coolant activity and even if written to the AOP the Knowledge of the system per the K/A was not applicable.

Mike Donithan provided HNP a new K/A on 4-15-2016.

076AG2.4.47 Tier/Group: T1G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must be knowledgable of application of Technical Specification required actions and surveillance requirements.

Thursday, May 19, 2016 5:04:42 PM 229

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 010/BANK/C/A//EOP-ES-0.2/NONE//WE10EA2.1/

Given the following plant conditions:

- Natural circulation cooldown and depressurization is in progress in accordance with EOP-ES-0.2, Natural Circulation Cooldown The following conditions exist:

- An estimated leak rate from 'A' RCP #1 seal is 20 gpm and rising slowly

- RCS Pressure is 825 psig and lowering

- Thot is 495°F and lowering

- PRZ level is 25% and lowering slowly

- RVLIS Train A Upper Range level is 92% and lowering

- RVLIS Train B Upper Range level Input Quality Code and Error Status is D0(BAD)

The Plant Staff determines that cooldown and depressurization must CONTINUE.

Which ONE of the following action(s) is correct?

A. Transition to EOP-ES-0.4, Natural Circulation Cooldown With Steam Void in Vessel Without RVLIS.

B. Transition to EOP-ES-0.3, Natural Circulation Cooldown With Steam Void in Vessel With RVLIS.

C. Raise RCS subcooling to collapse voids and remain in EOP-ES-0.2, Natural Circulation Cooldown.

D. Actuate SI and go to EOP-E-0, Reactor Trip or Safety Injection.

Thursday, May 19, 2016 5:04:42 PM 230

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: During natural circulation cooldown and depressurization with RVLIS available, the criteria for transition to ES-0.3 is met when unable to maintain RVLIS >94% and RCS depressurization must continue. In this instance RVLIS level is 92% and lowering and Plant staff has communicated that depressurization must continue. Transition to ES-0.3 is correct.

A. Incorrect. Plausible because would be the required transition if the RVLIS were not available, however with RVLIS Train 'A' available, ES-0.4 is not the correct procedure.

B. Correct.

C. Incorrect. Plausible because maintaining subcooling is directed in the depressurization steps, but to increase subcooling, pressurizing the RCS would be required, and Plant staff has communicated that depressurization must continue.

D. Incorrect: Plausible because PRZ level lowering, but SI Actuation Criteria is not currently met. Conditions (20 gpm leakage) currently do not warrant Safety Injection, and continuing to depressurize the RCS will lessen this leakage.

Thursday, May 19, 2016 5:04:42 PM 231

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal W/E09&E10 Natural Circ. / 4 WE10EA2.1 Ability to determine and interpret the following as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

(CFR: 43.5 / 45.13)

Importance Rating: RO 3.2 SRO 3.9 Technical

Reference:

EOP-ES-0.2, Note prior to Step 18, Rev. 0, Page 32, 33 References to be provided: None Learning Objective: EOP-LP-3.8, Objective 4.c Question Origin: Bank Comments: None Tier/Group: T1G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.

Thursday, May 19, 2016 5:04:42 PM 232

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 011/NEW/C/A//EOP-E-1/EOP-E-1 ATT 2//006G2.2.44/

Given the following plant conditions:

- A large break LOCA occurs with a loss of Off-site power

- EOP-E-1, Loss of Reactor Or Secondary Coolant, is in progress Subsequently the following occurs:

- A fire is reported from MCC 1B31-SB

- The crew is evaluating if cold leg recirculation capability exists Which ONE of the following completes the statement below?

Based on the conditions above AND the indications in the reference provided, the 1A-SA Safety Bus has (1) AND the CRS will transition to (2) .

(Reference provided)

A. (1) energized (2) EOP-ES-1.3, Transfer To Cold Leg Recirculation B. (1) energized (2) EOP-ECA-1.1, Loss Of Emergency Coolant Recirculation C. (1) failed to energize (2) EOP-ES-1.3, Transfer To Cold Leg Recirculation D. (1) failed to energize (2) EOP-ECA-1.1, Loss Of Emergency Coolant Recirculation Thursday, May 19, 2016 5:04:42 PM 233

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The main control board indicates the A' RHR pump has control power available by the green light for the pump, but the system valves which are powered by 480V MCC's do not have control power indicating that the 'A' EDG has not energized the 1A-SA Safety bus. Additionally MLB 2A-SA indicates the 'A' CSIP is not running confirming the 1A-SA Safety bus is not energized. Attachment 2 of EOP-E-1 provides the components and the condition of those components required to establish cold leg recirculation. The 'A' train has multiple components that do not meet the required conditions due to the loss of the 1A-SA Safety bus. After reviewing the Monitor light boxes, MLB 3B-SB indicates that 1SI-341, Low Head SI train B to cold leg valve is in the shut position by the status light being illuminated. This valve is normally open with control power removed, but the potential of a hot short due to the fire in 1B31-SB can reposition this valve. 1SI-323 which is powered from 1B31 does not have power as indicated by the lack of red or green valve indication on its control switch.

With 1SI-341 in the shut position and no power available to 1SI-323 the 'B' train is not capable of being placed in cold leg recirculation. With both trains of RHR not capable of cold leg recirculation the procedure directs the candidate to go to EOP-ECA-1.1, Loss Of Emergency Coolant Recirculation".

A. Incorrect. The first part is plausible since the 'A' RHR Pump and 1RH-1 have indication on the MCB; however this is incorrect because the MCB indication for the 'A' RHR Pump is an indication of the status of the breakers DC control power and 1RH-1 does have power but it is powered from the 1B-SB Safety bus. The second part is plausible since the 'B' RHR pump is running and has indication of flow; however this is incorrect because 1SI-341 is not in the correct position. Additionally the RWST level is 38% which does not meet the transition criteria of RWST level less than 23.4%

B. Incorrect. The first part is plausible see A(1). The second part is correct.

C. Incorrect. The first part is correct. The second part is plausible see A(2).

D. Correct.

006 Emergency Core Cooling System 006G2.2.44; Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.10 / 43.2 / 45.6)

Importance Rating: RO 4.2 SRO 4.4 Technical

Reference:

EOP-E-1, Step 12, Rev 1, Page 16 EOP-E-1, Attachment 2, Rev 1, Page 28 Thursday, May 19, 2016 5:04:42 PM 234

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal References to be provided: EOP-E-1, Attachment 2, Rev 1, Page 28 MCB Indication images Learning Objective: EOP-LP-3.1, Objective 2.a Question Origin: New Comments: Discuss with Mike about how to make SRO based on SRO guidance for AOP/EOP entry conditions not being SRO knowledge. Original K/A 007G2.4.4; Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Phonecon 3/23: The procedures mentioned above wont work for this K/A, so generated:

New K/A 006G2.2.44: Emergency Core Cooling System

- Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. SRO IR 4.4 Tier/Group: T2G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific subprocedures or emergency contingency procedures.

Thursday, May 19, 2016 5:04:42 PM 235

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 012/NEW/C/A//TS 3.4.9.4/TS 3.4.9.4/EARLY/005A2.02/ENHANCE Given the following plant conditions:

- A plant heat up is in progress on July 19th

- RCS temperature is 175°F and rising slowly

- The RCS is in a solid plant condition with both RHR Trains in service Subsequently the following occurs:

- At 0830, 1RH-30, RHR HX Outlet Isolation Valve fails closed

- ALB-010-5-1, RC Overpress, alarms

- The first PORV to operate, LTOPS PORV 445A, cycles open at 480 psig and shuts

- LTOPS PORV 445B remains shut during this event Which ONE of the following identifies (1) the operability status of the LTOPS PORV's AND (2) the required Technical Specification action(s), IF any, for the LTOPS?

(Reference Provided)

A. (1) ONE inoperable PORV (2) Plant heat up to draw a bubble may continue.

B. (1) ONE inoperable PORV (2) Prepare and submit a special report to the Commision by August 20th.

C. (1) TWO inoperable PORVs (2) Restore the inoperable PORV to operable by 0830 on July 20th.

D. (1) TWO inoperable PORVs (2) Depressurize and vent the RCS via a 2.9 square inch vent by 1630 on July 19th.

Thursday, May 19, 2016 5:04:42 PM 236

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: PLP-106 provides a set value of +/- 65 psig as the tolerance for the LTOPS PORV actuation setpoint. RCS pressure actuation setpoint is 465 psig and 475 psig respectively for the Low and High PORV. Based on LTOPS PORV 445A actuating at 480 psig the valve is inoperable. Because this PORV is the first to operate the candidate should determine that LTOPS PORV 445B did not operate as designed also and both LTOPS PORVs are inoperable. This condition requires the implentation of TS 3.4.9.4 action statement c. With both PORVs inoperable, depressurize and vent the RCS through at least a 2.9 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

A. Incorrect. The first part is plausible in the event the candidate misinterpets the Figure 3-4.4 curve and determines the PORV operated within the PLP-106 maximum tolerance of 80 psig for the Reactor Vessel P-T limits which would be 480 psig; however this is incorrect since the PORV actuation occurs at 480 this is higher than the allowed tolerance of 465 psig. The second part is plausible since the candidate determines the PORV remains operable the evolution the was in progress may continue; however this is incorrect since the PORV actuation occurs outside of the calculated tolerance therefore the PORV is inoperable and will not allow the plant heat to continue.

B. Incorrect. The first part is plausible see A1. The second part is plausible in the event the candidate misapplies the action d due to the operation of the PORV to reduce RCS pressure; however this is incorrect since the assumption is the PORV remains operable therefore the requirements of action d. do not apply.

C. Incorrect. The first part is correct. The second part is plausible since this action is correct if only one PORV is determine to inoperable, the candidate misinterpets the Figure 3-4.4 curve and determines the PORV operated within the PLP-106 maximum tolerance of 80 psig for the Reactor Vessel P-T limits which would be 490 psig for the high PORV; however this is incorrect since actuation pressure of the first PORV to operate is above the maximum actuation setpoint pressure for both the low and high pressure LTOPS PORV both PORVs are inoperable.

D. Correct.

Thursday, May 19, 2016 5:04:42 PM 237

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 005 Residual Heat Removal 005A2.02; Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Pressure transient protection during cold shutdown (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Importance Rating: RO 3.5 SRO 3.7 Technical

Reference:

Tech Spec 3.4.9.4 Action item c, Tech Spec PORV Setpoint Figure 3.4-4 PLP-106, Attachment 10, Rev 58, Page 75 References to be provided: Tech Spec 3.4.9.4, Tech Spec PORV Setpoint Figure 3.4-4 PLP-106, Attachment 10 Learning Objective: Student Text PRZPC, Objective 11.e Question Origin: New Comments: None Tier/Group: T2G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must be knowledgable of application of Technical Specification required actions and surveillance requirements. Requires knowledge of Technical Specification that is greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions, information below the LCO line and is not solely based on knowing the Safety Limits.

Thursday, May 19, 2016 5:04:42 PM 238

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 013/NEW/C/A//T.S. 3.7.3/PHOTOS AND TS 3.7.3//008G2.2.44/

Given the following plant conditions:

- The unit is operating at 100% power

- 'A' Train equipment is in service

- CCW indications on the MCB are as follows:

(See Reference Photo 1)

Subsequently on July 18, 2016 at 1100, multiple CCW low flow and both 'A' and 'B' low pressure annunciators alarmed on ALB-005.

- BOTH 'A' and 'B' CCW pumps have red running lights illuminated

- CCW indications are now as follows:

(See Reference Photo 2)

Which ONE of the following completes the statement below?

Based on the conditions above AND the indications provided in the reference, the 'A' CCW pump has a (1) . If the CCW system is not restored, in accordance with Technical Specification 3.7.3, Plant Systems: Component Cooling Water System the unit must in at least HOT STANDBY no later than (2) .

A. (1) shaft shear (2) 1100 on July 21, 2016 B. (1) shaft shear (2) 1700 on July 21, 2016 C. (1) leak upstream of flow transmitter FI-652.1 CCW HTX A Outlet Flow (2) 1100 on July 21, 2016 D. (1) leak upstream of flow transmitter FI-652.1 CCW HTX A Outlet Flow (2) 1700 on July 21, 2016 Thursday, May 19, 2016 5:04:42 PM 239

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: For the first part: Indications for the 'A' CCW pump shaft shear are a lack of outlet flow indication on FI-652.1 (reads 0 GPM), the red running light is still illuminated indicating that the pump is not tripped or just the green light would be on. Low flow and pressure annunciators alarming indicate that the 'A' CCW pump has at one time not provided adequate discharge flow to maintain conditions above the alarm setpoints. At 52 psig and decreasing the 'B' CCW pump which was in standby would have automatically started. The indication that this occurred was that the red running light is illuminated on the 'B' CCW pump and the flow indications on 'B' pump (FI-653.1) initially were at zero and are now at approximately 10,500 GPM.

There are no changes in other pressures, temperatures or flows that would indicate that a leak in the CCW system has occurred.

The second part: Tech Spec 3.7.3 LCO for only ONE CCW flow path being OPERABLE now applies. The CCW flow path needs to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant needs to be placed in HSB within the next 6 and CSD within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. July 18 at 1100 + 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> + 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to HSB would be July 21 at 1700.

A. Incorrect. The first part is correct. The second part is plausible if ONLY 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is used to place the plant in HSB.

B. Correct.

C. Incorrect. The first part is plausible since a leak could cause low pressure in the CCW system which could also reduce system pressure to below the autostart pressure (52 psig) of the standby pump. Multiple low flow and pressure alarms could also occur with a CCW leak. This leak can be elimated by the absense of reduced pressures and flows from reference photo 1 compared to reference photo 2. The second part is plausible (see A.2)

D. Incorrect. The first is plausible (see C.1). The second part is correct.

Thursday, May 19, 2016 5:04:42 PM 240

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 008 Component Cooling Water 008G2.2.44; Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR: 41.5 / 43.5 / 45.12)

Importance Rating: RO 4.2 SRO 4.4 Technical

Reference:

Photos indicating CCW before/after failure.

Tech Spec 3.7.3 References to be provided: None Learning Objective: Student Text CCW, Objective 7.a (Part 1)

Student Text CCW, Objective 12 (Part 2)

Question Origin: New Comments: None Tier/Group: T2G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must be knowledgable of application of Technical Specification required actions and surveillance requirements. Requires knowledge of Technical Specification that is greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions, information below the LCO line and is not solely based on knowing the Safety Limits.

Thursday, May 19, 2016 5:04:42 PM 241

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 014/NEW/C/A//TS 3.4.3, ALB-009/NONE//010A2.01/

Given the following plant conditions:

- The unit is at 20% power the crew is conducting a normal startup

- Backup Heater Groups A, B, and D are ON Subsequently the following conditions exist:

- ALB-009-7-5, Pressurizer Heater Overload Trip, is in alarm

- PRZ Heater Group A is de-energized due to overcurrent

- PRZ Pressure indicates 2215 psig and lowering slowly

- PRZ Level is 27% and rising Which ONE of the following completes the statements below?

Based on the indications above, the PRZ Heater Group A breaker must be racked out because (1) .

Heater Group A is (2) to satisfy the surveillance requirements of Technical Specification 3.4.3, Reactor Coolant System: Pressurizer.

A. (1) there is no mechanical lockout to prevent reclosure (2) required B. (1) there is no mechanical lockout to prevent reclosure (2) NOT required C. (1) subsequent closure of the breaker may render the Diesel Generator inoperable (2) required D. (1) subsequent closure of the breaker may render the Diesel Generator inoperable (2) NOT required Thursday, May 19, 2016 5:04:42 PM 242

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The PRZ Heater group breakers do not have anti pump logic to mechanically prevent the breaker from reclosing in the event that it trips open due to overcurrent. APP-ALB-009 directs the breaker to be racked out until that cause of the overcurrent conditions is investigated. Technical Specification 3.4.3 action a requires at least two groups of pressurizer heaters be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with only one group of heaters operable.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the LCO only requires 2 heater groups with a capacity of 125 KW and the capacity heater groups A, B C and D all meet this minimum 125 KW requirement.

However this is incorrect because in accordance with OWP-RC-08 and MST-E0023 group A and Group B heaters are required to exhibit output of 125 KW or greater to successfully comply with the T.S. 3.4.3 surveillance requirement.

C. Incorrect. The first part is plausible since components that are powered from the diesel backed safety bus may not properly load shed the EDG may be inoperable. This caution and limitation is found in multiple operations procedures as a reminder to evaluate the EDG operability for potential impacts to sequencing and load shedding requirements. However this is incorrect because the PRZ Heater Group A is powered from the emergency bus which is stripped from the safety bus during sequencer operations and only the 1A1 supply breaker is required to be stripped for EDG operability. The second part is correct.

D. Incorrect. The first is plausible see C(1). The second part is plausible see B(2)

Thursday, May 19, 2016 5:04:42 PM 243

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 010 Pressurizer Pressure Control 010A2.01; Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Heater failures (CFR: 41.5 / 43.5 / 45.3 / 45.13)

Importance Rating: RO 3.3 SRO 3.6 Technical

Reference:

APP-ALB-009, Window 7-5, Rev 17, Page 28 Technical Specification 3.4.3, Page 3/4 4-10 OWP-RC-08, Rev 6, Page 13 MST-E0023, Section 6.0 Step 1, Rev 12, Page 5 References to be provided: None Learning Objective: Student Text PRZPC, Objective 11.c Question Origin: New Comments: None Tier/Group: T2G1 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must be knowledgable of application of Technical Specification required actions and surveillance requirements. Requires knowledge of Technical Specification that is greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> actions, information below the LCO line and is not solely based on knowing the Safety Limits.

Thursday, May 19, 2016 5:04:42 PM 244

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 015/BANK/C/A//EOP-ES-1.3/NONE//026G2.4.9/

Given the following plant conditions:

- A LOCA occurred 45 minutes ago

- The crew is performing actions in accordance with EOP-ES-1.3, Transfer to Cold Leg Recirculation

- The OATC is in the process of performing the valve alignment During the valve alignment the following alarms are received:

- ALB-001-2-2, SPRAY PUMP A DISCHARGE LOW PRESS

- ALB-001-2-5, SPRAY PUMP A SUCTION LOW PRESS Both alarms are received and clear intermittently over the course of about 1 minute

- "A" RHR pump amps and discharge pressure are beginning to oscillate

- The CRS has determined that Train 'A' recirculation sump performance is degraded Which ONE of the following identifies (1) the procedure implementation strategy AND (2) the mitigating actions based on the determination that the recirculation sump is degraded?

A. (1) Remain in EOP-ES-1.3 (2) Stop 'A' Containment Spray Pump B. (1) Remain in EOP-ES-1.3 (2) Throttle CSIP flow to be slightly greater than the minimum flow requirements C. (1) Go to EOP-ECA-1.1, Loss of Emergency Coolant Recirculation (2) Stop 'A' Containment Spray Pump D. (1) Go to EOP-ECA-1.1, Loss of Emergency Coolant Recirculation (2) Throttle CSIP flow to be slightly greater than the minimum flow requirements Thursday, May 19, 2016 5:04:42 PM 245

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Reason the first part of question is correct: Once EOP ES-1.3 is entered a caution prior to step 1 states that steps 1 through 9 should be performed without delay. Even though there is indications of a degraded sump on the

'A' train of equipment, the contingency EOP-ECA-1.1 should not be transistioned to at this time because cold leg recirculation capibility is still available using the 'B' train equipment. EOP-ES-1.3 should be carried out until reaching step 10 where initation of monitoring of CSF Status Trees are started. The second part of question is correct in accordance with the APP for ALB-001-2-2 when the discharge pressure for the spray pump is low due to no suction source (degraded sump prevents suction to the pump) the pump should be secured.

A. Correct.

B. Incorrect. The first part is correct. In the second part of the question the stem information is that the RHR pumps are experiencing cavitaion. It is plausible that increasing the CSIP discharge which raises RCS level will give the RHR pumps a greater NPSH. The increased NPSH should stop the RHR pump cavitation problem making this answer plausible.

C. Incorrect. The first part is plausible since there are indications that the suction sources to the Containment Spray and RHR pumps are inadequate but during the alignment for Cold Leg Recirculation the procedure is implemented without transitioning to another procedure until after the alignment is completed (step 10). The second part is correct.

D. Incorrect. The first part is plausible (see C.1). The second part is plausible (see B.2)

Thursday, May 19, 2016 5:04:42 PM 246

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 026 Containment Spray 026G2.4.9; Knowledge of low power/shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: RO 3.8 SRO 4.2 Technical

Reference:

EOP-ES-1.3 Caution prior to step 1, Rev. 2, Page 4 APP ALB-001-2-2, Rev. 22, Page 6 References to be provided: None Learning Objective: EOP-LP-2.3, Objective 5.a, Student Text CSS, Objective 6.d Question Origin: Bank Comments: None Tier/Group: T2G1 SRO Justification: 10 CFR Part 55 Content - this question meets the SRO level of knowledge by assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations

[10 CFR 55.43(b)(5)], involving BOTH:

1) assessing plant conditions and then

2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Thursday, May 19, 2016 5:04:42 PM 247

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 016/NEW/C/A//CURVE F-20-1,3.1.3.6/PLP-106 AND TS//001G2.1.25/

Given the following plant conditions:

- A load reduction was initiated in accordance with GP-006, Normal Plant Shutdown From Power Operation To Hot Standby (Mode 1 To Mode 3)

The following indications are observed:

Time Power Control Bank C Control Bank D 0600 75% 228 steps 155 steps 0630 70% 228 steps 125 steps 0700 65% 228 steps 110 steps 0730 60% 223 steps 95 steps 0800 55% 213 steps 85 steps Which ONE of the following identifies (1) the EARLIEST time that the action statement is required to be entered for Technical Specification 3.1.3.6, Control Rod Insertion Limits AND (2) the action(s) required to safisfy the LCO at that time?

(Reference Provided)

A. (1) 0630 (2) Restore control banks to within the insertion limit specified by 0830.

B. (1) 0630 (2) Reduce Thermal Power to less than 65% by no later than 1030.

C. (1) 0730 (2) Restore control banks to within the insertion limit specified by 0930.

D. (1) 0730 (2) Reduce Thermal Power to less than 45% by no later than 1130.

Thursday, May 19, 2016 5:04:42 PM 248

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: The Rod Insertion technical specification 3.1.3.6 limit is a linear curve that increases the limit 1.86 steps for each percent power. With the reactor at 70% power the rod insertion limits for control bank C and D are 225 and 130 steps repectively. The control rods indicate they are at 228 on control bank C and 125 steps on control bank D therefore control bank D is clearly below the technical specification 3.1.3.6 limits at time 0630 as indicated. At that time the candidate must apply action statement a or b within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to either restore rods to above the insertion limits for action a or reduce thermal power below the required fraction of rated thermal power for the rod height at that time for action b.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the normal progression of LCO action statements is to perform the first action, i.e.

action a, then if not completed perform the second action, i.e. action b, within the following time frame after the elapse of the first action; however this is incorrect since the LCO allows the candidate to perform either action statement to restore compliance with the LCO within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> timeframe therefore the cumulative time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is improperly applying the technical specification 3.1.3.6 LCO.

C. Incorrect. The first part is plauible since it is correct at that current time; however this is incorrect since it is not the earliest time. The second part is plausible since it is the correct action based on the current time; however this is incorrect since it is not the earliest time.

D. Incorrect. The first part is plauible since it is correct at that current time; however this is incorrect since it is not the earliest time. The second part is plausible since the normal progression of LCO action statements is to perform the first action, i.e. action a, then if not completed perform the second action, i.e. action b, within the following time frame after the elapse of the first action; however this is incorrect since the LCO allows the candidate to perform either action statement to restore compliance with the LCO within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> timeframe therefore the cumulative time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is improperly applying the technical specification 3.1.3.6 LCO.

Thursday, May 19, 2016 5:04:42 PM 249

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 001 Control Rod Drive 001G2.1.25; Ability to interpret reference materials, such as graphs, curves, tables, etc.

(CFR: 41.10 / 43.5 / 45.12)

Importance Rating: RO 3.9 SRO 4.2 Technical

Reference:

Curve No F-20-1, Rev 0 Technical Specification 3.1.3.6, Page 3/4 1-21 References to be provided: PLP 106 Attachment 9 Sheet 10 of 14, Rev 58 Technical Specification 3.1.3.6, Page 3/4 1-21 Learning Objective: Student Text RODCS, Objective 15.d Question Origin: New Comments: None Tier/Group: T2G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the Technical Specifications and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know the Technical Specification actions of the limiting conditions of operation concerning control bank insertion limits. Requires knowledge of ability to apply Technical Specification action statements greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that are not system knowledge.

Thursday, May 19, 2016 5:04:42 PM 250

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 017/BANK/C/A//TS 3.9.2/NONE/EARLY/034A4.02/

Given the following plant conditions:

- The unit is in Mode 6 with defueling in progress

- NI-31 is selected for audible count rate At 0935, power is lost to NI-32 due to failure of the instrument power fuse Which ONE of the following statements describes (1) the requirements as a result of this failure in accordance with Technical Specification 3.9.2, Refueling Operations -

Instrumentation AND (2) the basis for the requirements?

A. (1) Verify the Wide Range Neutron Flux Monitor on the opposite side of the core from NI-31 is operable and refueling operations may continue.

(2) Ensures that redundant NEUTRON monitoring capability is available.

B. (1) Verify the Wide Range Neutron Flux Monitor on the opposite side of the core from NI-31 is operable and refueling operations may continue.

(2) Ensures that redundant AUDIBLE monitoring capability is available.

C. (1) Immediately suspend refueling operations.

(2) Minimizes reactivity changes during a REDUCED neutron flux monitoring capability event.

D. (1) Immediately suspend refueling operations.

(2) Minimizes reactivity changes due to the DELAYED neutron flux monitoring response time from N-31.

Thursday, May 19, 2016 5:04:42 PM 251

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: Per the asterisk in the LCO statement and it's associated note: WIth N-31 unaffected and selected for audible count rate, substitution of an operating Wide Range Neutron Flux Monitor is allowed, the LCO is met, and no action is required. Therefore refueling operations may continue.

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the neutron monitoring system is does provide audible indication of neutron flux level in both containment and the MCR; however this is incorrect since the Wide Range Flux monitors are indication only and not selectable for audible indication.

C. Incorrect. The first part is plausible since these are the required actions in the event that both required monitors are inoperable or not working; however this is not correct since NI-31 remains available the only action required is to suspend core alterations until Wide Range Flux monitor NI-61 is verified operable. The second part is plausible since it is the correct bases for the expected action statement; however this is not correct since NI-31 remains available the only action required is to suspend core alterations until Wide Range Flux monitor NI-61 is verified operable.

D. Incorrect. The first part is plausible since these are the required actions in the event that both required monitors are inoperable or not working; however this is not correct since NI-31 remains available the only action required is to suspend core alterations until Wide Range Flux monitor NI-61 is verified operable. The second part is plausible since it is the correct bases for the expected action statement; however this is not correct since NI-31 remains available the only action required is to suspend core alterations until Wide Range Flux monitor NI-61 is verified operable.

Thursday, May 19, 2016 5:04:42 PM 252

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 034 Fuel Handling Equipment System (FHES) 034A4.02; Ability to manually operate and/or monitor in the control room: Neutron levels (CFR: 41.7 / 45.5 to 45.8)

Importance Rating: RO 3.5 SRO 3.9 Technical

Reference:

Tech Spec 3.9.2 Note on bottom of page and Tech Spec Basis 3.9.2 References to be provided: None Learning Objective: Student Text NIS, Objective 2.d and 13 Question Origin: New Comments: None Tier/Group: T2G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the TS and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of the ability to apply Technical Specification action statements greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that are not system knowledge and the bases for these actions.

Thursday, May 19, 2016 5:04:42 PM 253

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 018/MODIFIED/FUNDAMENTAL//T.S. 3.11.2.5/NONE//071G2.2.42/

Which ONE of the following completes the statements below?

In accordance with Technical Specification 3.11.2.5, Radioactive Effluents/Explosive Gas Mixture, the oxygen limit downstream of the Hydrogen Recombiners in the Gaseous Radwaste Treatment System is required to be (1) when the hydrogen concentration exceeds 4% by volume.

The bases for this restriction is to (2) .

Noun Name:

10 CFR Part 50, Domestic Licensing Of Production And Utilization Facilities A. (1) 2%

(2) prevent an explosive mixture that has the likelihood of damaging equipment needed for safe shutdown capability B. (1) 2%

(2) provide assurance that the release of radioactive materials due to an explosion will be controlled within 10 CFR Part 50 requirements C. (1) 4%

(2) prevent an explosive mixture that has the likelihood of damaging equipment needed for safe shutdown capability D. (1) 4%

(2) provide assurance that the release of radioactive materials due to an explosion will be controlled within 10 CFR Part 50 requirements Plausibility and Answer Analysis Reason answer is correct: In accordance with Technical specification 3.11.2.5, the concentration of oxygen in the GASEOUS RADWASTE TREATMENT SYSTEM downstream of the hydrogen recombiners shall be limited to less than or equal to 2%

by volume whenever the hydrogen concentration exceeds 4% by volume. The bases for Technical Specification 3.11.2.5 states "Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50".

A. Incorrect. The first part is correct. The second part is plausible since EP-EAL, Emergency Action Level Techincal Bases, Hazards category states an fire or explosion can pose significant hazards to personnel and reactor safety.

Appropriate for classification are fires within the site Protected Area or Thursday, May 19, 2016 5:04:42 PM 254

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal which may affect operability of equipment needed for safe shutdown. The Waste Process building is a structure that contains Safe Shutdown Equipment. However this is incorrect because the Technical Specification basis is to confrom with GDC 60 of Appendix A to 10 CFR part 50.

B. Correct.

C. Incorrect. The first part is plausible since OP-102.07 precaution and limitation

1. 17 states any mixture of hydrogen, oxygen and nitrogen containing less than 4% by volume hydrogen, is nonflammable; however this is incorrect since the Technical Specification is more restrictive and the limt is 2%

oxygen when the hydrogen concentration is greater than 4%.

D. Incorrect. The first part is plausible see C(1). The second part is plausible see A(2).

Original question:

071 Waste Gas Disposal 071G2.2.42; Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

(CFR: 41.10 / 43.5 / 45.13)

Thursday, May 19, 2016 5:04:42 PM 255

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Technical

Reference:

Technical Specification 3.11.2.5, Page 3/4 11-15, (page 357)

Technical specification Bases 3.11.2.5, Page B 3/4 11-1, (page 100)

References to be provided: None Learning Objective: Student Text GWPS, Objective 8.a Question Origin: Modified - 2013 NRC RO 61 Comments: Discuss with Mike...we have entry into AOP-012 for loss of a CW pump but there are no immediate actions with the AOP not an EOP. We are NOT going to be able to write a question to the SRO level with this K/A.

K/A 075G2.4.1; Knowledge of EOP entry conditions and immediate action steps should be replaced.

Phonecon 3/23: During the call I provided new K/A 011G2.4.1, changing out System 075 for System 011, Pressurizer Level Control, but the real problem with the initial K/A (that I missed the importance of in the 3/22 phone call) is that G2.4.1, Knowledge of EOP entry conditions and immediate action steps, does not lend itself to an SRO question because the SRO-only screening criteria in Figure 2 would kick a question out on both entry conditions and immediate actions.

Given that the G2.4.1 piece wont work for an SRO question, and that Circ Water isnt important to safety, randomly generated a completely new T2G2 K/A:

New K/A 071G2.2.42: Waste Gas Disposal - Ability to recognize system parameters that are entry-level conditions for Tech Specs.

(WGD was moved out of HNP TS in 1995, but its in the ODCM, so its fair game for SROs per the SRO-only Tech Spec flowchart. Or could possibly go after TS 3.11.2.5, Explosive Gas Mixture.)

Tier/Group: T2G2 SRO Justification: 10 CFR Part 55 Content - 43(b)(2): Facility operating limitations in the TS and their bases. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of the ability to apply Technical Specification action statements greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that Thursday, May 19, 2016 5:04:42 PM 256

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal are not system knowledge and the bases for these actions.

Thursday, May 19, 2016 5:04:42 PM 257

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 019/NEW/FUNDAMENTAL//OMM-002/NONE/EARLY/G2.1.3/

The Main Control Room has been notified that the OATC has been selected for Fitness for Duty screening and must leave the Control Room for approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, another operator will relieve the OATC.

In accordance with OMM-002, Shift Turnover Package, which ONE of the following identifies the MINIMUM position(s) responsible for approval of this unscheduled shift relief?

A. Shift Manager OR Shift Technical Advisor B. Control Room Supervisor OR Shift Manager C. Control Room Supervisor OR Shift Technical Advisor D. Control Room Supervisor AND Shift Technical Advisor Thursday, May 19, 2016 5:04:42 PM 258

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: For reliefs occurring during the shift, use the following guidance: Unscheduled relief during the shift requires the approval of the SM or CRS.

A. Incorrect. Plausible since the Shift Manager assumes the command function and is responsible for maintaining the MCR environment in a highly professional manner and the on shift Shift Technical Advisor is responsible for verifying the on coming personnel are qualified to to assume the shift position they will fill the candidate may assume either of the two positions is allowed to approve the unscheduled relief; however this is incorrect because the procedure requires either SM or CRS to approve unscheduled shift reliefs.

B. Correct.

C. Incorrect. Plausible since the Control Room Supervisor assumes the control function and ensures that adequate control room staffing is maintained at all times and the on shift Shift Technical Advisor is responsible for verifying the on coming personnel are qualified to to assume the shift position they will fill the candidate may assume either of the two positions is allowed to approve the unscheduled relief; however this is incorrect because the procedure requires either SM or CRS to approve unscheduled shift reliefs.

D. Incorrect. Plausible since the Control Room Supervisor assumes the control function and ensures that adequate control room staffing is maintained at all times and the on shift Shift Technical Advisor is responsible for verifying the on coming personnel are qualified to to assume the shift position they will fill the candidate may assume approval of both is required; however this is incorrect because the procedure requires either SM or CRS to approve unscheduled shift reliefs.

Thursday, May 19, 2016 5:04:42 PM 259

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.1 Conduct of Operations G2.1.3; Knowledge of shift or short-term relief turnover practices.

(CFR: 41.10 / 45.13)

Importance Rating: RO 3.7 SRO 3.9 Technical

Reference:

OMM-002, Section 5.1 Step 24.a.1, Rev 64, Page 9 References to be provided: None Learning Objective: Lesson Plan PP-LP-3.1 Objective 3 Question Origin: New Comments: None Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(1): Condition and limitations in the facility license. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know procedural knowledge of the required actions for not meetng administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).

Thursday, May 19, 2016 5:04:42 PM 260

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 020/BANK/FUNDAMENTAL//EOP-E-3/NONE//G2.1.9/

The crew is implementing EOP-E-3, Steam Generator Tube Rupture. The CRS is at the step to isolate flow from the ruptured SG.

Which ONE of the following completes the statements below?

The CRS should direct the OATC to set the ruptured SG PORV controller setpoint to (1)

The bases for setting the controller to the new setpoint is to (2) .

A. (1) 1135 psig (87%)

(2) prevent lifting the SG code safety valves B. (1) 1135 psig (87%)

(2) minimize RCS to ruptured SG P C. (1) 1145 psig (88%)

(2) prevent lifting the SG code safety valves D. (1) 1145 psig (88%)

(2) minimize RCS to ruptured SG P Thursday, May 19, 2016 5:04:42 PM 261

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: In accordance with the Westinghouse Owners Group (WOG)

Background document for step descriptions of the Steam Generator Tube Rupture procedure (E-3) the setpoint for the ruptured SG PORV controller should be adjusted so the setpoint is greater than no load value (85% at HNP this would be 1106 psig) in order to minimize atmospheric releases from the rupture steam generator and less than the minimum safety valve setpoint to prevent lifting of the code safety valves, which at HNP there are 5 safety valves with lift settings of 1170, 1185, 1200, 1215, and 1230 psig. The 25 psig margin is a typical value to allow for opening of the PORV prior to lifting of the safety valve.

A. Incorrect. The first part is plausible since this is the SG PORV controller setpoint that the CRS would direct the RO set the PORV during plant startup operations (GP-005 Section 5.0 step 5.e). This higher setting is to accommodate plant startup by placing an artifical load on the Reactor without causing the PORVs to open. The second part of the answer is correct.

B. Incorrect The first part is plausible since this is the SG PORV controller setpoint that the CRS would direct the RO set the PORV during plant startup operations (GP-005, Power Operation, Section 5.0 step 5.e). This higher setting is to accommodate plant startup by placing an artifical load on the Reactor without causing the PORVs to open. The second part is plausible since changing the controller setpoint to a higher value will increase the SG pressure and minimize the delta P between the RCS to SG.

C. Correct.

D. Incorrect. The first part is correct. The second part is plausible since changing the controller setpoint to a higher value will increase the SG pressure and minimize the delta P between the RCS to SG.

Thursday, May 19, 2016 5:04:42 PM 262

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.1 Conduct of Operations G2.1.9; Ability to direct personnel activities inside the control room.

(CFR: 41.10 / 45.5 / 45.12 / 45.13)

Importance Rating: RO 2.9 SRO 4.5 Technical

Reference:

EOP-E-3, Step 5, Page 8, Rev. 1, WOG Background Doc, E-3, Rev. 2, Page 61 References to be provided: None Learning Objective: EOP-LP-3.2 Objective 4.b Question Origin: Bank Comments: None Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item.

Thursday, May 19, 2016 5:04:42 PM 263

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 021/NEW/FUNDAMENTAL//AD-OP-ALL-0203/NONE//G2.2.1/

Given the following plant conditions:

- A Reactor startup is in progress in accordance with GP-004, Reactor Startup (Mode 3 To Mode 2)

- Reactor power will be held below 3% until EST-923, Initial Criticality And Low Power Physics Testing is completed Which ONE of the following completes the statement below?

In accordance with AD-OP-ALL-0203, Reactivity Management this is an (1) category planned reactivity evolution AND a DEDICATED SRO (Reactivity Manager)

(2) expected to provide oversight during the implementation this evolution.

A. (1) R1 (2) is B. (1) R1 (2) is NOT C. (1) R2 (2) is D. (1) R2 (2) is NOT Thursday, May 19, 2016 5:04:42 PM 264

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: R1 Evolutions are defiened as an activity significantly affecting core power/reactivity which requires significant operator attention. Examples include but are not limited to: (1) Zero Power Physics Testing, (2) Reactor startups The Reactivity Manager oversight (Dedicated SRO, other than the CRS or STA, with no concurrent duties) is expected to be stationed during R1 evolutions A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the CRS is allowed to perfrom the role of Reactivity Manager during R2 activities and power is being held below 5% may misapply the expectations of an R2 evolution since power changes will be less than 10%. However this is incorrect because Low Power Physics testing and Reactor startups are considered R1 evolutions and the expectation is that the Reactivity Manager be a dedicated SRO, other than the CRS or STA.

C. Incorrect. The first part is plausible since power is being held below 5% and an example of an R2 evolution is power changes of less than 10%. However this is incorrect because Low Power Physics testing and Reactor startups are considered R1 evolutions. The second part is correct.

D. Incorrect. The first part is plausible see C(1). The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:42 PM 265

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.2 Equipment Control G2.2.1; Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

(CFR: 41.5 / 41.10 / 43.5 / 43.6 / 45.1)

Importance Rating: RO 4.5 SRO 4.4 Technical

Reference:

AD-ALL-OP-0203,Section 4.4.4, Rev 2, Page 15 Attachment, Rev 2, Page 72 References to be provided: None Learning Objective: PP-LP-2.0, SRO Only Objective 3 Question Origin: New Comments: None Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(6): Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know knowledge of the content of the administrative requirements associated with low power physics testing processes. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(6) by ensuring that the additional knowledge of the procedure requirement for the level of oversight and the reactivity management category is required to correctly answer the written test item.

Thursday, May 19, 2016 5:04:42 PM 266

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 022/NEW/C/A//AD-OP-ALL-0200/NONE//G2.2.15/

Given the following clearance for the 'C' CSIP which ONE of the following completes the statement below?

In accordance with AD-OP-ALL-0200, Clearance and Tagging the required isolation boundary (1) satisfied AND the SRO approver can approve the clearance (2) .

(Reference provided)

A. (1) is (2) as written, this is NOT an Exceptional Clearance B. (1) is (2) when an Exceptional Clearance is documented C. (1) is NOT (2) as written, this is NOT an Exceptional Clearance D. (1) is NOT (2) when an Exceptional Clearance is documented Thursday, May 19, 2016 5:04:42 PM 267

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: AD-OP-ALL-0200, section 5.7 provides the following general guidance for creating a clearance tagging boundary. The order of tagging application should be arranged as follows:

General cautions, notes or other associated clearance, Control switches, Power supplies - e.g., breakers, disconnects, fuses, Lifted leads. Mechanical isolation points (i.e., discharge and suction valves), Support systems (cooling water, air supplies, etc.),

Drains and Vents. If system has greater than 500 psid across boundary valves or fluids above 200°F (93°C), then double valve isolation shall be provided when available.

If double valve isolation is not provided, then designate the clearance as an Exceptional Clearance. In accordance with Attachment 6 clearance hang preparation checklist the details of an exceptional clearance are required to documented in the clearance details and if not complete the clearance should be locked out until exceptional clearance approvals are complete.

A. Incorrect. The first part is correct. The second part is plausible since the VCT temperature is less than 200°F and double valve isolation is not required below 200°F which would allow this clearance to be hung as written; however this is incorrect because the CSIP discharge pressure is greater than 500 psig which requires double isolation or the clearance to be identified as an Exceptional clearance.

B. Correct.

C. Incorrect. The first part is plausible since double valve isolation is not used for this clearance; however this is not correct because double valve isolation is not required since it is not available without shutting down the entire CVCS system. The second part is plausible see A(2).

D. Incorrect. The first part is plausible see C(1). The second part is correct.

Thursday, May 19, 2016 5:04:42 PM 268

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.2 Equipment Control G2.2.15; Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

(CFR: 41.10 / 43.3 / 45.13)

Importance Rating: RO 3.9 SRO 4.3 Technical

Reference:

AD-OP-ALL-0200, Rev 12, Page 38, 40, 117 References to be provided: None Learning Objective: PP-LP-2.4, Objective 7 Question Origin: New Comments: Provide reference...give a drawing that shows system with a clearance and then (Part 1) if the approval of the clearance should be or shouldn't be approved. Then (Part 2) the reason on why the clearance will or will not be approved.

Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(3): Facility licensee procedures required to obtain authority for design and operating changes in the facility. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know knowledge of the content of the procedure requirements to authorize an exceptional clearance. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(3) by ensuring that the additional knowledge of the procedure requirement for the approval of an exceptional clearance is required to correctly answer the written test item.

Thursday, May 19, 2016 5:04:42 PM 269

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 023/PREVIOUS/FUNDAMENTAL//PEP-330/NONE//G2.3.4/

Which ONE of the following completes the statements below in accordance with PEP-330, Radiological Consequences, Attachment 1, Limitations for Lifesaving and Emergency Reentry/Repair Actions?

Emergency worker exposures during life saving missions should be limited to (1)

REM TEDE.

Exposures in excess of 5 REM TEDE shall not be permitted unless specifically authorized by the (2) .

A. (1) 15 (2) Emergency Response Manager B. (1) 15 (2) Site Emergency Coordinator C. (1) 25 (2) Emergency Response Manager D. (1) 25 (2) Site Emergency Coordinator Thursday, May 19, 2016 5:04:42 PM 270

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: PEP-330, Attachment 1 states that 25 Rem TEDE is the maximum dose allowed for life saving missions during a declared emergency and the entry into radiation fields of greater than 25 Rem/hr or exposure in excess of 5 Rem TEDE shall not be permitted unless specifically authorized by the SEC.

A. Incorrect. 15 Rem TEDE is plausible because the dose limit for the lens of the eye is three times the 5 Rem limit, however this is not correct becuase the limit for protecting valuable equipment but life saving is 25 Rem TEDE.

The Emergency Response Manager (ERM) is plausible because this individual authorizes the issuance of KI tablets to offsite personnel assigned to the EOF.

B. Incorrect. 15 Rem TEDE is plausible because the dose limit for the lens of the eye is three times the 5 Rem limit, however this is not correct becuase the limit for protecting valuable equipment but life saving is 25 Rem TEDE.

Site Emergency Coordinator (SEC) is correct.

C. Incorrect. 25 Rem TEDE is correct. The Emergency Response Manager (ERM) is plausible because this individual authorizes the issuance of KI tablets to offsite personnel assigned to the EOF..

D. Correct.

Thursday, May 19, 2016 5:04:43 PM 271

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.3 Radiation Control G2.3.4; Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10)

Importance Rating: RO 3.2 SRO 3.7 Technical

Reference:

PEP-330, Rev. 9, pg 17, Attachment 1 References to be provided: None Learning Objective: EP-LP-2.0, SRO Obj 1 Question Origin: Previous 2013 NRC SRO Exam 98 randomly selected Comments: None Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(4): Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must have the ability to analysis and interpret radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Thursday, May 19, 2016 5:04:43 PM 272

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 024/NEW/FUNDAMENTAL//TE-RP-ALL-2000/NONE//G2.3.7/

With the unit at power, which ONE of the following Operatons tasks (1) would require utilization of a Specific RWP AND (2) the individual required to concur with the RP Manager to approve entry into the applicable area?

A. (1) Entry into Containment to inspect for RCS leakage in the PRZ cubicle.

(2) Shift Manager B. (1) Entry into Containment to inspect for RCS leakage in the PRZ cubicle.

(2) Assistant Operations Manager - Shift C. (1) Entry into a High Radiation Area on the 261' RAB to inspect a CVCS leak.

(2) Shift Manager D. (1) Entry into a High Radiation Area on the 261' RAB to inspect a CVCS leak.

(2) Assistant Operations Manager - Shift Plausibility and Answer Analysis Reason answer is correct: SRWP is required for entry into the Containment bioshield with the Reactor critical. In accordance with AP-545, The Nuclear Shift Manager (SM) responsibility shall not be designated and is responsible for: Concurring with entries into LHRAs (for example, areas inside the bio-shield, or on Elevation 286 or above, when the reactor is critical).

A. Correct.

B. Incorrect. The first part is correct. The second part is plausible since the AOM-Shift is one of the individuals who performs the function of Duty Manager and the Duty Manager is required to be briefed prior to entry into CNMT and other LHRAs with ALARA concerns; however this is incorrect because the AOM-Shift is not required to concur with the entry.

C. Incorrect. The first part is plausible since the equipment is located inside a HRA with ALARA concerns; however a SRWP is not required for routine tasks for operations rounds, HP surveillances, inspections, and routine PM's in an high radiation area. The second part is correct.

D. Incorrect. The first part is plausible see C(1). The second part is plausible see B(2).

Thursday, May 19, 2016 5:04:43 PM 273

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.3 Radiation Control G2.3.7; Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(CFR: 41.12 / 45.10)

Importance Rating: RO 3.5 SRO 3.6 Technical

Reference:

TE-RP-ALL-2000, Attachment 9, Rev 0, Page 37 AP-545, Step 4.11, Rev 57, Page 8 References to be provided: None Learning Objective: PP-LP-3.7 Objective 4 Question Origin: New Comments: None Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(4): Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must have knowledge of the individual responsiblities of the Shift Manager that may not be delegated to another indvidual as they pertain to the process of approving entry into containment at power.

Thursday, May 19, 2016 5:04:43 PM 274

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2016 NRC SRO 025/BANK/C/A//EOP-E-0, USERS GUIDE/NONE/EARLY/G2.4.35/SAT Given the following plant conditions:

- An Inadvertent Safety Injection has occurred from 100% Reactor Power The following conditions exist:

- The crew is terminating Safety Injection

- The OATC has SHUT 1SI-4, BIT Outlet Valve

- 1SI-3, BIT Outlet Valve, will NOT SHUT from the MCB Which ONE of the following identifies (1) the procedure that is being implemented at the time Safety Injection flow is terminated AND (2) the preferred procedural action(s) required for 1SI-3 in accordance with the EOP-User's Guide?

Valve Noun Name:

1SI-1, BIT Inlet Valve 1SI-2, BIT Inlet Valve 1SI-3, BIT Outlet Valve A. (1) EOP-ES-1.1, SI Termination (2) Locally SHUT 1SI-3 B. (1) EOP-ES-1.1, SI Termination (2) Locally SHUT 1SI-1 and 1SI-2 C. (1) EOP-E-0, Reactor Trip Or Safety Injection (2) Locally SHUT 1SI-3 D. (1) EOP-E-0, Reactor Trip Or Safety Injection (2) Locally SHUT 1SI-1 and 1SI-2 Thursday, May 19, 2016 5:04:43 PM 275

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal Plausibility and Answer Analysis Reason answer is correct: With the condition of an inadvertent SI, SI Termination Criteria will be met prior to exiting E-0 so SI flow will be terminated in E-0. The RNO for 1SI-3 directs the operator to locally shut or isolate the BIT Outlet valves but the User's Guide states preferentially to shut either 1SI-3 or 1SI-4 (the BIT Outlet Valves).

A. Incorrect. ES-1.1 is for SI Termination and will be entered for this event. However, with the condition of an inadvertent SI, SI Termination Criteria will be met in E-0 so SI flow will be terminated in E-0. The RNO for 1SI-3 directs locally shut or isolate the BIT Outlet valves.

B. Incorrect. ES-1.1 is for SI Termination and will be entered for this event. However, with the condition of an inadvertent SI, SI Termination Criteria will be met in E-0 so SI flow will be terminated in E-0. The RNO for 1SI-3 directs locally shut or isolate the BIT Outlet valves, so shutting the Inlet valves is incorrect. But it is plausible because the EOP Users Guide discusses closing the Inlet valves if the Outlet valves cannot be shut.

C. Correct.

D. Incorrect. With the condition of an inadvertent SI, SI Termination Criteria will be met in E-0 so SI flow will be terminated in E-0. The RNO for 1SI-3 directs locally shut or isolate the BIT Outlet valves, so shutting the Inlet valves is incorrect. But it is plausible because the EOP Users Guide discusses closing the Inlet valves if the Outlet valves cannot be shut.

Thursday, May 19, 2016 5:04:43 PM 276

QUESTIONS REPORT for Harris 2016 NRC RO SRO Written Exam REV 4 60 Day Submittal 2.4 Emergency Procedures / Plan G2.4.35; Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

(CFR: 41.10 / 43.5 / 45.13)

Importance Rating: RO 3.8 SRO 4.0 Technical

Reference:

EOP-E-0, Step 42, Rev 4, Page 34 EOP-User's Guide, Section 6, step 6.30, Rev 45, Page 54 References to be provided: None Learning Objective: EOP-LP-3.19, Objective 4.bb Question Origin: Bank Comments: None Tier/Group: T3 SRO Justification: 10 CFR Part 55 Content - 43(b)(5): Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Per ES-401 Attachment 2, Clarification Guidance for SRO-only Questions, this question meets the SRO level of knowledge by ensuring that the candidate must know knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose. The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item.

Thursday, May 19, 2016 5:04:43 PM 277