Text

5-Columbia-2017-02 Draft Operating Test
	ML17081A541
Person / Time
Site:	Columbia
Issue date:	03/09/2017
From:	Vincent Gaddy; Operations Branch IV
To:	; Energy Northwest
References
	Download: ML17081A541 (325)
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ES-301 Administrative Topics Outline (Rev 1 - 12/10/16) Form ES-301-1 Facility: Columbia Generating Station Date of Examination: 2/27/17 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic (see Note) Type Describe activity to be performed Code*

DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

A-1 Conduct of Operations

Description:

Determine that criticality has been (D)(R) achieved prior to reaching the minimum K/A: 2.1.37 (4.3 / 4.6) Estimated Critical Position (ECP) and correctly identify the next action to be taken due to being critical outside the ECP.

MAIN TURBINE (MT) LOAD RATE CHANGE A-2 DETERMINATION Conduct of Operations (D)(P)(R)

Description:

Determine the Main Turbine Load K/A: 2.1.25 (3.9 / 4.2) Change Recommendation when raising Main Turbine load from 15% to 85%.

A-3 VALIDATE FUSE INSTALLATION PER PPM 1.3.47 (FUSE REPLACEMENT CONTROL)

Equipment Control (N)(R)

K/A: 2.2.41 (3.5 / 3.9)

Description:

For the RO Candidate, given circumstance requiring fuse replacement and OPEX AR 00314141 an electrical print, determine correct replacement fuse and provide justification.

DETERMINE IF TAGOUT CAN BE HUNG A-4 Radiation Control (D)(R)

Description:

Determination will have to be K/A: 2.3.7 (3.5 / 3.6) made (with justification) whether or not a tag can be hung based on provided Clearance Order, RWP and Survey Map.

NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 3 for ROs) (3)

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

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

Attachment 8

ES-301 Administrative Topics Outline (Rev 1 - 12/10/16) Form ES-301-1 Facility: Columbia Generating Station Date of Examination: 2/27/17 Examination Level: RO SRO Operating Test Number: 1 Administrative Topic (see Note) Type Describe activity to be performed Code*

DETERMINE ACTION BASED ON PLANT CONDITIONS AND PROCEDURAL A-5 GUIDANCE Conduct of Operations (D)(R)

Description:

Given equipment status and an K/A: 2.1.7 (4.4 / 4.7) electrical bus lockout, determine required operator action based on existing plant conditions.

DETERMINE THE OPERABILITY OF THE SLC SYSTEM A-6 Conduct of Operations (D)(R)

Description:

Given the SLC portion of OSP-INST-H101 (Shift and Daily Instrument Checks K/A: 2.1.25 (3.9 / 4.2) for Modes 1, 2 & 3), determine the operability status of the Standby Liquid Control (SLC)

System.

A-7 VALIDATE FUSE INSTALLATION PER PPM 1.3.47 (FUSE REPLACEMENT CONTROL)

Equipment Control (N)(R)

Description:

For the SRO Candidate, given K/A: 2.2.41 (3.5 / 3.9) circumstance requiring fuse replacement and an electrical print, either authorize or do not OPEX AR 00314141 authorize fuse replacement with proposed fuse type. Provide justification.

ESTIMATE MAIN CONDENSER AIR EJECTOR GROSS GAMMA ACTIVITY RATE A-8 AND DETERMINE ACTIONS Radiation Control (D)(P)(R)

Description:

Estimate Main Condenser air ejector Gross gamma activity rate and K/A: 2.3.11 (3.8 / 4.3) determine that a reactor power reduction is required to maintain Main Condenser Gross activity LT the LCO 3.7.5 limit.

A-9 COMPLETE CLASSIFICATION NOTIFICATION FORM (CNF) FOR SAE Emergency Plan (D)(R)

Description:

Given a dose projection printout, K/A: 2.4.41 (2.9 / 4.6) classify the event and complete Classification Notification Form. (Time Critical)

Attachment 8

ES-301 Administrative Topics Outline (Rev 1 - 12/10/16) Form ES-301-1 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).

Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank ( 4 for SROs) (4)

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

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

Attachment 8

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR INITIAL TRAINING COURSE TITLE ADMIN JOB PERFORMANCE MEASURE LESSON TITLE DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

LESSON LENGTH .5 HRS MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001587 Rev. No. 4 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 5/11/06 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY: DATE INSTRUCTIONAL REVIEW BY: DATE APPROVED BY: DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use.

DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Setup Instructions:

Make a copy of the current pull sheet from the simulator. Determine where Minimum ECP is and indicate it on copy of pull sheet by placing an *1 next to the step and at the bottom of the column indicate that a *1 is Minimum ECP. Ensure it is AFTER the step indicated in initial conditions by about four control rods.

Place a *2 ten rods later and make that the maximum ECP.

Fill out the pull sheet pages. The Performed by column is initialed up to control rod 10-47. The Verified column, the Continuous Withdrawal Couple Check column and the Full Out Light columns are initialed to control rod 10-47 by the verifier. Fill in Neutron Flux Response column with a few Ns but mostly Ys.

Sign the Reactivity Manager Review on the bottom right ONLY on those move sheets for which the rod moves have been completed.

Have a separate copy of ATTACHMENT 1 (in the form of a handout) available for student reference.

JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: None Safety Items: None Task Number: RO-0156; SRO-0118 Validation Time: 11 minutes Alternate Path: No Time Critical: No PPM

Reference:

3.1.2 Rev. 81, 9.3.9 Rev 29 Location: Any NUREG 1123 Ref: 2.1.37 4.3 / 4.6 Performance Method: Perform Task Standard:

SRO -State that direction is to be given to the RO to drive control rods in the reverse order until they all are inserted (based on reactor being critical prior to minimum ECP position being reached).

RO - State that control rod withdrawal must be stopped and that the CRS must be informed that the reactor is critical (based on reactor being critical prior to minimum ECP position being reached).

LO001587 Rev. 4 Page 2 of 8

DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

JPM CHECKLIST INITIAL A plant startup is in progress. PPM 3.1.2 has been completed as shown on ATTACHMENT 1.

CONDITIONS:

CRO1 is pulling control rods and notes the following indications:

Time 0953

Coolant Temp 155°F

Control rod 10-47

Control rod position 48

Neutron level 8,000 CPS and rising

Period 145 seconds and stable Control rods have been pulled steadily since starting Group 1 of the Pull Sheet. Control rod motion stopped approximately 1 minute ago.

INITIATING Using the given information, PPM 3.1.2 (Attachment 1), and the supplied pull sheets, determine your next action.

CUE:

When you have determined your next action, write it on the page provided along with the basis for the decision and hand it to the examiner.

LO001587 Rev. 4 Page 3 of 8

DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step Examiner note:

Criticality will be determined using the criteria on Startup Flow Chart 3.1.2 Note N6 (see Attachment 1): Criticality usually occurs in the source range between 1x103 and 1x104 cps. For purposes of this procedure, criticality shall be identified by increasing neutron level, a constant steady period and no simultaneous control rod motion.

The information given identifies 8000 cps and rising (which is between the 1x103 and 1x104 cps) and a constant steady period with no rod motion -

these are indications of a critical reactor.

Examiner note:

If SRO position is being evaluated, perform steps 1 through 3 below else skip to step 4.

Using information provided, Recognized the reactor is critical. S / U*

1 determine that the reactor is critical.

Using information provided, Recognized the Minimum ECP S / U*

determine that criticality was position was not reached during rod 2 achieved prior to rods being withdraw (as shown on control rod withdrawn to the Minimum ECP pull sheets).

position.

Determines action to be taken using Directed that control rod motion be S / U*

PPM 3.1.2 (Attachment 1), steps stopped and that all control rods be 3

Q-14 and Q-16. inserted in reverse order until all control rods are fully inserted.

LO001587 Rev. 4 Page 4 of 8

DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

Examiner note:

If RO position is being evaluated, perform steps 4 through 6 below.

Using information provided, Recognized the reactor is critical. S / U*

4 determine that the reactor is critical.

Using information provided, Recognized the Minimum ECP S / U*

determine that criticality was position was not reached during rod 5 achieved prior to rods being withdraw (as shown on control rod withdrawn to the Minimum ECP pull sheets).

position.

Determines action to be taken using Notified CRS that he/she has stopped S / U*

6 PPM 3.1.2 (Attachment 1), steps control rod withdrawal due to reactor Q-14 and Q-16. being critical outside ECP.

Termination Criteria: When student hands Student JPM Information Card back to the examiner.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001587 Rev. 4 Page 5 of 8

DETERMINE ACTIONS FOR CRITICALITY OUTSIDE OF ECP (EARLY)

RESULTS OF JPM:

Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard:

SRO -State that direction is to be given to the RO to drive control rods in the reverse order until they all are inserted (based on reactor being critical prior to minimum ECP position being reached).

RO - State that control rod withdrawal must be stopped and that the CRS must be informed that the reactor is critical (based on reactor being critical prior to minimum ECP position being reached).

Overall Evaluation Exam Code SAT / UNSAT (Circle One)

Verified Procedure #/Rev. Used for Validation/Critical JPM Completion JPM (Initial Box) Time Time 11 Minutes / NA COMMENTS:

Evaluator's Signature: Date:

LO001587 Rev. 4 Page 6 of 8

STUDENT JPM INFORMATION CARD Initial Conditions:

A plant startup is in progress. PPM 3.1.2 has been completed as shown on ATTACHMENT 1.

CRO1 is pulling control rods and notes the following indications:

Time 0953

Coolant Temp 155°F

Control rod 10-47

Control rod position 48

Neutron level 8,000 CPS and rising

Period 145 seconds and stable Control rods have been pulled steadily since starting Group 1 of the Pull Sheet.

Control rod motion stopped approximately 1 minute ago.

Initiating Cue:

Using the given information, PPM 3.1.2 (Attachment 1), and the supplied pull sheets, determine your next action.

When you have determined your next action, write it on the page provided along with the basis for the decision and hand it to the examiner.

LO001587 Rev. 4 Page 7 of 8

STUDENT JPM ANSWER SHEET NEXT ACTION TO BE TAKEN:

BASIS FOR ACTION:

LO001587 Rev. 4 Page 8 of 8

LO001587 Rev. 4 (Attachment 1)

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LO001587 Rev. 4 (Attachment 1)

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LO001587 Rev. 4 (Attachment 1)

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LO001587 Rev. 4 (Attachment 1)

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LO001587 Rev. 4 (Attachment 1)

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LO001587 Rev. 4 (Attachment 1)

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INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE MAIN TURBINE CHANGE OF LOAD RATE DETERMINATION LESSON LENGTH .5 HRS MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001783 Rev. No. 2 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 10/21/14 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MAIN TURBINE CHANGE OF LOAD RATE DETERMINATION MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

Ensure student has access to a calculator, clear ruler, and a copy of SOP-MT-START (when asked for).

JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: Calculator; Clear Ruler Safety Items: None Task Number: RO-0325 Validation Time: 12 Minutes Alternate Path: No Time Critical: No PPM

Reference:

SOP-MT-START Rev. 26 Location: Any NUREG 1123 Ref: 245000 K5.02 (2.8 / 3.1) Performance Method: Perform Task Standard:

The time needed to change main turbine load has been calculated and written in the space provided on the Student JPM Information Card and is within the range allowed.

LO001783 Page 2 of 6

MAIN TURBINE CHANGE OF LOAD RATE DETERMINATION JPM CHECKLIST INITIAL CONDITIONS: Columbia is in the process of starting up. The Main Turbine is on the line and is currently 15% loaded.

INITIATING CUE: You have been directed to determine the time required to change load from 15% to a load of 85%. Assume a fatigue index of 20,000 cycles. Inform the CRS of your determination when complete by writing it in the space provided on the Student JPM Information Card and handing the card back to the examiner.

Items are Critical Steps Time Step Task Element Performance Standard Evaluators Cue Results 1 Obtains procedure. Recognized SOP-MT-START as being the correct procedure and S/U refers to Attachment 6.1.

2 SOP-MT-START Referred to example at bottom of Attachment 6.1 to determine Attachment 6.1 S/U use of graphs:

Percent Rated Load vs. First Stage Temp Change

Time to change Load vs. First Stage Temp Change 3 Correlated 15% load to a First Stage Steam Temperature of Accept a range of 105° to S/U*

112°F (accept 105°F to 120°F). 120° 4 Correlated 85% load to a First Stage Steam Temperature of Accept a range of 265° to S/U*

272°F (accept 265°F to 280°F). 280° 5 Calculated the temperature difference (272°F - 112°F) to be Accept a range of 145° to S/U*

160°F (accept 145°F to 175°F). 175° 6 Plotted First Stage Steam Temperature Change to Time to Accept a range of 0.95 S/U*

Change Load-Hours using the 20,000 cycles curve and hours (57 minutes) to 1.75 determined that the time to change load is 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (90 hours (105 minutes) minutes)

Termination Criteria: Student hands completed JPM Information Card to the examiner.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001783 Page 3 of 6

MAIN TURBINE CHANGE OF LOAD RATE DETERMINATION LO001783 Page 4 of 6

MAIN TURBINE CHANGE OF LOAD RATE DETERMINATION RESULTS OF JPM:

Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard:

The time needed to change main turbine load has been calculated and written in the space provided on the Student JPM Information Card and is within the range allowed.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001783 Page 5 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

Columbia is in the process of starting up. The Main Turbine is on the line and is currently 15% loaded.

Initiating Cue:

You have been directed to determine the time required to change load from 15% to a load of 85%.

Assume a fatigue index of 20,000 cycles.

Inform the CRS of your determination when complete by writing it in the space provided on the Student JPM Information Card and handing the card back to the examiner.

The time required to change load from 15% to 85% is:

LO001783 Page 6 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE VALIDATE FUSE INSTALLATION PER PPM 1.3.47 (RO)

LENGTH OF LESSON 0.5 Hour INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001860 Rev. No. 1 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Steve Bruce DATE 12/05/16 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

VALIDATE FUSE INSTALLATION PER PPM 1.3.47 MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

Obtain a cleared fuse (BUSS F10A/250V/1A) and a choice of 3 replacement fuses (BUSS F10A/250V/1A, BUSS F10A/250V/10A, and BUSS KTK/600V/20A) for the student to select from.

Provide a copy of EWD-15E-042 (see Attachment 1) and PPM 1.3.47 (Fuse Replacement Control).

JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: Fuses (as specified above) Safety Items: None Task Number: RO-0570, SRO-0212 Validation Time: 5 Minutes Alternate Path: No Time Critical: No PPM

Reference:

PPM 1.3.47 Rev. 11 Location: Any NUREG 1123 Ref: 2.2.41 (3.5 / 3.9) Performance Method: Perform Task Standard: Student determines that the replacement fuse should be a 10A fuse and hands the correct fuse to the CRS (evaluator).

LO001860 Rev. 1 Page 2 of 6

VALIDATE FUSE INSTALLATION PER PPM 1.3.47 JPM CHECKLIST SETUP: Provide the student with EWD-15E-042 (see Attachment 1), PPM 1.3.47, the cleared fuse, and three choices of replacement fuses.

INITIAL CRD-LIS-601B, MS-LIS-200B, and MS-LIS-300B have lost power. Plant conditions are now stable.

CONDITIONS:

EFIN troubleshooting has identified that fuse GG-F02 in H13-P611 has cleared.

INITIATING The CRS directs you to determine the correct replacement fuse.

CUE: Inform the CRS of your decision by completing the Student JPM Information Card provided with justification of your answer and by handing the CRS (evaluator) the card with the correct replacement fuse.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 1 Evaluates EWD-15E-042 Determines that the rating for fuse S/U*

(Attachment 1). GG-F02 is 10A.

2 Evaluates the cleared fuse. Determines that the cleared fuse May not refer to cleared (blown fuse) S/U rating is 1A. in determining the correct fuse.

3 Evaluates available replacement Determines that the fuse marked as a May refer to PPM 1.3.47 to help S/U*

fuses by looking at the rating printed 10A fuse is the correct fuse. identify correct fuse.

on the fuse.

4 Informs the CRS. Informs the CRS that the Evaluator may need to prompt the S/U*

replacement for fuse GG-F02 should candidate to provide the correct be a 10A fuse and provides the replacement fuse.

correct replacement fuse to the CRS.

Termination Criteria: The student turns in the completed answer sheet and selected fuse for replacement.

Termination Cue: This completes the JPM.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001860 Rev. 1 Page 3 of 6

RESULTS OF JPM:

VALIDATE FUSE INSTALLATION Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard: Candidate determines that the replacement fuse should be a 10A fuse and hands the correct fuse to the CRS (evaluator).

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001860 Rev. 1 Page 4 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

CRD-LIS-601B, MS-LIS-200B, and MS-LIS-300B have lost power. Plant conditions are now stable.

EFIN troubleshooting has identified that fuse GG-F02 in H13-P611 has cleared.

Initiating Cue:

The CRS directs you to determine the correct replacement fuse.

Inform the CRS of your decision by completing the Student JPM Information Card provided with justification of your answer and by handing the CRS (evaluator) the card with the correct replacement fuse.

LO001860 Rev. 1 Page 5 of 6

STUDENT JPM ANSWER SHEET The correct replacement fuse is: ________________

Provide your justification below:

LO001860 Rev. 1 Page 6 of 6

EWD-15E-042 LO001860 Rev. 1 (ATTACHMENT 1)

Page 1 of 1

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR INITIAL TRAINING COURSE TITLE ADMIN JOB PERFORMANCE MEASURE LESSON TITLE DETERMINE IF TAGOUT CAN BE HUNG (Admin)

MAXIMUM STUDENTS LESSON LENGTH .5 HRS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001585 Rev. No. 4 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 9/26/06 REVISED BY Dave E. Crawford DATE 12/7/16 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use.

DETERMINE IF TAGOUT CAN BE HUNG (Admin)

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Setup Instructions:

Print out copy of ATTACHMENT 1 for student use.

JPM Instructions:

Verify current procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: N/A Safety Items: N/A Task Number: RO-1293 Validation Time: 15 minutes Prerequisite Training: N/A Time Critical: No

Reference:

GEN-RPP-01 R8; GEN-RPP-02 R34 Location: Any NUREG 1123 Ref: 2.3.7 (3.5 / 3.6) Performance Method: Perform Task Standard: Candidate fills out the JPM Answer Sheet and indicates that the tagout cannot be hung due to HPCS-V-4 being in a High Radiation area and because the associated RWP is only authorized for Non-High Rad operation/investigation.

LO001585 Page 2 of 14

DETERMINE IF TAGOUT CAN BE HUNG (Admin)

JPM CHECKLIST INITIAL Columbia Generating Station is shutdown for a refueling outage.

CONDITIONS:

INITIATING You have been directed by the Control Room Supervisor to hang tagout D-HPCS-V-102R18-001. Health Physics has been contacted CUE: and directed the use of RWP-30001423 for the purpose of hanging tags in the Reactor Building.

Review the task and from the information provided, fill out the attached sheet indicating your ability or inability to perform the task assigned.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step Reviews the Recognizes that a tag is to be placed 1 Clearance Tag Hang on the handwheel for HPCS-V-4. S/U List Reviews the Survey Recognizes the area dose rates around 2 Map for the Reactor the handwheel for HPCS-V-4 exceed S/U Building 541 areas. 100mr/hr.

Reviews the Recognizes that the RWP does not 3 Radiation Work allow for entry into a high radiation S/U Permit (RWP) area (> 100 mr/hr).

Document whether Documents that the task cannot be or not the task can performed on the answer sheet. S/U*

4 be completed given the current conditions.

LO001585 Page 3 of 14

DETERMINE IF TAGOUT CAN BE HUNG (Admin)

Documents the basis Documents that the task cannot be for why the task completed because the RWP does not S/U*

cannot be allow entry into high radiation areas completed. on the answer sheet.

Termination Criteria: Candidate hands the examiner the completed answer sheet.

RECORD TERMINATION TIME: _________

Transfer to Results of JPM page the following information: Procedures validated prior to use; Comments from marked up evaluators procedure copy; Unsatisfactory critical tasks; Total JPM time; Marked Up procedure and remaining JPM pages may be discarded.

LO001585 Page 4 of 14

DETERMINE IF TAGOUT CAN BE HUNG (Admin)

RESULTS OF JPM:

Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard: Candidate fills out the JPM Answer Sheet and indicates that the tagout cannot be hung due to HPCS-V-4 being in a High Radiation area and because the associated RWP is only authorized for Non-High Rad operation/investigation.

Overall Evaluation Exam Code SAT / UNSAT (Circle One)

Verified Procedure #/Rev. Used for Validation/Critical JPM Completion JPM (Initial Box) Time Time 15 Minutes / NA COMMENTS:

Evaluator's Signature: Date:

LO001585 Page 5 of 14

STUDENT JPM INFORMATION CARD Initial Conditions:

Columbia Generating Station is shutdown for a refueling outage.

Cue:

You have been directed by the Control Room Supervisor to hang tagout D-HPCS-V-102R18-001.

Health Physics has been contacted and directed the use of RWP-30001423 for the purpose of hanging tags in the Reactor Building.

Review the task and from the information provided, fill out the attached sheet indicating your ability or inability to perform the task assigned.

LO001585 Page 6 of 14

STUDENT JPM ANSWER SHEET I will be able to perform the assigned task (Initial):

I will not be able to perform the assigned task (Initial): ____________________

Document your justification for the response above:

- - - SURVEY MAPS REDACTED*****

- - - SURVEY MAPS REDACTED*****

- - - SURVEY MAPS REDACTED*****

- - - SURVEY MAPS REDACTED*****

- - - SURVEY MAPS REDACTED*****

LO001585 Page 7 of 14

LO001585 Attachment 1 Page 10 of 14

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LO001585 Attachment 1 Page 12 of 14

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LO001585 Attachment 1 Page 14 of 14

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE Determine Action Based on Plant Conditions and Procedural Guidance LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001791 Rev. No. 2 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 02/15/15 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

Have copies of ABN-CORE and ABN-RRC-LOSS available for student reference.

JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: None Safety Items: None Task Number: SRO-0659 Validation Time: 15 minutes Alternate Path: No Time Critical: No PPM

Reference:

ABN-CORE Rev. 16; Location: Any ABN-RRC-LOSS Rev. 13 NUREG 1123 Ref: 2.1.7 (4.4 / 4.7) Performance Method: Perform Task Standard:

A determination is made (as documented on Student JPM Answer Sheet) and based on information provided that a Reactor Scram is required as directed by ABN-CORE (Step 3.2).

LO001791 Rev. 1 Page 2 of 6

JPM CHECKLIST INITIAL With Columbia operating at full power, a common cause failure of the OPRMs required the CRS to direct placing both OPRM CONDITIONS Manual Enable/Bypass switches in the BYPASS position. Three hours later a lockout on SH-5 occurs.

The following plant conditions are reported by CRO1: Reactor power is 50%. Active loop drive flow is 27500 gpm. Rod line is 85%.

INITIATING From the information given, determine the procedural action required. On the Student JPM Answer Sheet provided, indicate what that action is and provide the procedural reference (including step) for that action. When completed hand the Student JPM Answer CUE: Sheet back to the examiner.

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 1 Determine procedure to enter. Determines that entry into ABN-RRC-LOSS is S/U required due to lockout on SH-5.

2 Determine procedure transition. Determines entry into ABN-CORE is directed S/U from ABN-RRC-LOSS.

3 Plots operating points given. Determines operation is now in Region A of the S/U Single Loop Power to Flow map (using Attachment 6.1 of ABN-RRC-LOSS).

4 Determines required procedural Refers to Immediate Operator Actions of ABN-S/U action based on conditions given. CORE (Step 3.2) and determines that a manual reactor scram is required due to operating in Region A of the Power to Flow map and the OPRM is inoperable.

5 Fills out answer sheet. Indicates that a Manual Scram is required per S/U*

ABN-CORE (Step 3.2).

Termination Criteria: Candidate hands in the completed JPM Answer Sheet.

Transfer the following to the Results of JPM page: Any Unsat step(s) and JPM completion time.

LO001791 Rev. 1 Page 3 of 6

JPM RESULTS:

DETERMINE ACTION BASED ON PLANT CONDITIONS AND PROCEDURAL GUIDANCE Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: A determination is made (as documented on Student JPM Answer Sheet) and based on information provided that a Reactor Scram is required as directed by ABN-CORE (Step 3.2).

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001791 Rev. 1 Page 4 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

With Columbia operating at full power, a common cause failure of the OPRMs required the CRS to direct placing the OPRM Manual Enable/Bypass switches in the BYPASS position.

Three hours later a lockout on SH-5 occurs.

The following plant conditions are reported by CRO1:

Reactor power is 50%.

Active loop drive flow is 27500 gpm

Rod line is 85%

Initiating Cue:

From the information given, determine the procedural action required.

On the Student JPM Answer Sheet provided, indicate what that action is and provide the procedural reference (including step) for that action.

When completed hand the Student JPM Answer Sheet back to the examiner.

LO001791 Rev. 1 Page 5 of 6

STUDENT JPM ANSWER SHEET The action required is:

The action is per procedure/step: _____________/_______

LO001791 Rev. 1 Page 6 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR/STA REQUALIFICATION TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE DETERMINE THE OPERABILITY OF THE SLC SYSTEM (ADMIN)

MAXIMUM STUDENTS LESSON LENGTH .5 HRS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001574 Rev. No. 1 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/29/16 REVISED BY DATE TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use.

DETERMINE THE OPERABILITY OF THE SLC SYSTEM MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

ADMIN JPM - NO SIMULATOR SETUP NEEDED.

Special Setup Instructions:

Print out a copy of OSP-INST-H101 and fill in blocks 53, 54 and 55 with required information. Provide student a copy of Attachments 9.6 and 9.7 in addition to Page containing steps 53, 54 and 55.

JPM Instructions:

Verify current procedure against JPM and ensure procedure steps match. If steps have changed, the JPM should be revised.

The evaluator and student will use current procedure. The evaluator should mark off steps as they are completed, note comments, and transfer the comments to the Results of JPM page.

Tools/Equipment: N/A Safety Items: N/A Task Number: SRO-0163 Validation Time: 5 Minutes Prerequisite Training: N/A Time Critical: NO PPM

Reference:

OSP-INST-H101 Rev. 86 Location: Simulator/Classroom NUREG 1123 Ref: 2.2.40 3.4 / 4.7 Performance Method: Perform LO0001574 Page 2 of 6

DETERMINE THE OPERABILITY OF THE SLC SYSTEM JPM CHECKLIST INITIAL The plant is operating at 100% power. Per the SLC placard in the P603 ARP holder, CONDITIONS: SLC Concentration is 14.3 percent.

INITIATING CUE: You are reviewing OSP-INST-H101, the Shift and Daily Instrument Checks (MODES 1, 2, & 3) for a day shift review. Evaluate steps 53, 54, and 55 to determine if SLC is operable. Notify the CRS (examiner) of your answer by checking the appropriate block on your JPM ANSWER SHEET.

Items are Critical Steps Comments Element Standard Sat/Unsat RECORD START TIME: __________

Determines SLC From step 53 determines the need to operability use Att. 9.6 for comparison. S/U Using ATT. 9.6, determines SLC concentration/temperature is outside S/U*

of the acceptable region. Refers to NOTE 4.

From step 55 determines the need to S/U use Att. 9.7 for comparison.

Using Att. 9.7 determines SLC Tank volume is acceptable. S/U*

Notifies the CRS that SLC is NOT S/U*

operable.

Termination Criteria: Student hands JPM ANSWER SHEET (with a selected response) to the examiner.

RECORD TERMINATION TIME: _________

Transfer to Results of JPM page the following information: Procedures validated prior to use; Comments from marked up evaluators procedure copy; Unsatisfactory critical tasks; Total JPM time; Marked Up procedure and remaining JPM pages may be discarded.

LO0001574 Page 3 of 6

DETERMINE THE OPERABILITY OF THE SLC SYSTEM RESULTS OF JPM:

Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard: OSP-INST-H101 is used to determine that the SLC system is NOT operable and NOT OPERABLE is checked on the JPM ANSWER SHEET.

Overall Evaluation Exam Code SAT / UNSAT (Circle One)

Verified Procedure #/Rev. Used for Validation/Critical JPM Completion JPM (Initial Box) Time Time 5 Minutes / NA COMMENTS:

Evaluator's Signature: Date:

LO0001574 Page 4 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

The plant is operating at 100% power.

Per the SLC placard in the P603 ARP holder, SLC Storage Tank Concentration is 14.3 percent.

Cue:

You are reviewing OSP-INST-H101, the Shift and Daily Instrument Checks (MODES 1, 2, & 3), for a day shift review.

Evaluate steps 53, 54, and 55 to determine if SLC is operable.

Notify the CRS (examiner) of your answer by checking the appropriate block on the JPM ANSWER SHEET.

LO0001574 Page 5 of 6

JPM ANSWER SHEET The SLC System is:

OPERABLE NOT OPERABLE LO0001574 Page 6 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE VALIDATE FUSE INSTALLATION PER PPM 1.3.47 (SRO)

LENGTH OF LESSON 0.5 Hour INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001861 Rev. No. 1 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Steve Bruce DATE 12/05/16 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

VALIDATE FUSE INSTALLATION PER PPM 1.3.47 MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

Obtain a cleared fuse (BUSS F10A/250V/1A) and a choice of 3 replacement fuses (BUSS F10A/250V/1A, BUSS F10A/250V/10A, and BUSS KTK/600V/20A) for the student to select from.

Provide a copy of EWD-15E-042 (see Attachment 1) and PPM 1.3.47 (Fuse Replacement Control).

JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: Fuses (as specified above) Safety Items: None Task Number: RO-0570, SRO-0212 Validation Time: 5 Minutes Alternate Path: No Time Critical: No PPM

Reference:

PPM 1.3.47 Rev. 11 Location: Any NUREG 1123 Ref: 2.2.41 (3.5 / 3.9) Performance Method: Perform Task Standard: Student determines that the replacement fuse should be a 10A fuse and hands the correct fuse to the SM (evaluator).

LO001860 Rev. 1 Page 2 of 6

VALIDATE FUSE INSTALLATION PER PPM 1.3.47 JPM CHECKLIST SETUP: Provide the student with EWD-15E-042 (see Attachment 1), PPM 1.3.47, the cleared fuse, and three choices of replacement fuses.

INITIAL CRD-LIS-601B, MS-LIS-200B, and MS-LIS-300B have lost power. Plant conditions are now stable.

CONDITIONS:

EFIN troubleshooting has identified that fuse GG-F02 in H13-P611 has cleared.

The SM directs you as the FIN Team SRO to independently validate the correct replacement fuse.

INITIATING CUE: Inform the SM of your decision by completing the JPM Answer Sheet provided with justification of your answer and by handing the SM (evaluator) the correct replacement fuse.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 1 Evaluates EWD-15E-042 Determines that the rating for fuse S/U*

(Attachment 1). GG-F02 is 10A.

2 Evaluates the cleared fuse. Determines that the cleared fuse May not refer to cleared (blown fuse) S/U rating is 1A. in determining the correct fuse.

3 Evaluates available replacement Determines that the fuse marked as a May refer to PPM 1.3.47 to help S/U*

fuses by looking at the rating printed 10A fuse is the correct fuse. identify correct fuse.

on the fuse.

4 Informs the SM. Informs the SM that the replacement Evaluator may need to prompt the S/U*

for fuse GG-F02 should be a 10A candidate to provide the correct fuse and provides the correct replacement fuse.

replacement fuse to the SM.

Termination Criteria: The student turns in the completed answer sheet and selected fuse for replacement.

Termination Cue: This completes the JPM.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001860 Rev. 1 Page 3 of 6

RESULTS OF JPM:

VALIDATE FUSE INSTALLATION PER PPM 1.3.47 Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard: Candidate determines that the replacement fuse should be a 10A fuse and hands the correct fuse to the SM (evaluator).

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001860 Rev. 1 Page 4 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

CRD-LIS-601B, MS-LIS-200B, and MS-LIS-300B have lost power.

Plant conditions are now stable.

EFIN troubleshooting has identified that fuse GG-F02 in H13-P611 has cleared.

Initiating Cue:

The SM directs you as the FIN Team SRO to independently validate the correct replacement fuse.

Inform the SM of your decision by completing the JPM Answer Sheet provided with justification of your answer and by handing the SM (evaluator) the correct replacement fuse.

LO001860 Rev. 1 Page 5 of 6

STUDENT JPM ANSWER SHEET The correct replacement fuse is: ________________

Provide your justification below:

LO001860 Rev. 1 Page 6 of 6

EWD-15E-042 LO001860 Rev. 1 (ATTACHMENT 1)

Page 1 of 1

INSTRUCTIONAL COVER SHEET PROGRAM TITLE INITIAL LICENSED OPERATOR TRAINING COURSE TITLE ADMIN JOB PERFORMANCE MEASURE ESTIMATE MAIN CONDENSER AIR EJECTOR GROSS GAMMA ACTIVITY LESSON TITLE RATE AND DETERMINE ACTIONS (ADMIN)

LESSON LENGTH .5 HRS MAXIMUM STUDENTS 1 INSTRUCTIONAL MATERIALS INCLUDED Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code Rev. No.

JPM PQD Code LO001590 Rev. No. 5 Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 10/21/06 REVISED BY Dave Crawford DATE 12/19/16 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use.

ESTIMATE MAIN CONDENSER AIR EJECTOR GROSS GAMMA ACTIVITY RATE AND DETERMINE ACTIONS MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Setup Instructions:

Candidate needs a calculator and access to ABN-OG.

JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: N/A Safety Items: N/A Task Number: SRO-0658 Validation Time: 10 minutes Alternate Path: No Time Critical: No PPM

Reference:

ABN-OG Rev. 4 Location: Classroom NUREG 1123 Ref: 271000A2.04 (3.7 / 4.1) Performance Method: Perform Task Standard: Candidate fills out the JPM Answer Sheet and has determined that a power reduction per PPM 3.2.4 is required to maintain Main Condenser Gross gamma activity LT 332 mCi/sec.

LO001590 Rev. 5 Page 2 of 7

ESTIMATE MAIN CONDENSER AIR EJECTOR GROSS GAMMA ACTIVITY RATE AND DETERMINE ACTIONS JPM CHECKLIST INITIAL Columbia is operating at full power. Various alarms are locked in due to suspected fuel pin damage. Offgas system parameters are CONDITIONS: as follows: OFFGAS POST TREATMENT RADIATION MONITOR, OG-RIS-601A, is in alarm. OFFGAS SYSTEM EXHAUST FLOW, OG-FR-620, is reading 43 SCFM. SJAE CONDENSER OUTLET RADIATION MONITOR, OG-RR-604, is reading 7760 mr/hr.

INITIATING Based on the above information, determine what action, if any, should be taken. Fill in the result of your conclusion on the CUE: attachment provided.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 1 Determine procedure. Recognizes entry condition into ABN-OG and refers to procedure. S/U 2 ABN-OG Step 4.1.4 Main Condenser Gross gamma activity = 7760 mr/hr Estimate Main Condenser air times 43 SCFM divided by 1000 OR S/U*

ejector gross gamma activity rate Main Condenser Gross gamma activity = 333.68 using the following formula: mCi/sec (GT 332mCi/sec).

[OG Pretreatment (mRem/hr) (OG-RR-604)] X [OG System flow (scfm) (OG-FR-620)] divided by 1000 = Main Condenser Gross gamma activity (mCi/sec).

LO001590 Rev. 5 Page 3 of 7

ESTIMATE MAIN CONDENSER AIR EJECTOR GROSS GAMMA ACTIVITY RATE AND DETERMINE ACTIONS Time JPM Task Element Performance Standard Evaluators Cue Results Step 3 ABN-OG Step 4.1.5 Based on a Main Condenser Gross gamma activity S/U*

reading of 333.68 mCi/sec, candidate determines that Determines required action. a power reduction per PPM 3.2.4 to maintain Main Condenser Gross gamma activity LT 332 mCi/sec is required.

Termination Criteria: Hands the JPM Answer Sheet to the examiner.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001590 Rev. 5 Page 4 of 7

RESULTS OF JPM:

ESTIMATE MAIN CONDENSER AIR EJECTOR GROSS GAMMA ACTIVITY RATE AND DETERMINE ACTIONS Examinee (Please Print): ______________________________________________

Evaluator (Please Print): ______________________________________________

Task Standard: Candidate fills out the JPM Answer Sheet and has determined that a power reduction per PPM 3.2.4 is required to maintain Main Condenser Gross gamma activity LT 332 mCi/sec.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001590 Rev. 5 Page 5 of 7

STUDENT JPM INFORMATION CARD Initial Conditions:

Columbia is operating at full power.

Various alarms are locked in due to suspected fuel pin damage.

Offgas system parameters are as follows:

OFFGAS POST TREATMENT RADIATION MONITOR, OG-RIS-601A, is in alarm.

OFFGAS SYSTEM EXHAUST FLOW, OG-FR-620, is reading 43 SCFM.

SJAE CONDENSER OUTLET RADIATION MONITOR, OG-RR-604, is reading 7760 mr/hr.

Cue:

Based on the above, determine what action, if any, should be taken.

Fill in the result of your conclusion on the JPM Answer Sheet. Hand the JPM Answer Sheet to your examiner when complete.

Page 6 of 7

JPM ANSWER SHEET INITIAL HERE IF NO ACTIONS ARE REQUIRED:

REASON NO ACTIONS ARE REQUIRED:

INITIAL HERE IF ACTIONS ARE REQUIRED:

ACTION(S) IF REQUIRED AND REASON FOR ACTION:

Page 7 of 7

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE COMPLETE CLASSIFICATION NOTIFICATION FORM (SAE) (SRO) (TC)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LR001509 Rev. No. 6 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 05/17/01 REVISED BY Dave E. Crawford DATE 01/17/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

Run URI based upon initial conditions. Print the URI Dose Assessment form. Give these to the student after reading the initial conditions and initiating cue.

JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: None Safety Items: None Task Number: SRO- 0529, 0638 Validation Time: 12 Minutes Alternate Path: No Time Critical: Yes (30 Minutes)

PPM

Reference:

PPM 13.8.1 Rev. 36, CNF Rev. 24 Location: Any NUREG 1123 Ref: 2.4.41 (2.9/4.6) Performance Method: Perform Task Standard: Performs Site Area Emergency event classification within 15 minutes of when Initial Cue is provided and completes Classification Notification Form with required information within 15 minutes of event classification.

LR001509 Page 2 of 6

JPM CHECKLIST INITIAL The plant has experienced an event that has resulted in the following conditions: The plant scrammed an hour ago. A release has CONDITIONS: been ongoing for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 44 minutes. Turbine Building exhaust flow is 355,000 cfm with activity reading of 1.30E-01 µCi/cc.

Wind direction is from 300°, wind speed is 7 mph, and there is no precipitation. Stability class is A. URI Dose Assessment has been performed.

INITIATING The Shift Manager has directed you to complete the Classification Notification Form based only upon the results of the completed CUE: Dose Assessment. This is the initial classification of this event. Present the completed CNF to the Shift Manager for signature. This is a Time Critical JPM and your time starts now.

EVALUATOR Record start time: ___________

NOTE:

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat EVALUATOR NOTE: RECORD CLASSIFICATION TIME: ___________ FOR STEP 1 BELOW.

1 Classifies event Classifies event as a SAE based upon Thyroid CDE dose at 1.2 miles of S/U*

GT 500 mrem within 15 minutes of start time.

EVALUATOR NOTE: THIS STOPS THE FIRST 15 MINUTE CLOCK AND STARTS THE NEXT 15 MINUTE CLOCK 2 Completes Classification Fills in following information on the Notification Form. CNF:

3 Block 1 Checks b. (Drill) S/U 4 Block 2 Enters a 1 S/U 5 Block 3 Enters name and number S/U 6 Block 4 Checks a. (Initial Classification) and S/U*

enters date and time LR001509 Page 3 of 6

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 7 Block 5 Checks c. (Site Area Emergency) S/U*

8 Block 6 N/A (left blank) S/U 9 Block 7 Checks No S/U 10 Block 8 Enters 7 for Wind Speed S/U 11 Block 8 Enters 300 for degrees S/U 12 Block 8 Checks No for Precipitation S/U 13 Block 8 Enters A as Stability Classification S/U 14 Block 9 Checks Release S/U*

15 Block 10 Checks Airborne S/U 16 Block 11 Enters a time for Estimated Start of S/U Release (part of initial conditions) 17 Block 12 Checks No S/U 18 Block 13 Enters 5.1.S.2 for EAL# with short S/U*

description 19 Block 14 Checks a., b., or c. S/U Termination Criteria: Student hands the instructor the completed CNF.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LR001509 Page 4 of 6

RESULTS OF JPM COMPLETE CLASSIFICATION NOTIFICATION FORM (SAE)

Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: Performs Site Area Emergency event classification within 15 minutes of when Initial Cue is provided and completes Classification Notification Form with required information within 15 minutes of event classification.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LR001509 Page 5 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

The plant has experienced an event that has resulted in the following conditions:

The plant scrammed an hour ago

A release has been ongoing for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 44 minutes

Turbine Building exhaust flow is 355,000 cfm with activity reading of 1.30E-01

µCi/cc

Wind direction is from 300°

Wind speed is 7 mph

No precipitation

Stability class is A AN URI Dose Assessment has been performed.

Initiating Cue:

The Shift Manager has directed you to complete a Classification Notification Form based only on the results of the completed Dose Assessment.

This is the initial classification of this event.

Present the completed CNF to the Shift Manager for signature.

THIS IS A TIME CRITICAL JPM and your time starts now LR001509 Page 6 of 6

1 Type of Event: 2 COLUMBIA GENERATING STATION

a. Emergency

b. Drill CLASSIFICATION NOTIFICATION FORM (CNF) No:

3 4 Classification/Status Time: Date:

Notification Provided By: Phone:

(Emergency Director) a. Initial Classification Name (Print): (509) b. Reclassification

c. Termination

d. PAR Changes/Additions

e. Information Section Map 5 a. UNUSUAL EVENT No Offsite Protective Actions Recommended

b. ALERT No Offsite Protective Actions Recommended

c. SITE AREA EMERGENCY Automatic Protective Action Recommendation EVACUATE:

Columbia River

Ringold Fishing Area

Wahluke Hunting Area

Schools in EPZ

Horn Rapids Recreation Area/ORV Park

d. GENERAL EMERGENCY Automatic Protective Action Recommendation EVACUATE:

Columbia River

Ringold Fishing Area

Wahluke Hunting Area

Schools in EPZ

Horn Rapids Recreation Area/ORV Park 8 Meteorological Data: 6 PROTECTIVE ACTION RECOMMENDATIONS Wind Speed: mph from degrees IF a General Emergency is declared, Precipitation: Yes No THEN Refer to PPM 13.2.2 Determining PARs.

Stability Classification IF A GE is NOT declared, This section is Not Applicable 9 No Release(Block 10,11&12 are N/A) Release Basis for PARs: Not Applicable Radiological Plant Type of release: 11 Estimated Start of Release: 0-2 miles 2-10 miles 10 N/A N/A All Sections Section 1 Section 2 Section 3 Section 4 Monitor & Monitor & Monitor & Monitor &

Airborne Time/Date:

Prepare Prepare Prepare Prepare Water Release Terminated: Shelter In Shelter In Shelter In Shelter In Shelter In Time/Date: Place Place Place Place Place Evacuate Evacuate Evacuate Evacuate Evacuate 12 State Criteria met for administering KI(Information only) 7 Security Event: Yes No N/A Responding personnel are to report to:

No On-Site Facilities Yes 250 mrem/hr thyroid Alternate Facilities, Energy Northwest Office Complex, 1.4 x 10-7 µci/cc I-131 3000 George Washington Way Unfiltered or unmonitored release 13 EAL# Description; Additional Information; 14 Prognosis of Situation: a. Unknown b. Stable c. Escalating d. Improving 15 Emergency Director Approval Signature:

24075 R24

Completion of Classification Notification Form (CNF)

Completing the form Block 1. Type of event: For actual emergencies, the block Emergency should be checked.

For drills or exercises, the block Drill should be checked.

Block 2. Classification Form Number: This is a sequential number indicating the order of offsite notifications.

The first CNF is #1 followed by #2, etc.

Block 3. Notification provided by. This is the name of the Emergency Director providing the information for the Crash call. Phone number is the number at which the notifier can be contacted.

Block 4. Classification/Statues: a-e.

Item a or b: The time listed is the time at which the ED declares the emergency classification or upgrade.

This time starts the 15-minute notification requirement.

Item c.: Termination, no Classification Level should exist or be marked.

A CNF and Crash must be initiated at the termination of a drill or actual event.

Item d.: If additional PARs are required after the CNF for the GE has been transmitted, complete this block.

The need for additional PARs requires notifications be completed within 15 minutes of the time in the block.

Item e.: Periodic information updates such as release information, KI, prognosis, and changes in Met conditions should be provided at least once an hour.

Block 5. Check block for appropriate emergency classification.

(UNUSUAL EVENT, ALERT, SITE AREA EMERGENCY, GENERAL EMERGENCY)

Block 6. When a General Emergency is declared, Refer to PPM 13.2.2 Determining Protective Action Recommendations, Check applicable sections/actions and communicated during the Crash call for the GE.

If a GE is NOT declared, this section is N/A and does NOT need to be filled in.

Block 7. Identify whether the event is security based (Auto Dialer Scenario 191) and reporting location for Offsite Response Organization (ie. County and State Personnel) responding to CGS.

Block 8. Enter Meteorological data. Following a release, if meteorological data changes ensure additional PARs are considered and provide offsite notification. To convert Delta T to stability class, refer to PPM 13.8.1.

Block 9. If there is a No RELEASE, then blocks 10, 11 & 12 are N/A.

If there is a RELEASE then enter information in blocks 10, 11 & 12.

If RELEASE starts after CNF and CRASH notification has been completed, then provide new CNF and Crash notifications to offsite agencies as soon as RELEASE Criteria has been met.

Block 10. If there is a RELEASE, mark it as airborne or water.

Block 11. If there is a RELEASE, enter the start time. Enter stop time following release termination.

Block 12. The block with information on the States criteria for KI is an information notification not a PAR.

Block 13. Enter the EAL number. Provide a short description of the event. Do not use jargon and avoid acronyms.

Block 14. Enter Prognosis of Situation. This is a judgment call primarily relating to the condition of the reactor.

Block 15. Ensure the Emergency Director has signed the form prior to transmittal to the offsite agencies.

Additional information to consider when completing the CNF

CNF must be filled out in entirely prior to transmittal to offsite agencies. Transmittal of the CNF should occur prior to initiation of each Crash Call. The requirement to complete 15-minute notifications to the offsite agencies should not be delayed if the time needed to complete the form would impact the notification requirement. In cases where the Crash Call is initiated prior to transmittal, the form should be filled out and transmitted as soon as possible.

When the Control Room is providing emergency classifications, they will ensure the SCC has received the CNF at which time the SCC will follow up with the offsite agencies to ensure they have received the information. If the SCC is not available, the Control Room Notifier must provide the information block by block to the offsite agencies.

If the CNF information is being communicated from the EOF or TSC, all information on the form must be verbally communicated. When communicating the CNF information, it must be communicated block by block for each of the blocks.

If an error on the CNF is recognized during the Crash Call, the correction should be noted on the CNF, initialed, and communicated during the Crash Call.

If an error is recognized in block 4, 5, 6, 8, 9, 10, 11, 12 or 13 after the Crash Call has concluded, a new corrected CNF with the next sequential number should be completed, transmitted, and followed up with a Crash Call.

24075 R24

ES-301 Control Room/In-Plant Systems Outline (Rev 1 - 12/10/16) Form ES-301-2 Facility: Columbia Generating Station Date of Examination: 2/27/17 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S-1: TRANSFER BUS SM-3 FROM TR-S TO TR-N

Description:

Transfer 4160 VAC Bus SM-3 from the Startup Transformer to (D)(S) 6 the Normal Transformer.

K/A: 262001.A4.04 (3.6/3.7)

S-2: (RHR) RESPOND TO LOSS OF SHUTDOWN COOLING

Description:

Restore Residual Heat Removal (RHR) Loop B shutdown cooling per SOP-RHR-SDC (RHR Loop B Shutdown Cooling Quick Restart). (A)(L)(N)(S) 4 Following pump start, insufficient pump head (and consequently flow) will necessitate securing pump per CAUTION in procedure.

K/A: 205000.A2.06 (3.4/3.5)

S-3: (HPCS) HPCS SYSTEM INITIATION

Description:

Initiate High Pressure Core Spray (HPCS) system per SOP-HPCS-INJECTION and restore RPV level back to directed band. Following (A)(N)(EN) start of the HPCS pump its minimum flow valve will fail to automatically close 2 (L)(S) once RPV injection has occurred (resulting in a lower injection rate into the RPV). Valve must be manually closed to maximize injection.

K/A: 209002.A4.04 (3.1/3.1)

S-4: (CR HVAC) INITIATE CR HVAC MANUAL PRESSURIZATION MODE

Description:

Place both trains of Control Room Ventilation in the Manual Pressurization Mode of operation per SOP-HVAC/CR-OPS (inlet damper for (A)(D)(EN) 9 one of the Control Room Emergency Filter Units fail to auto open and must be (S) opened manually).

K/A: 290003.A4.01 (3.2/3.2)

S-5: (RB HVAC) RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC

Description:

Restart Reactor Building (RB) HVAC using RB Outside Air Fan (D)(P)(S) 5 1A and RB Exhaust Air Fan 1A per SOP-HVAC RB-RESTART-QC to re-establish Secondary Containment integrity.

K/A: 290001.A4.01 (3.3/3.4)

S-6: (DEH) LOWER RPV PRESSURE USING DEH

Description:

Recognize that auto control of bypass valves to lower RPV pressure to a target of 550 psig does not work and that the manual lowering of (A)(D)(L)(P) 3 RPV pressure at a rate LE 50 psig per minute through manual control of (S)

Bypass Valves would be required.

K/A: 241000.A4.02 (4.1/4.1)

S-7: (RPS) RESTORE RPS A FROM ALTERNATE POWER SOURCE

Description:

Transfer RPS A to its Alternate power supply by performing (D)(P)(S) 7 subsequent steps in ABN-RPS.

K/A: 212000.A2.01 (3.7/3.9)

Attachment 9 Page 1 of 6

ES-301 Control Room/In-Plant Systems Outline (Rev 1 - 12/10/16) Form ES-301-2 S-8: (Fire Protection) ABN-FIRE Immediate Actions

Description:

Given the report of heavy smoke in the field and the observance (N)(S) 8 of several fire alarms, perform the Immediate Actions of ABN-FIRE.

K/A: 286000.A4.01 (3.3/3.2)

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

P-1: RESTART RPS-MG-1 AND CLOSE EPA BREAKERS

Description:

Direction is provided to restart the RPS Motor Generator (RPS-MG-1) which supplies power to RPS Bus A using SOP-RPS-START. During (A)(D)(R) 6 the start the expected voltage indication is not present requiring manual reset of the MG overvoltage trip.

K/A: 212000.A2.01 (3.7/3.9)

P-2: CHOOSE METHOD - INSERT CONTROL RODS BY VENTING SCRAM AIR HEADER

Description:

Based on initial conditions provided, recognize that manually (D)(E)(R) 1 venting the scram air header is the next action to take in an attempt to insert control rods.

K/A: 295037.EA1.05 (3.9/4.0)

P-3: (RSD) REMOTE SHUTDOWN PANEL ACTIVATION DURING A CONTROL ROOM EVACUATION (Time Critical)**

Description:

Based on a Main Control Room evacuation due to fire, and from (D)(E)(R) 7 a designated starting point, transit to the Remote Shutdown Panel and activate panel within required time using ABN-CR-EVAC Attachment 7.2.

K/A: 295016 AA1.07 (4.2/4.3) ** Ref: OI-69, TCOA-3/TCOA-4

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 (5)

(C)ontrol room (D)irect from bank 9 (8)

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

(EN)gineered safety feature 1 (2) (control room system)

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

(N)ew or (M)odified from bank including 1(A) 2 (3)

(P)revious 2 exams 3 (3) (randomly selected)

(R)CA 1 (3)

(S)imulator Attachment 9 Page 3 of 6

ES-301 Control Room/In-Plant Systems Outline (Rev 1 - 12/10/16) Form ES-301-2 Facility: Columbia Generating Station Date of Examination: 2/27/17 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S-1: TRANSFER BUS SM-3 FROM TR-S TO TR-N

Description:

Transfer 4160 VAC Bus SM-3 from the Startup Transformer to (D)(S) 6 the Normal Transformer.

K/A: 262001.A4.04 (3.6/3.7)

S-2: (RHR) RESPOND TO LOSS OF SHUTDOWN COOLING

Description:

Restore Residual Heat Removal (RHR) Loop B shutdown cooling per SOP-RHR-SDC (RHR Loop B Shutdown Cooling Quick Restart). (A)(L)(N)(S) 4 Following pump start, insufficient pump head (and consequently flow) will necessitate securing pump per CAUTION in procedure.

K/A: 205000.A2.06 (3.4/3.5)

S-3: (HPCS) HPCS SYSTEM INITIATION

Description:

Initiate High Pressure Core Spray (HPCS) system per SOP-HPCS-INJECTION and restore RPV level back to directed band. Following (A)(N)(EN) start of the HPCS pump its minimum flow valve will fail to automatically close 2 (L)(S) once RPV injection has occurred (resulting in a lower injection rate into the RPV). Valve must be manually closed to maximize injection.

K/A: 209002.A4.04 (3.1/3.1)

S-4: (CR HVAC) INITIATE CR HVAC MANUAL PRESSURIZATION MODE

Description:

Place both trains of Control Room Ventilation in the Manual Pressurization Mode of operation per SOP-HVAC/CR-OPS (inlet damper for (A)(D)(EN) 9 one of the Control Room Emergency Filter Units fail to auto open and must be (S) opened manually).

K/A: 290003.A4.01 (3.2/3.2)

S-5: (RB HVAC) RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC

Description:

Restart Reactor Building (RB) HVAC using RB Outside Air Fan (D)(P)(S) 5 1A and RB Exhaust Air Fan 1A per SOP-HVAC RB-RESTART-QC to re-establish Secondary Containment integrity.

K/A: 290001.A4.01 (3.3/3.4)

S-6: (DEH) LOWER RPV PRESSURE USING DEH

Description:

Recognize that auto control of bypass valves to lower RPV pressure to a target of 550 psig does not work and that the manual lowering of (A)(D)(L)(P) 3 RPV pressure at a rate LE 50 psig per minute through manual control of (S)

Bypass Valves would be required.

K/A: 241000.A4.02 (4.1/4.1)

S-8: (Fire Protection) ABN-FIRE Immediate Actions

Description:

Given the report of heavy smoke in the field and the observance (N)(S) 8 of several fire alarms, perform the Immediate Actions of ABN-FIRE.

K/A: 286000.A4.01 (3.3/3.2)

Attachment 9 Page 3 of 6

ES-301 Control Room/In-Plant Systems Outline (Rev 1 - 12/10/16) Form ES-301-2 In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

P-1: RESTART RPS-MG-1 AND CLOSE EPA BREAKERS

Description:

Direction is provided to restart the RPS Motor Generator (RPS-MG-1) which supplies power to RPS Bus A using SOP-RPS-START. During (A)(D)(R) 6 the start the expected voltage indication is not present requiring manual reset of the MG overvoltage trip.

K/A: 212000.A2.01 (3.7/3.9)

P-2: CHOOSE METHOD - INSERT CONTROL RODS BY VENTING SCRAM AIR HEADER

Description:

Based on initial conditions provided, recognize that manually (D)(E)(R) 1 venting the scram air header is the next action to take in an attempt to insert control rods.

K/A: 295037.EA1.05 (3.9/4.0)

P-3: (RSD) REMOTE SHUTDOWN PANEL ACTIVATION DURING A CONTROL ROOM EVACUATION (Time Critical)**

Description:

Based on a Main Control Room evacuation due to fire, and from (D)(E)(R) 7 a designated starting point, transit to the Remote Shutdown Panel and activate panel within required time using ABN-CR-EVAC Attachment 7.2.

K/A: 295016 AA1.07 (4.2/4.3) ** Ref: OI-69, TCOA-3/TCOA-4

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 (5)

(C)ontrol room (D)irect from bank 8 (7)

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

(EN)gineered safety feature 1 (2) (control room system)

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

(N)ew or (M)odified from bank including 1(A) 2 (3)

(P)revious 2 exams 3 (2) (randomly selected)

(R)CA 1 (3)

(S)imulator Attachment 9 Page 5 of 6

ES-301 Control Room/In-Plant Systems Outline (Rev 1 - 12/10/16) Form ES-301-2 Facility: Columbia Generating Station Date of Examination: 2/27/17 Exam Level: RO SRO-I SRO-U Operating Test No.: 1 Control Room Systems:* 8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System / JPM Title Type Code* Safety Function S-1: TRANSFER BUS SM-3 FROM TR-S TO TR-N

Description:

Transfer 4160 VAC Bus SM-3 from the Startup Transformer to (D)(S) 6 the Normal Transformer.

K/A: 262001.A4.04 (3.6/3.7)

S-2: (RHR) RESPOND TO LOSS OF SHUTDOWN COOLING

Description:

Restore Residual Heat Removal (RHR) Loop B shutdown cooling per SOP-RHR-SDC (RHR Loop B Shutdown Cooling Quick Restart). (A)(L)(N)(S) 4 Following pump start, insufficient pump head (and consequently flow) will necessitate securing pump per CAUTION in procedure.

K/A: 205000.A2.06 (3.4/3.5)

S-3: (HPCS) HPCS SYSTEM INITIATION

Description:

Initiate High Pressure Core Spray (HPCS) system per SOP-HPCS-INJECTION and restore RPV level back to directed band. Following (A)(N)(EN) start of the HPCS pump its minimum flow valve will fail to automatically close 2 (L)(S) once RPV injection has occurred (resulting in a lower injection rate into the RPV). Valve must be manually closed to maximize injection.

K/A: 209002.A4.04 (3.1/3.1)

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

P-2: CHOOSE METHOD - INSERT CONTROL RODS BY VENTING SCRAM AIR HEADER

Description:

Based on initial conditions provided, recognize that manually (D)(E)(R) 1 venting the scram air header is the next action to take in an attempt to insert control rods.

K/A: 295037.EA1.05 (3.9/4.0)

P-3: (RSD) REMOTE SHUTDOWN PANEL ACTIVATION DURING A CONTROL ROOM EVACUATION (Time Critical)**

Description:

Based on a Main Control Room evacuation due to fire, and from (D)(E)(R) 7 a designated starting point, transit to the Remote Shutdown Panel and activate panel within required time using ABN-CR-EVAC Attachment 7.2.

K/A: 295016 AA1.07 (4.2/4.3) ** Ref: OI-69, TCOA-3/TCOA-4

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.

Attachment 9 Page 5 of 6

ES-301 Control Room/In-Plant Systems Outline (Rev 1 - 12/10/16) Form ES-301-2 Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 2-3 (2)

(C)ontrol room (D)irect from bank 4 (3)

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

(EN)gineered safety feature 1 (1) (control room system)

(L)ow-Power / Shutdown 1 (2)

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

(P)revious 2 exams 2 (0)

(R)CA 1 (2)

(S)imulator Attachment 9 Page 2 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE RESTART OF RPS-MG-1 AND REPOWER RPS BUS (Plant) (Alt Path)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001641 Rev. No. 4 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 6/10/08 REVISED BY John Miller DATE 9/14/16 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

None Special Setup Instructions:

None JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

The student is given SOP-RPS-START Sections 5.1 and 5.3.

Tools/Equipment: None Safety Items: None Task Number: RO-0248 Validation Time: 12 Minutes Alternate Path: Yes Time Critical: No PPM

Reference:

SOP-RPS-START Section 5.1 and 5.3 Rev. 5 Location: Plant NUREG 1123 Ref: 212000A2.01 (3.7/3.9) Performance Method: Simulate Task Standard: RPS-MG-1 is running and the A RPS EPA breakers have been closed.

LO001641 Rev. 4 Page 2 of 12

JPM CHECKLIST INITIAL RPS Division A has been de-energized due to a fault. The fault has been identified and corrected.

CONDITIONS:

INITIATING The CRS directs you to restart RPS-MG-1 and repower the Division A RPS bus in accordance with SOP-RPS-START section 5.1 CUE: and 5.3. Inform the CRS when the RPS EPA breakers have been closed. The performance of this JPM is simulated. Control manipulations will not be performed.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 1 Step 5.1.1 Observes RPS-DISC-7A1Bs handle. The handle is pointing to On.

S/U Verify RPS-DISC-7A1B is Closed (RPS Bus Mtr Gen MG-1 Supply Breaker) (E-MC-7A).

2 Step 5.1.2 Performs this step.

S/U Perform the following at E-CP-C72/S001A (RPS-MG-1 Control Panel):

3 Step 5.1.2a Observes the Green Motor Off The Green light is on.

indicating light is on. S/U Verify Motor Off indicating light illuminated (Green).

4 Step 5.1.2b Observes RPS-CB-MG1 is open with Indicate that the lever is pointed lever in Off position. downward (towards off). S/U Verify RPS-CB-MG1 Open (Generator Output Breaker).

LO001641 Rev. 4 Page 3 of 12

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 5 Step 5.1.2c Simulates depressing and holding The start pushbutton is RPS-RMS-MG1/START, MOTOR depressed. S/U*

Depress and Hold RPS-RMS-MG1/ START, ON pushbutton depressed.

pushbutton (Motor On). Red light On, Green light Off The motor starts to make noise and starts to spin.

6 Step 5.1.2d Observes the Green Motor Off The Green light is off and the indicating light off and the Red Red light is on. S/U Verify the following:

Motor On indicating light is on.

Motor Off indicating light extinguished (Green)

Motor On light illuminates (Red) 7 Step 5.1.2e Simulates releasing the MOTOR ON (If the MG set is actually pushbutton (when cue is read). running in the plant) The motor S/U*

When RPS-MG-1 has come up to speed, then release RPS-RMS- MG1/START is now spinning as you see it and making the noise you now pushbutton.

hear it; OR (If the MG set is not actually running in the plant) The MG motor speed is not visible rising and motor noise is constant.

NOTE: Motor On pushbutton doubles as an Over Voltage Trip Reset pushbutton.

LO001641 Rev. 4 Page 4 of 12

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 8 Step 5.1.2f Verbalizes that voltage indication Indicate zero volts on RPS-S/U*

If voltage is not indicated at rated speed, then would be expected on RPS-VM- VM-MG1A.

MG1A.

momentarily depress RPS-RMS-MG1/

START, Motor On pushbutton to reset the overvoltage trip.

Alt Path 9 Recognizes need to momentarily depress the Simulates momentarily depressing The pushbutton has been Step Motor On pushbutton. the RPS-RMS-MG1/START, Motor depressed and released. S/U*

On pushbutton. Indicate 120 volts on RPS-VM-MG1A.

10 Step 5.1.2g Observes voltage on Continue to indicate 120 volts RPS-VM-MG1A. on RPS-VM-MG1A. S/U Verify RPS-VM-MG1A voltage stabilizes at about 120 VAC.

11 5.1.2h Simulates closing RPS-CB-MG1 by (If RPS-CB-MG1 is closed) pushing up on lever to On. The lever is as you see it. S/U*

Close RPS-CB-MG1.

OR (If RPS-CB-MG1 is not closed)

Indicate that the lever is pointed up (towards On).

12 Step 5.1.3 Performs section 5.3 as follows:

S/U Proceed to Section 5.3.

13 Step 5.3.1 Recognizes Section 5.1 was just completed. S/U Verify Section 5.1 completed.

LO001641 Rev. 4 Page 5 of 12

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 14 Step 5.3.2 Verbalizes where the EPA breaker The student does not have to go keys # 166 & 168 are located to the control room to obtain S/U*

Obtain required EPA breaker keys from the (Control Room in key locker outside keys - an explanation on where Control Room key locker: Shift Managers office). the keys are is sufficient.

Key 166 (RPS C72-S003-A Div A Test)

(RPS-EPA-3A) You have obtained the keys.

Key 168, (RPS C72-S003-B Div A Test)

(RPS-EPA-3B) 15 Step 5.3.3 Performs this step.

S/U Close RPS-EPA-3A as follows (EPA Breaker) (RPS-MG2 Room):

16 Step 5.3.3a Observes S-1 is in Normal position. The keylock switch is pointed S/U to Normal.

Verify breaker keylock switch S-1 in Normal.

17 Step 5.3.3b Observes S-2 is in OPER position. The keylock switch is pointed S/U to OPER.

Verify breaker keylock switch S-2 in OPER.

18 Step 5.3 3c Observes Power In Indicator is on. The Red, RPS-MG-1 Power In S/U light is on.

Verify the Power In indicator illuminated.

LO001641 Rev. 4 Page 6 of 12

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 19 Step 5.3.3d Observes the following indicators: All indicator lights are off.

S/U If any of the following indicators are

Overvoltage illuminated, then rotate keylock switch S-2

Undervoltage to RESET, and return to OPER

Underfrequency

Overvoltage

Power Out

Undervoltage

Underfrequency

Power Out 20 Step 5.3.3e Verifies the following indicators are All indicator lights are off.

S/U extinguished:

Verify the following indicators extinguished:

Overvoltage

Overvoltage

Undervoltage

Undervoltage

Underfrequency

Underfrequency

Power Out

Power Out 21 Step 5.3.3f Simulates opening EPA breaker RPS- Indicate that the breakers lever S/U*

EPA-3A by pushing down on lever is in the lowered position Open RPS-EPA-3A to reset it. towards OFF position. (towards Off).

22 Step 5.3.3g Simulates closing EPA breaker RPS- Indicate that the breakers lever S/U*

EPA-3B by pushing up on lever is in the raised position Close RPS-EPA-3A. towards ON position. (towards On).

23 Step 5.3.3h Observes the Power Out indicator is The Red RPS-EPA-3C Power S/U on. Out light is on.

Verify the Power Out indicator illuminated.

NOTE: EPA breakers are designed such that the undervoltage lights for RPS-EPA breakers may illuminate indicating an undervoltage condition without activating the undervoltage trip circuit.

LO001641 Rev. 4 Page 7 of 12

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 24 Step 5.3.4 Verifies under voltage light is not on The under voltage light is off.

S/U with breaker closed.

If the under voltage light is illuminated and the breaker is closed, then initiate a work request.

25 Step 5.3.5 Performs this step. S/U Close RPS-EPA-3C as follows (EPA Breaker) (RPS-MG2 Room).

26 Step 5.3.5a Observes S-1 is in Normal position. The keylock switch is pointed S/U to Normal.

Verify breaker keylock switch S-1 in the NORMAL.

27 Step 5.3.5b Observes S-2 is in OPER position. The keylock switch is pointed S/U to OPER.

Verify breaker keylock switch S-2 in the OPER.

28 Step 5.3.5c Observes Power In Indicator is on. The Power In light is lit.

S/U Verify the Power In indicator illuminated.

Alt Path 29 Step 5.3.5d Observes the following indicators: All indicators are off except the Step S/U*

white under frequency light is If any of the following indicators are not

Overvoltage on.

extinguished, then rotate keylock switch S-2

Undervoltage to the RESET position, and return to OPER

Underfrequency

Power Out

Overvoltage

Undervoltage

Underfrequency

Power Out LO001641 Rev. 4 Page 8 of 12

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step Alt Path 30 Recognizes need to reset indicators. Simulates rotating Keylock switch S- The switch has been turned to Step S/U*

2 to RESET and back to OPER. RESET and back to OPER.

31 Step 5.3.5e Verifies the following indicators are All indicator lights are off.

S/U extinguished:

Verify the following indicators extinguished:

Overvoltage

Overvoltage

Undervoltage

Undervoltage

Underfrequency

Underfrequency

Power Out

Power Out 32 Step 5.3.5f Simulates opening EPA breaker RPS- Indicate that the breakers lever S/U*

EPA-3C by pushing down on lever is in the lowered position Open RPS-EPA-3C to reset it. towards OFF position. (towards Off).

33 Step 5.3.5g Simulates closing EPA breaker RPS- Indicate that the breakers lever S/U*

EPA-3C by pushing up on lever is in the raised position Close RPS-EPA-3C. towards ON position. (towards On).

34 Step 5.3.5h Observes the Power Out indicator is The Power Out light is on.

S/U on.

Verify the Power Out indicator illuminated.

NOTE: EPA breakers are designed such that the undervoltage lights for RPS-EPA breakers may illuminate indicating an undervoltage condition without activating the undervoltage trip circuit.

35 Step 5.3.5i Verifies under voltage light is off The under voltage light is off.

S/U with the breaker closed.

If the under voltage light is illuminated and the breaker is closed, then initiate a work request.

Termination Criteria: Student informs the CRS that RPS-EPA-3A and RPS-EPA-3C breakers are closed.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001641 Rev. 4 Page 9 of 12

LO001641 Rev. 4 Page 10 of 12

RESULTS OF JPM RESTART RPS MG-1 AND REPOWER THE RPS BUS Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: RPS-MG-1 is running and the A RPS EPA Breakers have been closed.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001641 Rev. 4 Page 11 of 12

STUDENT JPM INFORMATION CARD Initial Conditions:

RPS Division A has been de-energized due to a fault The fault has been identified and corrected Initiating Cue:

The CRS directs you to restart RPS-MG-1 and repower the Division A RPS bus in accordance with SOP-RPS-START section 5.1 and 5.3.

Inform the CRS when the RPS EPA breakers have been closed.

THE PERFORMANCE OF THIS JPM IS SIMULATED.

CONTROL MANIPULATIONS WILL NOT BE PERFORMED.

Page 12 of 12

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE INSERT CONTROL RODS BY VENTING SCRAM AIR HEADER (Plant)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001593 Rev. No. 1 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 5/11/06 REVISED BY Ron Hayden DATE 5/01/11 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

None JPM Instructions:

The student is handed the Student Information JPM Card and PPM 5.5.11 Tab D.

Tools/Equipment: Pre staged EOP Tools Safety Items: Hard Hat; Safety Glasses; Gloves Task Number: RO-0680 Validation Time: 9 Minutes Alternate Path: No Time Critical: No PPM

Reference:

PPM 5.5.11 Rev. 8 Location: Plant NUREG 1123 Ref: 295037 EA1.05 (3.9/4.0) Performance Method: Simulate Task Standard: The scram air header has been vented and PPM 5.5.11 Tab D has been completed.

LO001593 Rev. 1 Page 2 of 6

JPM CHECKLIST INITIAL A scram has been initiated and the blue scram lights are extinguished at H13-P603. Reactor pressure is stable at 930 psig and CONDITIONS: Reactor Power is 38%.

INITIATING The CRS has directed you to insert control rods by venting the Scram Air Header per PPM 5.5.11 Tab D. Inform the CRS when CUE: Tab D has been completed. The performance of this JPM will be simulated. Control manipulations will not be performed.

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 1 Close CRD-V-95, Scram Air Header Simulates turning the handwheel for Handwheel stops moving S/U*

Isolation CRD-V-95 clockwise to close valve 2 Close CRD-V-729, CRD-PI-13 Simulates turning the handwheel for Handwheel stops moving S/U*

Isolation CRD-V-729 clockwise to close valve Note/Caution: Pressurized air will be released when drain plug is removed from CRD-PI-13 which could cause personnel injury 3 Remove instrument drain plug for Using pre-staged wrench, simulates Drain plug is in hand S/U*

CRD-PI-13 rotating the instrument drain plug counterclockwise on CRD-PI-13 until drain plug is removed 4 Open CRD-V-729, CRD-PI-13 Simulates turning the handwheel for Air can be heard/felt coming from S/U*

isolation CRD-V-729 counter-clockwise to drain line open the valve Handwheel stops moving LO001593 Rev. 1 Page 3 of 6

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 5 When scram air header is fully Verifies CRD-PI-13 indicates the air Air can no longer be heard/felt S/U depressurized and no further rod header is depressurized coming from drain line motion observed Indicate 0 psig on the gauge face Contacts the Control Room to verify Inform the candidate that no further the status of rod motion. rod motion is observed 6 Restore system alignment as follows:

7 Close CRD-V-729 Simulates turning the handwheel for Handwheel stops moving S/U*

CRD-V-729 clockwise until valve is closed 8 Install instrument drain plug for Simulates inserting the drain plug Plug is connected to the pipe and has S/U*

CRD-PI-13 back into the pipe and simulates stopped turning turning the drain plug for CRD-PI-13 clockwise to reinstall it 9 Open CRD-V-729 Simulates turning the handwheel Handwheel stops moving S/U*

CRD-V-729 counter-clockwise until valve is opened 10 Open CRD-V-95 Simulates turning the handwheel for Handwheel stops moving S/U*

CRD-V-95 counter-clockwise until valve is opened Termination Criteria: Student informs the CRS that actions to vent Scram Air Header have been completed.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001593 Rev. 1 Page 4 of 6

RESULTS OF JPM INSERT CONTROL ROD BY VENTING SCRAM AIR HEADER Examinee (Print): __________________________________________________________________

Evaluator (Print): __________________________________________________________________

Task Standard: The scram air header has been vented and PPM 5.5.11 Tab D has been completed.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001593 Rev. 1 Page 5 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

A scram has been initiated and the blue scram lights are extinguished at H13-P603.

Reactor pressure is stable at 930 psig and Reactor Power is 38%.

Initiating Cue:

The CRS has directed you to insert control rods by venting the Scram Air Header per PPM 5.5.11 Tab D.

Inform the CRS when actions for Tab D have been completed THE PERFORMANCE OF THIS JPM IS SIMULATED CONTROL MANIPULATIONS WILL NOT BE PERFORMED Page 6 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE REMOTE SHUTDOWN PANEL ACTIVATION DURING A CONTROL LESSON TITLE ROOM EVACUATION (Plant) (Time Critical)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LR001767 Rev. No. 4 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 8/24/06 REVISED BY Steve Bruce DATE 11/15/16 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

N/A Special Setup Instructions:

This JPM should be started from the outside of the exit door from the Control Room. The student should be handed the JPM Information Card and the examiner will read the initial conditions and cue to the student.

It expected that the student will go to the Remote Shutdown Room to get a copy of ABN-CR-EVAC. When the procedure is identified in the RSD Room, then hand the student a copy of Attachment 7.2.

JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: None Safety Items: Hard Hat, Safety Glasses Task Number: RO-1057, SRO-0251 Validation Time: 9 Minutes Alternate Path: No Time Critical: YES - 10 Minutes PPM

Reference:

ABN-CR-EVAC Rev. 35 Location: Plant NUREG 1123 Ref: 295016AK2.01 (4.4/4.5) Performance Method: Simulate Task Standard: Remote Shutdown Panel Activation has been accomplished within the time requirement.

LR001767 Rev. 4 Page 2 of 7

JPM CHECKLIST INITIAL The Control Room has been evacuated due to fire. The immediate and subsequent operator actions of PPM ABN-CR-EVAC have CONDITIONS: been completed.

INITIATING The CRS has directed you to perform Attachment 7.2 to activate the Remote Shutdown Panel. The performance of this JPM will be CUE: simulated. No control manipulations will be performed. This is a time critical JPM and your time starts now.

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat WHEN THE STUDENT HAS ENTERED THE RSD ROOM AND HAS A COPY OF ABN-CR-EVAC IN HAND, GIVE THE STUDENT HIS COPY OF ATTACHMENT 7.2.

Note: The RSD panel must be activated within 10 minutes from the time the Shift Manager (or designee) orders a reactor scram due to a design basis fire.

CAUTION: Failure to transfer RCIC flow control to EMERG may cause RCIC to trip when DP-S1-1A feeder is tripped in the subsequent step.

1 Step 7.2.1 Simulates placing RCIC-RMS- The switches arrow is pointing to RSTS7 (transfer switch 1) in the EMERG S/U*

Place RCIC-RMS-RSTS7 in EMERG position EMERG (RCIC FLOW CONTROL RCIC-FIC-1R POWER TRANSFER) (C61-P001, RSD).

Note: De-energizing DP-S1-1A will defeat the automatic ADS function from Division 1.

LR001767 Rev. 4 Page 3 of 7

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 2 Step 7.2.2 Simulates opening breakers on DP- As each breaker is opened: The S1/1: handle is pointing to the OFF S/U*

Verify open the following breakers

IN-3A feeder (Cubicle 2B - if ON position on DP-S1/1 within 15 minutes (Battery Charger Room 1): - simulates turning handle CW to OFF position)

E-DISC-DPS11/2B (IN-3A feeder)

DP-S1-1A feeder (Cubicle 2C - if

E-DISC-DPS11/2C (DP-S1-1A ON - simulates turning handle CCW When step 7.2.2 is completed, feeder) to OFF position) inform the student that the time critical portion of the JPM has been

E-DISC-DPS11/2D (IN-3B

IN-3B feeder (Cubicle 2D - if completed.

feeder) ON - simulates turning handle CW to OFF position) 3 Step 7.2.3 In ARSD Room, simulates placing As each switch is turned: The the following power transfer switches arrow is pointing to S/U*

Place the following four (4) power switches to the EMERG position: EMERG transfer switches to EMERG (E-CP-ARS, ARSD):

41

47

41

48

47

59

48

59 4 Step 7.2.4 In RSD Room, simulates placing the As each switch is turned: The following power transfer switches to switches arrow is pointing to S/U*

Place all five (5) FRTS power transfer switches to EMERG (E-CP- EMERG: EMERG FRTP, RSD):

31

31

32

32

33

33

34

34

35

35 LR001767 Rev. 4 Page 4 of 7

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 5 Step 7.2.5 Simulates placing the following As each switch is turned: The power transfer switches to EMERG: switches arrow is pointing to S/U*

Place the following twelve (12)

2 and 5 EMERG power transfer switches to EMERG (EC61-P001, RSD):

6 and 7

8 and 11

2 and 5

12 and 13

6 and 7

15 and 16

8 and 11

17 and 18

12 and 13

15 and 16

17 and 18 6 Step 7.2.6 Simulates placing the following As each switch is turned: The power transfer switches to EMERG: switches arrow is pointing to S/U*

Place the following four (4) power

21 EMERG transfer switches to EMERG (H22-P100, RSD):

22

23

21

24

22

23

24 7 Notify the CRS that Attachment 7.2 Notifies CRS that attachment 7.2 is Inform the student that the JPM is S /U is complete. complete. complete.

Termination Criteria: Student informs CRS that Attachment 7.2 is complete.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LR001767 Rev. 4 Page 5 of 7

RESULTS OF JPM REMOTE SHUTDOWN PANEL ACTIVATION DURING A CONTROL ROOM EVACUATION Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: Remote Shutdown Panel Activation has been accomplished within the time requirement.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LR001767 Rev. 4 Page 6 of 7

STUDENT JPM INFORMATION CARD Initial Conditions:

The Control Room has been evacuated due to fire.

The immediate and subsequent operator actions of ABN-CR-EVAC have been completed.

Initiating Cue:

The CRS has directed you to perform Attachment 7.2 to activate the Remote Shutdown Panel.

THE PERFORMANCE OF THIS JPM WILL BE SIMULATED.

CONTROL MANIPULATIONS WILL NOT BE PERFORMED.

THIS IS A TIME CRITICAL JPM.

Page 7 of 7

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE TRANSFER SM-3 FROM TR-S TO TR-N (SIMULATOR)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LR001898 Rev. No. 1 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 09/04/08 REVISED BY Dave Crawford DATE 12/21/16 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

Any IC where SM-1 is being powered from N1-1 Special Setup Instructions:

None JPM Instructions:

Verify the current procedure against the JPM. If the procedure is a different revision than listed in the JPM, ensure the critical steps still match. If the critical steps have changed, the JPM should be revised.

The evaluator and student shall use current procedure. The evaluator should mark off steps as they are completed, note comments, and transfer the comments to the Results of JPM page.

Tools/Equipment: None Safety Items: None Task Number: R0-0414 Validation Time: 5 minutes Alternate Path: No Time Critical: No PPM

Reference:

SOP-ELEC-4160V-OPS Section 5.3 Rev. 12 Location: SIMULATOR NUREG 1123 Ref: 262001A4.04 (3.6/3.7) Performance Method: PERFORM Task Standard: SM-3 has been transferred from the Startup Transformer to the Normal Transformer per SOP-ELEC-4160V-OPS.

LR001898 Rev. 1 Page 2 of 7

JPM CHECKLIST INITIAL SM-3 is currently powered from the Startup Transformer. Work on breaker CB-N1/3 has been completed and CB-N1/3 has been CONDITIONS: racked in. All conditions, limitations, and prerequisites for this evolution are completed.

INITIATING The CRS has directed you to transfer SM-3 from the Startup Transformer to the Normal Transformer. Inform the CRS when SM-3 CUE: is being powered from the Normal Transformer.

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat NOTE: The control and indications necessary to perform this section are located at H13-P800 (Bd C).

1 Step 5.3.1 Observes CB-N1/3 white Lockout S/U Circuit Avail light is illuminated Verify CB-N1/3 white Lockout Circuit Avail light is illuminated 2 Step 5.3.2 Observes CB-N1/3 green light S/U illuminated and green flag displayed Verify CB-N1/3 green light illuminated and green flag displayed 3 Step 5.3.3 Observes CB-S3 white Lockout S/U Circuit Avail light illuminated Verify CB-S3 white Lockout Circuit Avail light illuminated 4 Step 5.3.4 Observes CB-S3 red light S/U illuminated Verify CB-S3 red light illuminated LR001898 Rev. 1 Page 3 of 7

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 5 Step 5.3.5 Places CB-N1/3 Sync Selector S/U*

switch to the MAN position Place CB-N1/3 Sync Selector switch in MAN 6 Step 5.3.6 Observes voltages present on both S/U incoming and running buses Verify voltages present on both incoming and running buses NOTE: The blue Sync Permit light for E-CB-N1/3 is illuminated from initiation of breaker closure until closure actually occurs.

NOTE: E-CB-S3 should automatically trip when E-CB-N1/3 closes.

NOTE: H13-800.C3.3-3, BKR S3 TRIP will alarm when the following step is performed.

7 Step 5.3.7 Places CB-N1/3 control switch to S/U*

close by turning to the right Close CB-N1/3 8 Step 5.3.8 Observes CB-S3 green light S/U illuminated and red light out Verify CB-S3 auto trips 9 Step 5.3.9 Places CB-S3 control switch in TRIP S/U by turning to the left Place CB-S3 control switch in TRIP 10 Step 5.3.10 Observes S3 green light illuminated S/U and green flag displayed Verify CB-S3 green light illuminated and green flag displayed LR001898 Rev. 1 Page 4 of 7

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 11 Step 5.3.11 Places CB-N1/3 Sync Selector S/U switch to the OFF position Place CB-N1/3 Sync Selector switch in OFF Termination Criteria: Student informs CRS that SM-3 is being powered by the Normal Transformer.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LR001898 Rev. 1 Page 5 of 7

RESULTS OF JPM TRANSFER SM-3 FROM TR-S TO TR-N Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: SM-3 has been transferred from the Startup Transformer to the Normal Transformer per SOP-ELEC-4160V-OPS.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LR001898 Rev. 1 Page 6 of 7

STUDENT JPM INFORMATION CARD Initial Conditions:

SM-3 is currently powered from the Startup Transformer.

Work on breaker CB-N1/3 has been completed and CB-N1/3 has been racked in.

All conditions, limitations, and prerequisites for this evolution are completed.

Initiating Cue:

The CRS has directed you to transfer SM-3 from the Startup Transformer to the Normal Transformer.

Inform the CRS when SM-3 is being powered from the Normal Transformer.

Page 7 of 7

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE RESPOND TO A LOSS OF SHUTDOWN COOLING LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001862 Rev. No. 1 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/29/16 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

RESPOND TO A LOSS OF SHUTDOWN COOLING MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

INITIALIZE to IC-210 - OR -

Manually place simulator in MODE 4 with RHR-P-2B lined-up for SDC with SW-P-1B operating. Insert inadvertent closure of RHR-V-8 and RHR-V-9. Perform steps to secure RHR Loop B Shutdown Cooling Lineup using SOP-RHR-SDC (Section 5.5).

Special Setup Instructions:

Place Danger Tag on RHR-V-64B (RHR B Min Flow valve control switch)

Have copy of SOP-RHR-SDC, section 5.7, available to student with previously performed steps marked complete (all of section 3 and 4 initialed (or N/Ad and initialed), as appropriate for plant conditions and steps 5.7.1 & 5.7.2 initialed with step 5.7.3 marked N/A and initialed.)

Clear BISIs on H13-P601 after simulator placed in RUN JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: None Safety Items: None Task Number: RO-1300 Validation Time: 25 Minutes Alternate Path: Yes Time Critical: No PPM

Reference:

SOP-RHR-SDC rev 25 Location: Simulator NUREG 1123 Ref: 205000.A2.06 (3.4 / 3.5) Performance Method: Perform Task Standard: Perform RHR Loop B Shutdown Cooling Quick Restart in accordance with SOP-RHR-SDC Section 5.7. When RHR Pump B is started, recognize that RHR-P-2B fails to develop sufficient head following pump start, and secure RHR-P-2B per the procedural caution on page 48 of SOP-RHR-SDC.

LO001862 Rev. 1 Page 2 of 10

RESPOND TO A LOSS OF SHUTDOWN COOLING JPM CHECKLIST INITIAL Given the following:

CONDITIONS:

The reactor is in Mode 4.

RHR Loop B was in Shutdown Cooling when RHR-V-8 and RHR-V-9 were inadvertently closed, tripping RHR-P-2B.

Shutdown Cooling has been off for 30 minutes.

No activities have occurred that could cause the formation of voids in RHR-B.

RHR Loop B Shutdown Cooling lineup has been secured per SOP-RHR-SDC, Section 5.5.

SW-P-1B is operating per SOP-SW-START.

OSP-RCS-C102 (Cooldown Rate Log) is secured.

INITIATING The CRS directs you to restore RHR Loop B shutdown cooling per SOP-RHR-SDC, section 5.7, RHR Loop B Shutdown Cooling CUE: Quick Restart. Steps 5.7.1 - 5.7.3 are complete; begin at step 5.7.4.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step CAUTION: To prevent failure of the RHR pumps due to excessive radiation exposure, alternate shutdown cooling, by Suppression Pool Cooling, is the only allowable mode for shutdown cooling once a degraded core condition has been identified.

CAUTION: Two Loop RHR Shutdown Cooling operations may cause actuation of Excess Flow Trip Isolation if ICP-RHR-Q901, RHR SDC Mode High Flow Isolation - CFT/CC, has not been completed within its required surveillance interval.

CAUTION: Failure to warm the RHR pump suction line may cause excessive thermal stress on the RHR injection line/Recirculation piping tee.

NOTE: This section is used if the Delta-T between RHR B Heat Exchanger Outlet and RRC-P-1A Suction (RRC-TR-650, pt. 1, or TDAS pt. X292) is LT 80 degrees F.

NOTE: If normal Shutdown Cooling cannot be used, then refer to ABN-RHR-SDC-ALT.

NOTE: Technical Specifications require Reactor Vessel and head flange temperatures be maintained GT 80 degrees F when the Vessel head bolting studs are being tensioned. (SR 3.4.11.7)

NOTE: If core decay heat is present, or if system metal temperature is high, a recirculation pump of RHR pump (do not use a recirculation pump if an RHR pump is available) along with a means of determining Reactor water temperature should be kept in service.

Examiner Note: Candidate may bring up Shutdown Cooling screen on PDIS computer to validate system parameters.

LO001862 Rev. 1 Page 3 of 10

RESPOND TO A LOSS OF SHUTDOWN COOLING

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 1 VERIFY RHR-V-8 OPEN Observed RHR-V-8 CLOSED.

Repositioned RHR-V-8 Switch to S/U*

(RHR Shutdown Cooling Suction open. Observed that RHR-V-8 Red Outboard Isolation) (H13-P601).

Light is ON and Green Light is OFF.

2 VERIFY RHR-V-9 OPEN (RHR Observed RHR-V-9 CLOSED. S/U*

Shutdown Cooling Suction Inboard Repositioned RHR-V-9 Switch to Isol) (H13-P601). open. Observed that RHR-V-9 8 Red Light is ON and Green Light is OFF.

3 VERIFY RHR-V-4B CLOSED Observed RHR-V-4B Red Light is S/U (Pump Suction from Supp Pool) OFF and Green Light is ON.

(H13-P601).

4 VERIFY RHR-V-6B OPEN Observed RHR-V-6B Red Light is S/U (Shutdown Cooling Suction) (H13- ON and Green Light is OFF.

P601).

Examiner Note: The following step is N/A per initial cue provided to the candidate.

5 IF RHR-V-6B, RHR-V-8, or RHR- N/A per initial cue. S/U V-9 were closed, AND any activity occurred that could cause the formation of voids while they were closed, THEN FILL and VENT RHR-P-2B suction piping as follows:

6 VERIFY RRC-P-1B OFF per SOP- Observed RRC-P-1B Red RUN Light is OFF and Green STOP Light S/U RRC-SHUTDOWN.

is ON.

NOTE: IF CRD Recirculation Pump seal purge is in operation, do not close both RRC-V-67B and RRC-V-23B.

LO001862 Rev. 1 Page 4 of 10

RESPOND TO A LOSS OF SHUTDOWN COOLING

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 7 VERIFY one of the following Verified RRC-V-67B Red Light is CLOSED. N/A the other. OFF and Green Light is ON or RRC- S/U RRC-V-67B (preferred) (Recirc V-23B Red Light is OFF and Green Pump B Discharge) (H13-P602) Light is ON.

RRC-V-23B (Recirc Pump B Suction) (H13-P602)

Examiner Note: The following step is N/A.

8 IF unable to close either RRC-V-23B N/A S/U or RRC-V-67B, THEN REFER to ABN-RHR-SDC-ALT.

NOTE: IF the RRC pump is not in service, than an alternate RRC temperature as determined by the CRS/Shift Manager should be used.

Examiner Note: The following step is N/A per initial cue.

9 IF RHR-SDC-B has been off GT two N/A per initial cue. S/U hours, THEN VERIFY the T between RHR B Heat Exchanger Outlet (RHR-TRS-601 or TDAS pt. X059) and RRC-P-1A Suction (RRC-TR-650, pt. 1, or TDAS pt.

X292) is LT 80° F.

RRC pump suction temperature.

______F RHR B Heat Exchanger (HX) Outlet temperature. ______F T between suction and HX Outlet temperatures. ______

10 VERIFY RHR-V-3B is CLOSED Verified RHR-V-3B Red Light is S/U (RHR-HX-1B Isolation) (H13-P601). OFF and Green Light is ON.

LO001862 Rev. 1 Page 5 of 10

RESPOND TO A LOSS OF SHUTDOWN COOLING

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 11 CLOSE or VERIFY CLOSED If open, repositioned RHR-V-48B S/U RHR-V-48B (RHR-HX-1B Shell control switch to closed and/or Side Bypass) (H13-P601). verified RHR-V-48B Red Light is OFF and Green Light is ON.

NOTE: The following steps are designed to minimize the potential of a water hammer when RHR-P-2B is started.

Examiner Note: Power is available to RHR-V-53B.

12 IF power is available to RHR-V-53B, Verified either RHR-V-53B Red or S/U*

THEN THROTTLE OPEN RHR- Green Light is ON. Repositioned V-48B approximately 8 seconds. and held RHR-V-48B to open for approximately 8 seconds and released. Verified both RHR-V-48B Red and Green Lights are ON.

Examiner Note: The following step is N/A. OSP-RCS-C102 secured in MODE 4.

13 IF required, N/A S/U THEN LOG RPV cooldown rate per OSP-RCS-C102.

Examiner Note: The following step is N/A. No refueling activities in progress.

14 IF starting the RHR pump during N/A S/U refuel activities, THEN NOTIFY the Refuel Supervisor RPV water clarity may be reduced.

15 VERIFY SW-P-1B operating per Verified SW-P-1B Red Light is ON S/U SOP-SW-START. and Green Light is OFF and flow is indicated in normal band.

CAUTION: Exceeding a flow of 8000 gpm or failure to maintain 800 gpm may cause RHR pump damage/failure.

NOTE: Attachment 6.1 shows RHR recommended operating conditions.

NOTE: With RHR B PUMP DISCH PRESS HIGH/LOW in alarm, RHR-P-2B may be started for EOP related activities.

Examiner Note: Two-handed operation is allowed.

LO001862 Rev. 1 Page 6 of 10

RESPOND TO A LOSS OF SHUTDOWN COOLING

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 16 START RHR-P-2B. Rotated RHR-P-2B control-switch to S/U*

start. Verified Red Light is ON and Green Light is OFF and discharge pressure increased.

Examiner Note: The student may secure RHR-P-2B any time following pump start, within 10 minutes of pump start.

Examiner Note: Power is available to RHR-V-53B. Two-handed operation is allowed.

Alt Path 17 IF power is available to RHR-V-53B, Verified either RHR-V-53B Red or S/U*

THEN IMMEDIATELY OPEN Green Light is ON. Rotated RHR-V-RHR-V-53B (Shutdown Cooling 53B control-switch to Open. Verified Return) (H13-P601). RHR-V-53B Red Light is ON and Green Light is OFF.

Examiner Note: The following step is N/A.

Alt Path 18 IF power is not available to RHR-V- N/A S/U 53B, AND RHR-V-53B has been left OPEN, THEN IMMEDIATELY THROTTLE OPEN RHR-V-48B approximately 8 seconds:

Alt Path 19 IF flow is not GE 3000 gpm, Verified flow LT 3000 gpm then S/U THEN THROTTLE RHR-V-48B to rotated and held RHR-V-48B establish approximately 3000 gpm. control-switch in Open in an attempt to raise flow (which is unsuccessful).

Informs CRS of the inability to raise flowrate as required.

Alt Path 20 After 30 seconds, THROTTLE May completely open RHR-V-48B in S/U RHR-V-48B to establish GT 5400 an attempt to establish flow (which is gpm, and preferably GT 7000 gpm. unsuccessful).

Alt Path 21 Secures RHR-P-2B. Rotated RHR-P-2B control-switch to S/U*

stop, within 10 minutes of pump start.

LO001862 Rev. 1 Page 7 of 10

RESPOND TO A LOSS OF SHUTDOWN COOLING

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step TERMINATING CUE: This JPM is Complete.

LO001862 Rev. 1 Page 8 of 10

RESULTS OF JPM RESPOND TO A LOSS OF SHUTDOWN COOLING Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: Perform RHR Loop B Shutdown Cooling Quick Restart in accordance with SOP-RHR-SDC Section 5.7. When RHR Pump B is started, recognize that RHR-P-2B fails to develop sufficient head following pump start, and secure RHR-P-2B per the procedural caution on page 48 of SOP-RHR-SDC.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001862 Rev. 1 Page 9 of 10

STUDENT JPM INFORMATION CARD Initial Conditions:

Given the following:

The reactor is in Mode 4.

RHR Loop B was in Shutdown Cooling when RHR-V-8 and RHR-V-9 were inadvertently closed, tripping RHR-P-2B.

Shutdown Cooling has been off for 30 minutes.

No activities have occurred that could cause the formation of voids in RHR-B.

RHR Loop B Shutdown Cooling lineup has been secured per SOP-RHR-SDC, Section 5.5.

SW-P-1B is operating per SOP-SW-START.

OSP-RCS-C102 (Cooldown Rate Log) is secured.

Initiating Cue:

The CRS directs you to restore RHR Loop B shutdown cooling per SOP-RHR-SDC, section 5.7, RHR Loop B Shutdown Cooling Quick Restart.

Steps 5.7.1 - 5.7.3 are complete (begin at step 5.7.4).

Page 10 of 10

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE HPCS SYSTEM INITIATION (MIN FLOW VALVE FAILS TO AUTO CLOSE)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001864 Rev. No. 1 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/29/16 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE SAT Coordinator APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

Initialize to IC-212 Execute associated Schedule/Event files Special Setup Instructions:

Insert a manual scram and allow RPV/L to recover to about 0 then trip both RFPs. Reduce RPV inventory to -40 inches Insert malfunction MOV-CSS004F to FAIL_AUTO_CLOSE Insert malfunction PMP-CSS001H to 30.00000 JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: None Safety Items: None Task Number: RO-0235 Validation Time: 13 minutes Alternate Path: Yes Time Critical: No PPM

Reference:

SOP-HPCS-INJECTION-QC Rev. 4 Location: Simulator ABN-SW Rev. 15 NUREG 1123 Ref: 209002 A1.01 (3.6 / 3.7) Performance Method: Perform Task Standard: HPCS Min Flow valve (HPCS-V-12) has been manually closed within 2 minutes of HPCS-V-4 being fully open. RPV level is in the band of +13 to +54 with HPCS-V-4 closed.

LR001567 Rev. 5 Page 2 of 6

JPM CHECKLIST INITIAL A failure of the master controller caused RPV level to drop. The Control Room Supervisor directed a manual scram. Both Reactor CONDITIONS: Feed Pumps tripped as RPV level approached 0. PPM 5.1.1, RPV Control, has been entered. RCIC system in not available.

INITIATING The CRS has directed you to initiate the HPCS system, verify proper system operation, and restore RPV level back to a band of CUE: +13 to +54. Inform the CRS when system operation has been verified, RPV level is in the band of +13 to +54, and HPCS-V-4 has been closed.

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 1 Step 2.1 Observes Level 8 amber seal-in light S/U Verify Reactor Level 8 Seal-in is not lit (HPCS-RMS-E22A/S6) is reset 2 Step 2.2 Rotates the collar in the clockwise If not already running, then ARM direction and depresses the S/U*

and DEPRESS the HPCS MANUAL pushbutton to initiate HPCS INITIATION pushbutton 3 Step 2.3 Observes HPCS-P-1 Red light on and Green light off. May also verify S/U Verify HPCS-P-1 running discharge pressure and amps 4 Step 2.4 2 minute clock starts upon step completion.

Observes HPCS-V-4 Red light on Verify HPCS-V-4 open (RPV S/U and Green light off Injection) Start Time: ___________ (min:sec)

LR001567 Rev. 5 Page 3 of 6

Evaluator note: Min Flow valve (HPCS-V-12) should automatically close when HPCS injection rate exceeds ~1300 gpm. Student should recognize that HPCS-V-12 is full open when it should have closed.

5 Recognize HPCS-V-12 failed to Manually close HPCS-V-12 when S/U*

close with RPV injection flow above RPV injection flow rate exceeds 1330 gpm. ~1330 gpm and within 2 minutes of start time (see previous step).

Evaluator note: Min Flow valve (HPCS-V-12) should automatically close when HPCS injection rate exceeds ~1300 gpm. Student should recognize that HPCS-V-12 is full open when it should have closed.

6 Step 2.5 Observes RPV level and when level S/U*

is approaching +13 and prior to Operate HPCS-V-4 as necessary to RPV level reaching +54 takes the maintain the desired RPV level control switch for HPCS-V-4 to closed and observes Green light on and Red light off. Level is in band immediately following HPCS-V-4 closure.

7 Verifies proper HPCS System Observes indications that HPCS DG S/U operation (given in Initiating Cue to is running perform)

Observes that HPCS-P-2 Green light S/U is off and Red light is on Termination Criteria: Student informs CRS that RPV level is in the band given, HPCS-V-4 is closed and may also add that HPCS-V-12 did not automatically close but was closed manually.

Terminating Cue: This JPM is complete.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LR001567 Rev. 5 Page 4 of 6

RESULTS OF JPM HPCS INITIATION WITH HPCS-P-2 FAILS TO AUTO START Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: HPCS Min Flow valve (HPCS-V-12) has been manually closed within 2 minutes of HPCS-V-4 being fully open. RPV level is in the band of +13 to +54 with HPCS-V-4 closed.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LR001567 Rev. 5 Page 5 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

A failure of the master controller caused RPV level to drop.

The Control Room Supervisor directed a manual scram.

Both Reactor Feed Pumps tripped as RPV level approached 0.

PPM 5.1.1, RPV Control, has been entered.

RCIC system in not available.

Initiating Cue:

The CRS has directed you to initiate the HPCS system, verify proper system operation, and restore RPV level back to a band of +13 to +54.

Inform the CRS when system operation has been verified, RPV level is in the band of +13 to +54, and HPCS-V-4 has been closed.

Page 6 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001823 Rev. No. 1 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Steve Bruce DATE 11/16/16 REVISED BY Dave E. Crawford DATE 01/13/17 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

INITIALIZE to IC-210 or any MODE with CR HVAC in normal line-up.

AND perform ONE the following:

Load JPM S-4.SCH file and JPM S-4.EVT files

Enter following into a newly created Schedule file and Event file:

Schedule file:

Insert malfunction MOV-RWB005F to FAIL_AUTO_OPEN

Insert malfunction MOV-RWB002F to CLOSE on event 1

Insert override OVR-RWB027A to OFF on event 1

Insert override OVR-RWB027B to OFF on event 1 Event file:

XWNI096C == 1 Special Setup Instructions:

Note: TRIGGER 1 - Removal of Fuse 3 in HVAC Panel COHV-2 on RW 525 can be accomplished by turning switch for TMU-V-18A (on H13-P824) to CLOSE (performed by evaluator) or via the Booth.

Provide copy of SOP-HVAC/CR-OPS with steps 5.12.1 & 5.12.2 initialed as complete.

JPM Instructions:

Verify current procedure against JPM. Revise JPM if any steps have changed.

Tools/Equipment: None Safety Items: None Task Number: RO-0502 Validation Time: 7 Minutes Alternate Path: Yes Time Critical: No PPM

Reference:

SOP-HVAC/CR-OPS Rev. 24 Location: Simulator NUREG 1123 Ref: 290003 A2.01 (3.1 / 3.2) Performance Method: Perform Task Standard: Control Room Ventilation Train B has been placed in Control Room Pressurization Mode.

LO001780 Rev. 1 Page 2 of 9

INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE JPM CHECKLIST INITIAL Control Room HVAC is normal operation with WMA-FN-51B running.

CONDITIONS:

INITIATING CRS has directed Control Room Ventilation Train B be placed in Pressurization Mode per SOP-HVAC/CR-OPS section 5.12.

CUE: Steps 5.12.1 and 5.12.2 are complete. Inform CRS when task is complete.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step Note: Unless otherwise noted, all control switches and annunciators are located on H13-P826.

Examiner Note: Start at step 5.12.4 for CR HVAC Train B.

1 Step 5.12.3 is N/A N/A because direction is to lineup S/U CR HVAC Train B 2 Step 5.12.4.a. Verified WMA-FN-51B Red light S/U ON and the Green light OFF VERIFY WMA-FN-51B running (Recirc Fan).

Examiner Note: CUE: If asked, these valves are verified locked open in the Lock Valve Checklist PPM 1.3.29 page 38.

Simulator Operator: If asked, verify as field operator that the valves are locked open on RW 525.

LO001780 Rev. 1 Page 3 of 9

INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 3 Step 5.12.4.b. Verified (4) valves open by If asked, valves in this step are S/U observing Red Light ON and Green verified locked open as annotated in VERIFY the following intake Light OFF on H13-P826. the Lock Valve Checklist PPM pathways locked open: (H13-P826) 1.3.29, page 38.

Verified LOCKED OPEN by Lock Remote Intake Number 1 (NW) Valve Checklist or Field Operator. If asked to verify as a field operator, Isol): report the valves in this step are verified locked open.

WOA-V-51A (Remote Air Intake No. 1 LOCKED OPEN WOA-V-52A (Remote Air Intake No. 1 LOCKED OPEN Remote Intake Number 2 (SE) Isol):

WOA-V-51B (Remote Air Intake No. 2 LOCKED OPEN WOA-V-52B (Remote Air Intake No. 2 LOCKED OPEN 4 Step 5.12.4.c. Rotated WOA-V-51C control-switch S/U*

to close and verified Red Light OFF CLOSE the following: and Green Light ON. Rotated WOA-V-52C control-switch to close and WOA-V-51C (Outside Air Intake) verified Red Light OFF and Green Light ON.

WOA-V-52C (Outside Air Intake)

LO001780 Rev. 1 Page 4 of 9

INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE Examiner Note: ROLEPLAY: Grant permission as CRS/Shift Manager to install temporary modification.

5 Step 5.12.4.d. Formally requested permission from ROLEPLAY: Grant permission as S/U CRS/Shift Manager to install CRS/Shift Manager to install REQUEST PERMISSION from temporary Modification per SOP- temporary modification.

CRS/Shift Manager to install HVAC/CR-OPS.

temporary modification to disable (fail closed) WMA-AD-51B1 (Fresh Air Inlet).

Examiner Note: Trigger 1 may be inserted by you using the switch for TMU-V-18A (located on H13-P824) by taking it to CLOSE or through coordination with the Booth Operator. Method to be used should be discussed before starting JPM.

Simulator Operator: If Trigger 1 is not being inserted directly by the evaluator, and when requested to remove fuse, insert Trigger 1 then call CRO as field operator and simultaneous verifier and report that Fuse 3 in HVAC Panel COHV-2 on RW 525 has been REMOVED.

6 Step 5.12.4.e. Confirmed field operator and If using local (switch) activation of S/U*

simultaneous verifier removed Fuse Trigger 1:

REMOVE Fuse 3 in HVAC Panel 3 in HVAC Panel COHV-2.

COHV-2. (Ref. EWD-84E-002) Take switch for TMU-V-18A (located on H13-P824) to CLOSE ROLEPLAY: Report as field operator and simultaneous verifier that Fuse 3 in HVAC Panel COHV-2 on RW 525 has been removed.

7 Step 5.12.4.f. Rotated WMA-FN-54B control- S/U*

switch to ON, observed Red Light START WMA-FN-54B by placing ON and Green Light OFF.

control switch in ON (Emergency Filter Unit Fan).

LO001780 Rev. 1 Page 5 of 9

INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE Step 5.12.4.g.1.

8 WMA-FN-54B Red Light ON and S/U Green Light OFF.

VERIFY the following occurs:

1) WMA-FN-54B starts.

Examiner Note: This next step constitutes the Alternate Path portion of the JPM.

Alt Path 9 Step 5.12.4.g.2. WMA-AD-54B1 Red Light OFF and S/U*

Green Light ON.

VERIFY the following occurs:

Examinee recognizes WMA-AD-

2) WMA-AD-54B1 OPEN (WMA- 54B1 failed to OPEN and turns FU-54B Inlet). control-switch to open and verifies Red Light ON and Green Light Off.

10 Step 5.12.4.g.3. WEA-FN-51 Red Light OFF and S/U Green Light ON VERIFY the following occurs:

3) WEA-FN-51 stops (Toilet/Kitchen Exhaust Fan).

11 Step 5.12.4.g.4. WEA-AD-51 Red Light OFF and S/U Green Light ON.

VERIFY the following occurs:

4) WEA-AD-51 CLOSED (Outlet Damper).

12 Step 5.12.4.g.5. Verified the following: S/U VERIFY the following occurs: WMA-AD-54B2 Red Light OFF and Green Light ON.

5) WMA-AD-54B2) CLOSED (WMA-FU-54B Inlet Bypass)

LO001780 Rev. 1 Page 6 of 9

INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE Termination Criteria: Control Room HVAC Train B has been placed in Control Room Pressurization Mode.

Terminating Cue: Student reports Control Room HVAC Train B has been placed in Control Room Pressurization Mode.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001780 Rev. 1 Page 7 of 9

RESULTS OF JPM INITIATE CR HVAC IN MANUAL PRESSURIZATION MODE Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard:

Control Room Ventilation Train B has been placed in Control Room Pressurization Mode.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001780 Rev. 1 Page 8 of 9

STUDENT JPM INFORMATION CARD Initial Conditions:

Control Room HVAC is in normal operation with WMA-FN-51B running.

Initiating Cue:

CRS has directed Control Room Ventilation Train B be placed in Pressurization Mode per SOP-HVAC/CR-OPS section 5.12.

Steps 5.12.1 and 5.12.2 are complete.

Inform CRS when task is complete.

Page 9 of 9

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC (Sim)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001637 Rev. No. 2 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 06/03/08 REVISED BY Dave E. Crawford DATE 12/6/16 TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC (Sim)

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

None Special Setup Instructions:

Reset to any IC. Turn off both ROA and REA fans. Allow secondary D/P to decay such that all expected annunciators are received. Acknowledge all associated annunciators.

JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: None Safety Items: None Task Number: RO-0497 Validation Time: 8 Minutes Alternate Path: No Time Critical: No PPM

Reference:

SOP-HVAC/ RB-RESTART-QC Rev. 1 Location: Simulator NUREG 1123 Ref: 290001 A4.01 (3.3 / 3.4) Performance Method: Perform Task Standard: ROA-FN-1A and REA-FN-1A are running and REA-DPIC-1A is in Auto and adjusted to achieve approximately -0.8 W.G.

LO001637 Rev.2 Page 2 of 7

RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC (Sim)

JPM CHECKLIST INITIAL A series of events occurred that resulted in no running Reactor Building Supply or Exhaust fan. PPM 5.3.1, Secondary Containment CONDITIONS: Control, was entered due to high Reactor Building differential pressure. Prior to starting Standby Gas Treatment, the Control Room received information that Reactor Building HVAC could be restarted.

INITIATING The CRC directs you to restart RB HVAC by starting ROA-FN-1A and REA-FN-1A per SOP-HVAC RB-RESTART-QC. Inform CUE: the CRS when Secondary Containment may be declared operable.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Results Step Cue Student uses the quick card to perform this JPM.

1 Step 2.1 Places toggle for REA-DPIC-1A to the Manual position. S/U*

Place REA-DPIC-1A(1B) (P Control RX Bldg/Outside) in manual.

2 Step 2.2 Depresses the open and/or closed pushbutton for S/U*

REA-DPIC-1A to have red indicator at Set REA-DPIC-1A(1B) output signal at approximately 60% of scale.

approximately 60% of scale 3 Step 2.3 Place the control switch for the following fans in Turns the handles counter-clockwise and pulls out to PULL-TO-LOCK: engage the Pull-To-Lock position for:

ROA-FN-1A (Reactor Bldg Supply Fan)

ROA-FN-1A (Reactor Bldg Supply Fan) S/U*

ROA-FN-1B (Reactor Bldg Supply Fan)

ROA-FN-1B (Reactor Bldg Supply Fan)

REA-FN-1A (Reactor Building Exhaust Fan)

REA-FN-1A (Reactor Building Exhaust Fan)

REA-FN-1B (Reactor Building Exhaust Fan)

REA-FN-1B (Reactor Building Exhaust Fan)

LO001637 Rev.2 Page 3 of 7

RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC (Sim)

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Results Step Cue 4 Step 2.4 Verify the following valves are open: Observes the Red light on and Green light off for:

S/U

ROA-V-1 (RB Supply Outboard Isolation)

ROA-V-1

ROA-V-2 (RB Supply Inboard Iso)

ROA-V-2

REA-V-1 (RB Exhaust Inboard Iso)

REA-V-1

REA-V-2 (RB Exhaust Outboard Isol)

REA-V-2 5 Step 2.5 (2H) Depresses the control switch handle for REA-FN-1A If student attempts to make a plant to the neutral position. Turns the same handle announcement S/U*

Place REA-RMS-FN1A(B) in Start (Reactor Bldg clockwise to the Start position and then releases it. STOP THEM.

Exhaust Fan Control Switch). Inform them announcement has been made.

6 Step 2.6 Observes the red light for REA-FN-1A comes on.

S/U*

When REA-FN-1A(1B) breaker closure is observed Depresses the control switch handle for ROA-FN-(red light), then immediately place ROA-RMS- 1A to the neutral position. Turns the same handle to FN1A(B) in Start (Reactor Bldg Supply Fan the Start position and releases it.

Control Switch).

7 Step 2.7 S/U*

Manually adjust REA-DPIC-1A(1B) controller Adjusts REA-DPIC-1A to achieve approximately -

output until Reactor Building pressure on 0.8 W.G.. on REA-DPR-1A.

REA-DPR-1A(1B) is approximately -0.8 W.G.

8 Step 2.8 Turns thumbwheel until REA-DPIC-1A is nulled or waits until red arrow lines up with green band and S/U*

Null REA-DPIC-1A (1B), and place then moves lever to AUTO position.

REA-DPIC-1A (1B) in AUTO.

LO001637 Rev.2 Page 4 of 7

RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC (Sim)

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Results Step Cue 9 Step 2.9 Depresses the control switch handles for ROA-FN-1B and REA-FN-1B from PTL and allows switches Place the control switch for the following non- S/U to go to the neutral position.

running fans in the NORMAL-after-STOP position.

ROA-FN-1B(1A) Observes the green flag is visible on each switch.

REA-FN-1B(1A) 10 Step 2.10 Notify the CRS that Secondary Containment may Informs the CRS that Secondary Containment may S/U be declared operable. be declared operable.

Termination Criteria: Student informs CRS that Secondary Containment may be declared operable.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001637 Rev.2 Page 5 of 7

RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC (Sim)

RESULTS OF JPM RE-ESTABLISH SECONDARY CONTAINMENT/START RB HVAC Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: ROA-FN-1A and REA-FN-1A are running and REA-DPIC-1A is in Auto and adjusted to achieve approximately -0.8 W.G.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001637 Rev.2 Page 6 of 7

STUDENT JPM INFORMATION CARD Initial Conditions:

A series of events occurred that resulted in no running Reactor Building Supply or Exhaust fan.

PPM 5.3.1, Secondary Containment Control, was entered due to high Reactor Building differential pressure.

Prior to starting Standby Gas Treatment, the Control Room received information that Reactor Building HVAC could be restarted.

Initiating Cue:

The Control Room Supervisor has directed you to restart RB HVAC by starting ROA-FN-1A and REA-FN-1A per SOP-HVAC RB-RESTART-QC.

Inform the CRS when Secondary Containment may be declared operable.

LO001637 Rev.2 Page 7 of 7

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE LOWER RPV PRESSURE USING DEH (CR/SIM) (Alt Path)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001780 Rev. No. 0 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 10/21/14 REVISED BY DATE TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

Reset to IC with reactor shutdown and pressure being controlled by bypass valves.

Special Setup Instructions:

Insert MAL-DEH017.

Set Pressure Rate to any value other than 50 psig.

JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: None Safety Items: None Task Number: RO-0348 Validation Time: 7 Minutes Alternate Path: Yes Time Critical: No PPM

Reference:

SOP-DEH-QC Rev. 5 Location: Simulator NUREG 1123 Ref: 241000 A4.02 (4.1 / 4.1) Performance Method: Perform Task Standard: Recognize that auto control of bypass valves to lower RPV pressure to a target of 550 psig does not work and RPV pressure has been lowered at a rate LE 50 psig per minute and in a controlled manner by taking manual control of the Bypass Valves.

LO001780 Rev.0 Page 2 of 9

JPM CHECKLIST INITIAL Columbia was operating at full power when RFW-P-1A tripped followed by RFW-P-1B tripping five minutes later. CRO1 has lined CONDITIONS: up on the startup flow control valves.

INITIATING The CRS directs you to lower RPV pressure to 550 psig at the rate of 50 psig per minute to facilitate feeding the RPV with the CUE: Condensate Booster pumps per SOP-DEH-QC. Inform the CRS when RPV pressure is 550 psig.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step NOTE: If the plant is operating in Mode 1 and is GT 25% power, then the DEH set point should be 960 psi. If a reactor pressure change is desired refer to ABN-PRESSURE.

1 Step 2.1.1 Selects a display screen.

S/U Initiate Pressure setpoint change as follows (Turbine Start Up; Reactor Startup Display) or (Main Display):

2 a. Select Pressure Target. Selects Pressure Target. S/U*

3 b. Enter desired pressure. Enters 5,5,0 psig. S/U*

4 c. Select OK. Selects OK. S/U*

5 d. If a change in pressure rate is Observes Pressure Rate is not 50 S/U desired, then perform the psig and performs step.

following:

6 1) Select Pressure Rate. Selects Pressure Rate. S/U*

7 2) Enter desired Pressure rate. Enters 5,0. S/U*

8 3) Select OK. Selects OK. S/U*

9 e. Select GO. Selects GO. S/U*

10 f. Select YES. Selects YES. S/U*

LO001780 Rev.0 Page 3 of 9

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step ALTERNATE 11 g. Verify Press Demand and Throttle Observes no change in Pressure CRS - What actions would you PATH STEP S/U*

Press change at the Pressure Rate. Demand or Bypass Valve position. recommend to lower RPV pressure to 550 psig at 50 psig per minute?

Observes green Hold light is still illuminated.

Informs the CRS.

EVALUATOR: If SOP-DEH-OPS is referenced, when section for manual bypass valve operation is found, cue to use the DEH Quick Card.

[OPTION A]

12 SOP-DEH-QC Step 2.2 Performs this section.

S/U Manual Bypass Valve Operation.

NOTE: In Manual, raising BPV demand will open the BPVs and cause Reactor pressure to lower. The BPVs will not respond to pressure changes in Manual.

13 Step 2.2.1 Performs this step.

S/U Operate the Bypass Valves Manually as follows (Turbine Start-up, Reactor Start screen):

NOTE: In manual, raising BPV demand will open the BPVs and cause Reactor pressure to lower. The BPVs will not respond to pressure changes in Manual.

14 a. Select BPV MANUAL. Selects BPV Manual. S/U*

15 b. Select YES. Selects Yes. S/U*

16 c. If rapid Bypass Valve movement is Verbalizes step (it is anticipated that desired, then select FAST ACTION. S/U this step will not be performed but it is OK if it is performed).

17 d. If opening Bypass Valves, then Selects BPV Raise.

S/U*

select BPV Raise.

18 e. If closing Bypass Valves, then Does not perform this step.

S/U select BPV Lower.

LO001780 Rev.0 Page 4 of 9

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step NOTE: The JOG button illuminates green when the command is accepted, and extinguishes when the command is complete.

EVALUATOR: Either step f or steps g, h, and i are performed to lower RPV pressure. Whichever is performed makes the other steps not critical steps. It is anticipated that step f will be used to reduce RPV pressure.

19 f. If incremental Bypass Valve Depresses the JOG button to achieve movement is desired, then depress JOG S/U*

approximately a 50 psig pressure button once for each 1% of valve drop per minute.

demand change desired.

If JPM step 19 was performed, skip JPM steps 20, 21, and 22 (which are now NOT critical step).

20 g. Select GO for full range motion to Selects Go.

100% demand or 0% demand.

S/U*

21 h. Select YES. Selects Yes and observes bypass S/U*

valves starting to open.

22 i. If desired to stop BPV motion, then Selects Hold to stop bypass valve depress hold. S/U*

motion.

EVALUATOR: No matter which steps were performed to lower pressure - When you determine that RPV pressure is being lowered in a controlled manner at less than or equal to 50 psig per minute, inform the operator that at the next RPV pressure 50 psig increment to close the Bypass Valves and stop the pressure reduction.

23 Stops RPV pressure reduction and Selects BPV Lower. S/U*

closes Bypass Valves.

EVALUATOR: Student may perform JPM step 24 or steps 25, 26, and 27 to close the Bypass Valves. Whichever is performed makes the other NOT critical.

24 Selects the JOG button until the S/U*

BPVs are closed.

LO001780 Rev.0 Page 5 of 9

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step Step 24 or 25-27 will be performed.

25 Selects Fast Action. S/U*

26 Selects Go. S/U*

27 Selects Yes. S/U*

IF SOP-DEH-OPS is utilized, these steps will be performed. [OPTION B]

28 SOP-DEH-OPS Step 5.5.1: Performs this step. S/U Operate the Bypass Valves Manually as follows:

29 a. Select BPV Manual. Selects BPV Manual S/U*

30 b. Select YES. Selects Yes S/U*

31 c. Verify BPV Manual illuminates. Observes BPV Manual illuminates. S/U NOTE: In BPV Manual mode the ramp rate is 1 %/sec (valve position) if the BPV RAISE or LOWER and GO buttons are used. The ramp rate is 5 %/sec if the FAST ACTION button is used with BPV RAISE or LOWER and GO. BPV position changes in 1% increments if the JOG button is used with BPV RAISE or LOWER.

32 d. If incremental Bypass Valve Depresses the JOG button to achieve movement is desired, then depress S/U*

approximately a 50 psig pressure JOG button once for each 1% of drop per minute.

valve demand change desired.

Candidate may decide to do rapid valve movement and perform the following steps instead of step d above. If the below steps are performed then above step is not critical. If d is performed then e, f, g ,and h are not critical.

33 e. If rapid Bypass valve movement is Selects Fast Action. S/U*

desired, then select fast action.

34 f. If lowering pressure, then select Selects BPV raise. S/U*

BPV raise.

35 g. If raising pressure, then select Does not perform this step. S/U*

BPV lower.

36 h. Select GO for full range motion to Selects Go. S/U*

100% demand or 0% demand.

LO001780 Rev.0 Page 6 of 9

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 37 i. Select yes. Select yes. S/U*

38 j. Verifies Go illuminates Observes Go illuminates. S/U*

39 k. Monitor BPV position and RPV Monitors BPV position. S/U pressure during BPV motion.

40 l. If desired to stop BPV motion, then Selects Hold S/U*

perform the following:

1. Select Hold Observes Hold illuminates.

2. Verify Hold illuminates.

EVALUATOR: No matter which steps were performed to lower pressure - When you determine that RPV pressure is being lowered in a controlled manner at less than or equal to 50 psig per minute, inform the operator that at the next RPV pressure 50 psig increment to close the Bypass Valves and stop the pressure reduction.

41 Stops RPV pressure reduction and Selects BPV Lower. S/U*

closes Bypass Valves.

EVALUATOR: Student may perform JPM step 42 or steps 43, 44, and 45 to close the Bypass Valves. Whichever is performed makes the other NOT critical.

42 Selects the JOG button until the S/U*

BPVs are closed.

Step 42 or 43-45 will be performed.

43 Selects Fast Action. S/U*

44 Selects Go. S/U*

45 Selects Yes. S/U*

Termination Criteria: When the BPVs are closed, inform the Student that the termination point of the JPM has been reached.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001780 Rev.0 Page 7 of 9

RESULTS OF JPM LOWER RPV PRESSURE USING BPVs IN MANUAL Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: Recognize that auto control of bypass valves to lower RPV pressure to a target of 550 psig does not work and RPV pressure has been lowered at a rate LE 50 psig per minute and in a controlled manner by taking manual control of the Bypass Valves.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001780 Rev.0 Page 8 of 9

STUDENT JPM INFORMATION CARD Initial Conditions:

Columbia was operating at full power when RFW-P-1A tripped followed by RFW-P-1B tripping five minutes later.

CRO1 has lined up on the startup flow control valves.

Initiating Cue:

The CRS directs you to lower RPV pressure to 550 psig at the rate of 50 psig per minute to facilitate feeding the RPV with the Condensate Booster pumps per SOP-DEH-QC.

Inform the CRS when RPV pressure is 550 psig.

Page 9 of 9

INSTRUCTIONAL COVER SHEET PROGRAM TITLE INITIAL LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE RESTORE RPS A FROM ALTERNATE POWER SOURCE (SIMULATOR)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE LO001597 Rev. No. 1 EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Ron Hayden DATE 9/02/06 REVISED BY Dave Crawford DATE 12/20/16 TECHNICAL REVIEW BY INSTRUCTIONAL REVIEW BY APPROVED BY Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

RESTORE RPS A FROM ALTERNATE POWER SOURCE MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

Any IC with a normal electrical lineup - all load centers energized Ensure AR-EX-1B is in service Special Setup Instructions:

Open RPS EPA BKR 3A, acknowledge all annunciators, and allow plant to stabilize.

JPM Instructions:

Verify the current procedure against the JPM. If the procedure is a different revision than listed in the JPM, ensure the critical steps still match. If the critical steps have changed, the JPM should be revised.

The evaluator and student shall use current procedure. The evaluator should mark off steps as they are completed, note comments, and transfer the comments to the Results of JPM page.

Tools/Equipment: None Safety Items: None Task Number: RO-0248 Validation Time: 15 minutes Alternate Path: Time Critical: NO PPM

Reference:

ABN-RPS Rev. 11 Location: Simulator NUREG 1123 Ref: 212000 A4.14 (3.8/3.8) Performance Method: Perform Task Standard: The subsequent actions for ABN-RPS have been completed and RPS A has been re-powered from Alternate Power Supply.

LO001597 Rev. 1 Page 2 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE JPM CHECKLIST INITIAL A loss of RPS A occurred 20 minutes ago. All maintenance and surveillance testing has been stopped. Investigation revealed a CONDITIONS: failure of the A RPS MG set motor.

INITIATING The CRS has directed you to transfer A RPS to its Alternate power supply by performing steps 4.1 through 4.8 of ABN-RPS.

CUE: Inform the CRS when the subsequent actions for ABN-RPS have been completed and A RPS has been restored.

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat Examiner Note: The candidate is given ABN-RPS 1 Step 4.1 Takes control switch for RWCU-V- S/U*

104 to the open position until it IF power is available to RWCU-V-indicates dual (both Red and Green 104 (Cleanup System Bypass),

lights lit for RWCU-V-104)

THEN THROTTLE OPEN RWCU-V-104.

2 Step 4.2 N/A S/U IF the alternate Gland Exhauster AR-EX-1B continues to run.

(AR-EX-1A(B) is required, THEN START the alternate Gland Exhauster, AND PLACE the tripped Gland Exhauster in OFF.

LO001597 Rev. 1 Page 3 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 3 Step 4.3 N/A S/U IF the alternate Mechanical Vacuum AR-EX-1B continues to run.

Pump (AR-P-1A(B) is required, THEN START the alternate Mechanical Vacuum Pump.

4 Step 4.4 Verifies: S/U ENSURE automatic actions have

Half Scram occurred.

Half trip on RC-1

Half trip on RC-2

The following valves close:

o RRC-V-20 o FDR-V-4 o EDR-V-20 o RWCU-V-4 o RHR-V-8 o RHR-V-40 o RHR-V-23 o RHR-V-53A and 53B o MS-V-67A-D o MS-V-19

AR-EX-1A trips

AR-P-1A trips

APRM Chassis 1 and 3 default to RUN Refers to Attachment 7.1 for list of annunciation Note: Due to loss of RPS A power to APRM Voter 1 and 3, the APRM Chassis 1 and 3 will default to the RUN setpoint. Due to loss of RPS B power to APRM Voter 2 and 4, the APRM Chassis 2 and 4 will default to the RUN setpoint.

LO001597 Rev. 1 Page 4 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 5 Step 4.5 N/A S/U IF the Reactor Mode switch is not in Mode Switch is in RUN.

the RUN position, THEN REFER to Technical Specifications for the required actions.

6 Step 4.6 Proceeds to 4.6.1. S/U PERFORM the following to reenergize RPS:

7 Step 4.6.1 N/A S/U IF the condition of the RPS MG set The status of the A RPS MG set is is known to be operable, NOT known to be operable. The initiating cue directs the candidate to AND the RPS bus is known to be restore power from the Alternate operable, Source.

THEN RESTART the RPS MG set, AND REPOWER the bus per SOP-RPS-START and SOP-RPS-OPS.

8 Step 4.6.2. Proceeds to 4.6.2.a S/U IF the condition of the RPS MG set is uncertain, THEN REPOWER RPS A or B from H13-P610 as follows:

LO001597 Rev. 1 Page 5 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 9 Step 4.6.2.a. Check power available from the RPS Alternate Power Supply, MC-6B, by VERIFY power available from the observing the Alternate Feed white Reactor Protection System Alternate light illuminated.

Power Supply, MC-6B, by observing S/U the Alternate Feed white light illuminated.

CAUTION: The MG Set Transfer switch is break before make and positioning it to the wrong supply will result in a full REACTOR SCRAM.

10 Step 4.6.2.b. Place the RPS power source selector S/U*

IF repowering RPS A, switch in the position (ALT A) to be powered from the Alternate Supply THEN PLACE RPS Power Source Select switch in ALT A position.

11 Step 4.6.2.c. N/A S/U IF repowering RPS B, RPS B was not de-energized.

THEN PLACE RPS Power Source Select switch in ALT B position.

12 Step 4.7. If restoring RPS A then perform the S/U following:

When RPS power has been restored stabilized, then perform the following:

LO001597 Rev. 1 Page 6 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 13 Step 4.7.a. Reset the Half Scram at H13-P603 S/U*

RESET the Half SCRAM at H13-P603.

14 Step 4.7.b. Reset Main Steam Line Rad Monitor S/U*

alarms at H13-P606.

RESET Main Steam Line Rad Monitor alarms at H13-P606:

MS-RIS-610A

MS-RIS-610C 15 Step 4.7.c. Depresses the Isolation logic A & S/U*

DEPRESS the following pushbuttons B and Isolation logic C & D reset at H13-P601: pushbuttons at H13-P601.

Isolation logic A&B reset pushbutton

Isolation logic C&D reset pushbutton 16 Step 4.7.1.d. N/A Inform the candidate that another RETURN RWCU to service per Read the candidate the cue for this operator is placing RWCU into S/U SOP-RWCU-START. step. service and to continue with ABN-RPS.

17 Step 4.7.1.e Resets RC-1 by depressing WMA- S/U*

RESET RC-1 by depressing WMA- RMS-FAZ/3AXY pushbutton.

RMS-FAZ/3AXY pushbutton.

18 Step 4.7.1.f. Resets RC-2 by depressing WMA- S/U*

RMS-FAZ/3BXY pushbutton.

RESET RC-2 by depressing WMA-RMS-FAZ/3BXY pushbutton.

LO001597 Rev. 1 Page 7 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE

Items are Critical Steps Time Step Element Standard Cue Sat/Unsat 19 Step 4.7.1.g. N/A S/U RHR SDC was in service, SDC was not in service.

THEN REFER to ABN-RHR-SDC-LOSS.

20 Step 4.7.1.h. Opens RRC-V-20 S/U*

OPEN RRC-V-20.

21 Step 4.7.1.i. Opens EDR-V-20 S/U*

OPEN EDR-V-20.

22 Step 4.7.2 N/A S/U RPS B remains energized.

23 Step 4.8. Opens FDR-V-3 and 4 FDR-V-3 and FDR-V-4 have been isolated for LT 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , S/U*

THEN OPEN the following: (H13-P601)

FDR-V-3

FDR-V-4 Termination Criteria: Candidate completes steps 4.1 through 4.8 of ABN-RPS.

Termination Cue: This completes the JPM.

Transfer the following to the Results of JPM page: Any Unsat step - indicate if step was a Critical Step; JPM completion time.

LO001597 Rev. 1 Page 8 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE RESULTS OF JPM RESTORE RPS A FROM ALTERNATE POWER SOURCE Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: The subsequent actions for ABN-RPS have been completed and RPS A has been re-powered from Alternate Power Supply.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

LO001597 Rev. 1 Page 9 of 10

RESTORE RPS A FROM ALTERNATE POWER SOURCE STUDENT JPM INFORMATION CARD Initial Conditions:

A loss of RPS A occurred 20 minutes ago.

All maintenance and surveillance testing has been stopped.

Investigation revealed a failure of the A RPS MG set motor.

Initiating Cue:

The CRS has directed you to transfer A RPS to its Alternate power supply by performing steps 4.1 through 4.8 of ABN-RPS.

Inform the CRS when the subsequent actions for ABN-RPS have been completed and A RPS has been restored.

Page 10 of 10

INSTRUCTIONAL COVER SHEET PROGRAM TITLE LICENSED OPERATOR TRAINING COURSE TITLE JOB PERFORMANCE MEASURE LESSON TITLE RESPOND TO INDICATIONS OF A FIRE (CR/SIM)

LESSON LENGTH .5 HRS INSTRUCTIONAL MATERIALS INCLUDED LESSON PLAN PQD CODE Rev. No.

SIMULATOR GUIDE PQD CODE Rev. No.

JPM PQD CODE Rev. No.

EXAM PQD CODE Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave Crawford DATE 12/20/16 REVISED BY DATE TECHNICAL REVIEW BY DATE INSTRUCTIONAL REVIEW BY DATE APPROVED BY DATE Operations Training Manager Verify materials current IAW SWP-TQS-01 prior to use

MINOR REVISION RECORD Minor Description Affected Entered Effective Manager Rev Number of Revision Pages By Date Approval JPM SETUP Simulator ICs; Malfunctions; Triggers; Overrides:

INITIALIZE to IC-212 or ANY MODE.

AND PERFORM the following:

Load NRCjpms-8.sch OR insert the following:

1. START FP-P-2A

2. START FP-P-2B

3. Insert ANN-FCP1A05A to ON (TG Bldg 471 East Mezzanine)

4. Insert ANN-FCP1A05C to ON (TG Bldg 471 East Mezzanine FIRE)

5. Insert ANN-FCP2C03A to ON (SYS 9 Wet Pipe TG Bldg 471 East Mezzanine)

6. Insert ANN-FCP2C03C to ON (SYS 9 Wet Pipe TG Bldg 471 East Mezzanine Fire)

Special Setup Instructions:

N/A JPM Instructions:

Verify Current Procedure against JPM. If any steps have changed, the JPM should be revised.

Tools/Equipment: None Safety Items: None Task Number: RO-0124 Validation Time: 5 Minutes Alternate Path: No Time Critical: No PPM

Reference:

ABN-FIRE Location: Simulator NUREG 1123 Ref: 286000.A4.01 (3.3 / 3.2) Performance Method: Perform Task Standard: ABN-FIRE Immediate Actions have been completed.

Page 2 of 6

JPM CHECKLIST INITIAL Given the following:

CONDITIONS:

Report from OPS-3 that there is heavy smoke on the TB-501 East End coming up from below and he was leaving the area.

Alarms on FCP-1, FCP-2, and FCP-3:

TG BLDG 471 EAST END Mezzanine Fire alarm.

SYS 9 WET PIPE TG BLDG 471 EAST MEZZANINE Fire alarm.

FP-P-2A PUMP RUNNING.

FP-P-2B PUMP RUNNING.

INITIATING

The CRS has directed you to take actions in response to the given conditions.

CUE:

From memory, take appropriate actions and report to the CRS when you are complete.

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 1 Reviews alarms on FCP-1, FCP-2, Determined by multiple conditions S/U and FCP-3: that a Fire exists in the TB 471 East

TG BLDG 471 End.

EAST END Mezzanine Fire alarm.

SYS 9 WET PIPE TG BLDG 471 EAST MEZZANINE Fire alarm.

FP-P-2A PUMP RUNNING.

FP-P-2B PUMP RUNNING.

Page 3 of 6

Items are Critical Steps Time JPM Task Element Performance Standard Evaluators Cue Results Step 2 Reviews report from OPS-3. Determined that local verification S/U requirement has been met.

3 Enters ABN-FIRE. Enters ABN-FIRE. S/U 4 Performs Immediate Actions of Performed Immediate Actions of S/U ABN-FIRE. ABN-FIRE.

5 SOUND the alerting tone for ~5 Sounded the Alerting Tone for S/U*

seconds. approximately 5 seconds.

6 ANNOUNCE the location of the Announced the location of the fire as S/U*

fire. TB 471 East End.

7 IF the fire is WITHIN a Protected Dispatched Fire Brigade. S/U*

Area Plant Building, THEN DISPATCH the Fire Brigade 8 REPEAT the above three steps Repeat steps 3.1, 3.2, and 3.3. S/U*

9 IF the fire is not extinguished, Depressed the Hanford Fire S/U*

THEN DEPRESS the Hanford Fire Department pushbutton on FCP-1.

Department pushbutton on FCP-1.

Examiner Note: Stop the JPM if examinee attempts to continue past the Immediate Actions.

Terminating Cue: This JPM is complete.

Page 4 of 6

RESULTS OF JPM RESPOND TO INDICATIONS OF A FIRE Examinee (Print): _________________________________________________________________

Evaluator (Print): _________________________________________________________________

Task Standard: ABN-FIRE Immediate Actions have been completed.

Overall Evaluation JPM Completion Time SAT / UNSAT (Circle One) Minutes COMMENTS:

Evaluator's Signature: Date:

Page 5 of 6

STUDENT JPM INFORMATION CARD Initial Conditions:

Given the following:

Report from OPS-3 that there is heavy smoke on the TB-501 East End coming up from below and he was leaving the area.

Alarms on FCP-1, FCP-2, and FCP-3:

TG BLDG 471 EAST END Mezzanine Fire alarm.

SYS 9 WET PIPE TG BLDG 471 EAST MEZZANINE Fire alarm.

FP-P-2A PUMP RUNNING.

FP-P-2B PUMP RUNNING.

Initiating Cue:

The CRS has directed you to take actions in response to the given conditions.

FROM MEMORY, take appropriate actions and report to the CRS when you are complete.

Page 6 of 6

INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE COLUMBIA GENERATING STATION SIMULATOR EXAMINATION RFW-DPT-4A fails downscale (Tech Spec); RWCU NRHX fouling causes high temperature isolation on RWCU-V-4; CRD-P-1A trips requiring CRD-P-1B to be started; HPCS-P-1 LESSON TITLE control power failure (Tech Spec); RRC-FT-14A fails low causing APRM-CHS-1 to trip; SRV MS-RV-2B inadvertently opens (will close upon fuse removal); LOCA from RRC-P-1B suction line requiring manual scram; Spray Wetwell and Drywell; RFW-FIC-620 controller failure with RFW-V-109 failing to open and RFW-V-112A & B failing to open once closed; RCIC-FIC-600 fails low on startup requiring manual trip of RCIC turbine; Initiate Emergency Depressurization (ED) on low RPV level and restore RPV level to above TAF LENGTH OF LESSON 1.5 Hours Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code LO001856 Rev. No. 0 JPM PQD Code Rev. No.

Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/22/16 REVISED BY DATE VALIDATED BY DATE TECHNICAL REVIEW DATE INSTRUCTIONAL REVIEW DATE APPROVED DATE NRC Scenario 1 Page 1 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Facility: Columbia Generating Scenario No.: 1 Op Test No.: 1 Station Examiners: Operators:

The reactor is in Mode 1 at 100% power. RCIC Operability Test surveillance was just completed to satisfy Post Maintenance Testing (PMT) requirements and has been returned to a Standby status and declared operable. RHR-SYS-B was placed in Suppression Pool Cooling three (3) hours ago to Initial Conditions:

restore Suppression Pool temperature following the testing and to satisfy RHR-P-2B PMT requirements. LCO 3.5.1 A.1, LCO 3.6.1.5 A.1, LCO 3.6.2.3 A.1, and RFO 1.6.1.5 A.1 have been entered for RHR-SYS-B being inoperable.

Turnover: Maintain RHR-P-2B in operation for the next three (3) hours to satisfy pump PMT requirements.

Critical Tasks:

Initiate Drywell sprays when Wetwell pressure exceeds 12 psig but prior to exceeding PSP, after verifying CT-1 Drywell parameters are within DSIL and RHR is NOT required for adequate core cooling.

CT-2 Initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS preferred) after RPV water level reaches TAF (-161 inches) and within 10 minutes of level dropping below TAF.

CT-3 After ED, and within 10 minutes of RPV pressure lowering to 200 psig, restore and maintain RPV water level above TAF (-161 inches) using Low Pressure ECCS systems.

Event Malf Event Type* Event Description No. No.

I (ATC,SRO) 1 TRG-1 RFW-DPT-4A fails downscale (Tech Spec)

TS (SRO) 2 TRG-2 C (BOP,SRO) RWCU NRHX fouling causes high temperature isolation on RWCU-V-4 3 TRG-3 C (ATC,SRO) CRD-P-1A trips requiring CRD-P-1B to be started 4 TRG-4 TS (SRO) HPCS-P-1 control power failure (Tech Spec) 5 TRG-5 I (ATC,SRO) RRC-FT-14A fails low causing APRM-CHS-1 to trip C (BOP,SRO) 6 TRG-6 SRV MS-RV-2B inadvertently opens (will close upon fuse removal)

R (ATC,SRO)

LOCA from RRC-P-1B suction line requiring manual scram 7 TRG-7 M (ALL)

Spray Wetwell and Drywell (CT #1)

RFW-FIC-620 controller failure with RFW-V-109 failing to open and RFW-8 N/A C (ATC,SRO)

V-112A & B failing to open once closed 9 N/A C (BOP) RCIC-FIC-600 fails low on startup requiring manual trip of RCIC turbine Initiate Emergency Depressurization (ED) on low RPV level and restore 10 N/A ---

RPV level to above TAF (CT #2) (CT #3)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications NRC Scenario 1 Page 2 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Target Quantitative Attributes Actual Description Inability to inject with feedwater; RCIC-FIC-600 output fails Malfunctions after EOP entry (1-2) 2 low RWCU NRHX fouling; CRD-P-1A trip; SRV MS-RV-2B Abnormal events (2-4) 3 opens Major transients (1-2) 1 LOCA from RRC-P-1B suction line PPM 5.1.1 (RPV Control); PPM 5.2.1 (Primary Containment EOPs entered/requiring substantive actions (1-2) 2 Control)

EOP contingencies requiring substantive actions 1 PPM 5.1.3 (Emergency RPV Depressurization)

(0-2)

Critical tasks (2-3) 3 See Critical Task Determination table Trigger Evaluator How Purpose Malfunction Numbers (TRG-x) Directed Triggered TRG-1 YES Manually Event Initiator XMT-RRS106A TRG-2 YES Manually Event Initiator HTX-RCC010F TRG-3 YES Manually Event Initiator BKR-CRD001 TRG-4 YES Manually Event Initiator BKR-CSS001 TRG-5 YES Manually Event Initiator XMT-RRS036A TRG-6 YES Manually Event Initiator OVR-RRS022D TRG-7 YES Manually Event Initiator MAL-RRS004B TRG-8 Automatically Malf Trigger MAL-RRS004D TRG-9 Automatically Malf Trigger MOV-CFW044F TRG-10 Automatically Malf Trigger MOV-CFW045F TRG-11 YES(2) Manually Malf Trigger BKR-CFW004; BKR-CFW005; BKR-CFW006 Initial Condition MAL-FWC011 Initial Condition MOV-CFW043F Initial Condition CNH-RCI002B Initial Condition MAL-RWU008 (2) Contingency action (see Event 8 description).

NRC Scenario 1 Page 3 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 SCENARIO 1

SUMMARY

Event 1 (TRG-1) The first event is a failure of RFW-DPT-4A/RFW-LI-606A downscale. The crew performs actions per Alarm Response Procedure (ARP) 603.A8.3-7 and enters ABN-INSTRUMENTATION and ABN-LEVEL. The CRS directs the BOP to verify FWLC system has automatically shifted to Channel B and directs the ATC to place the Reactor Vessel Level Control Channel selector switch to Channel B. The SRO refers to Technical Specifications and determines that LCO 3.3.2.2, action A.1 applies which requires tripping the affected channel within 7 days.

Event 2 (TRG-2) Reactor Water Cleanup Non-Regenerative Heat Exchanger (RWCU-HX-2A/2B NRHX) fouling causes a rising temperature at the outlet of the NRHX leading to the RWCU filter demineralizers. The crew takes actions per ARP 4.602.A5 6-8 (CLEANUP FLTR INLET TEMP HI) to include monitoring temperature, verifying system lineup, and ensuring proper Reactor Closed Cooling (RCC) flow to the RWCU NRHX exist. When the crew recognizes that the RWCU NRHX outlet temperature is approaching 140°F, and isolation appears imminent, the BOP operator will stop the running RWCU pump (RWCU-P-1A) and close the RWCU Suction Outboard Isolation Motor-Operated Valve (MOV) (RWCU-V-4). The crew may also elect to close the Inboard isolation MOV (RWCU-V-1).

Note that if the crew does not take action to prevent automatic closure of RWCU-V-4 (which occurs at 140°F), RWCU-V-4 will automatically close but the running RWCU pump (RWCU-P-1A) will fail to trip.

The BOP operator will have to manually trip the pump.

Event 3 (TRG-3) Control Rod Drive Pump 1A (CRD-P-1A) inadvertently trips requiring the ATC operator to start CRD-P-1B per ARP H13-P603 A-7 3-8 (CRD CHARGE WATER PRESS LOW). Actions include placing the CRD Flow Controller in Manual, zeroing the output and then starting CRD-P-1B. The controller is then nulled and placed back in Auto and CRD system parameters restored.

Depending on the time required to restore CRD flow, one or more control rod HCU accumulator alarms may come in. If asked, local accumulator pressures will be reported to be 980 psig which is below the alarm setpoint but above the LCO 3.1.5 minimum limit of 940 psig required for operability. No Technical Specification actions will be required.

Event 4 (TRG-4) High Pressure Core Spray (HPCS-P-1) control power fails (fuses blow) due to electrical fault.

The BOP operator refers to ARP 601.A1 6-8 (HIGH PRESSURE CORE SPRAY SYSTEM OUT OF SERVICE). If directed to investigate, the HPCS pump control power fuses are reported as blown. Any attempt to replace fuses will result in fuses again blowing.

With both RHR-SYS-B and HPCS inoperable, the CRS refers to Technical Specifications and determines the following additional actions apply:

LCO 3.5.1 B.1 - Immediately verify by administrative means that RCIC is operable

LCO 3.5.1 B.2 - Restore HPCS system to operable status within 14 days

LCO 3.5.1 C.1 - Restore RHR-SYS-B or HPCS system to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months NRC Scenario 1 Page 4 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Event 5 (TRG-5) A downscale failure of Reactor Recirculation Flow Transmitter 14A (RRC-FT-14A) occurs causing Channel 1 of the Average Power Range Monitor (APRM-CHS-1) to trip. With only one (1) vote sent to the 2-out-of-4 voter logic no half-scram or reactor trip signals are generated. The crew takes actions per annunciator 603.A8 3-6 (FLOW REFERENCE OFF NORMAL). The CRS directs the ATC operator to bypass APRM-CHS-1.

With APRM-CHS-1 inoperable (and bypassed), the CRS refers to Technical Specifications and determines that only three (3) APRM channels are required to be operable and that no Technical Specification actions are required.

Event 6 (TRG-6) Non-ADS Safety Relief Valve (SRV) MS-RV-2B inadvertently opens. The crew confirms this by observing at least one of the following: 1) Rise on MS-RV-2B tailpipe temperature on MS-TR-614; 2)

Rising Suppression Pool temperature or level; or 3) Reduction in Main Generator output of ~70 MWe.

The CRS enters ABN-SRV and directs the ATC operator to reduce reactor power to < 90% using Reactor Recirculation (RRC) flow. The BOP attempts to close the SRV using the control switch. The valve will not close requiring the BOP to remove solenoid fuses per Attachment 7.1. Once fuses are removed the SRV closes. Entry into PPM 5.2.1 (Primary Containment Control) will be required if Suppression Pool level exceeds +2 inches or wetwell temperature exceeds 90°F.

Event 7 (TRG-7) A primary leak from the RRC-P-1B suction line occurs. The crew takes actions to identify and isolate the leak per ABN-LEAK which will not be successful. The leak continues to increase until degrading plant parameters require a manual reactor scram. The crew takes actions per PPM 3.3.1 (Reactor Scram), PPM 5.1.1 (RPV Control), and PPM 5.2.1 (Primary Containment Control). The crew initiates Wetwell sprays when Wetwell pressure reaches 2 psig and initiates Drywell sprays when Wetwell pressure exceeds 12 psig but prior to exceeding the Pressure Suppression Pressure (PSP) limit (PPM 5.2.1 Figure F) and after verifying Drywell parameters are within the Drywell Spray Initiation Limit (DSIL)

(PPM 5.2.1 Figure E) and RHR is NOT required for adequate core cooling (CT #1). RHR will be re-aligned from Drywell spray to LPCI injection after emergency depressurization is initiated. Due to a loss of sufficient RPV injection, RPV level continues to lower requiring the crew to emergency depressurize the RPV because sufficient high pressure injections system are not available.

Event 8 Total loss of feedwater injection occurs: Reactor Feedwater Flow Indicating Controller (RFW-FIC-620) output fails low and FWH 6A/6B Bypass Valve (RFW-V-109) fails to open preventing RFW injection into the RPV. RFW-HX-6A & B Discharge to Rx Discharge MOVs (RFW-V-112A & B) fail to open (if attempted) after being initially closed to support feeding with the RFW Flow Control Valves (RFW-FCV-10A/B).

Examiner Note: If the ATC operator fails to close either RFW-V-112A or RFW-V-112B then with specific Examiner direction, Trigger 11 will be entered to cause a trip of all running Condensate Booster pumps to ensure a total loss of feedwater injection occurs which is needed to support Critical Tasks.

Event 9 Reactor Core Isolation Cooling Flow Indicating Controller (RCIC-FIC-600) fails low on RCIC system startup requiring a manual trip of the RCIC turbine.

NRC Scenario 1 Page 5 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Event 10 With insufficient high pressure injection sources available, and with RPV level continuing to lower, the CRS enters PPM 5.1.3 (Emergency RPV Depressurization) and initiates emergency depressurization by opening seven (7) Safety Relief Valves (ADS preferred) after RPV water level reaches TAF (-161 inches) and within 10 minutes of level dropping below TAF. (CT #2) After ED, and within 10 minutes of RPV pressure lowering to 200 psig, the crew will restore and maintain RPV water level above TAF (-161 inches) using Low Pressure ECCS systems. (CT #3) Wetwell and Drywell sprays can be reinitiated per PPM 5.2.1 when not needed for adequate core cooling.

TERMINATION CRITERIA: The scenario will be terminated when Drywell sprays have been initiated, an Emergency Depressurization has been performed and RPV level is being controlled in the prescribed band OR as directed by the Examination Team.

NRC Scenario 1 Page 6 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Critical Task Determination Measurable Performance Critical Task Safety Significance Cueing Performance Feedback Indicators CT #1 - Initiate Drywell Primary containment Procedural direction The operator Valve position will sprays when Wetwell pressures at or above in PPM 5.2.1 (Primary will manually change and Drywell pressure exceeds 12 specified limits pose a Containment Control - open Drywell spray flow will psig but prior to direct threat to primary step P-7) when spray isolation increase.

exceeding PSP, after containment integrity Wetwell pressure valves.

verifying Drywell and the pressure exceeds 12 psig.

parameters are within suppression function.

DSIL and RHR is NOT required for adequate (Ref: PPM 13.1.1A core cooling. (Classifying the Emergency - Technical Bases) Attachment 4.1 section 3)

CT #2 - Initiate Preclude core damage Procedural direction The operator The valve light Emergency by establishing in PPM 5.1.1 (RPV will manually indications for each Depressurization (ED) conditions that allow Control - step L-15) open 7 Safety of the 7 Safety by opening seven (7) low pressure ECCS when RPV Level Relief Valves Relief Valves will Safety Relief Valves systems to restore cannot be restored (ADS preferred) change from Green (ADS preferred) after water level above TAF and maintained above to emergency lit to Red lit when RPV water level (Safety Limit) -186 inches. depressurize control switch is reaches TAF (-161 the RPV. taken to Open.

inches) and within 10 (Ref: CGS Technical minutes of level Specifications - 2.1.1.3) Reactor pressure dropping below TAF. will lower in response.

CT #3 - After ED, and Preclude core damage Procedural direction All available low Indication of within 10 minutes of by establishing in PPM 5.1.1 (RPV pressure ECCS applicable ECCS RPV pressure lowering conditions that allow Control - step L-16) systems are system flow.

to 200 psig, restore low pressure ECCS which directs aligned to and maintain RPV systems to restore restoring and restore RPV RPV level rises to water level above TAF water level above TAF maintaining RPV level level. greater than TAF.

(-161 inches) using (Safety Limit) above -186 inches Low Pressure ECCS and ultimately above systems. (Ref: CGS Technical TAF.

Specifications - 2.1.1.3)

NRC Scenario 1 Page 7 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SIMULATOR SETUP Unload simulator (between each scenario)

Verify in ILC load Reload simulator Reset to IC-208 (reset, go to Run, reset again)

Load Scenario 1 Schedule file Load Scenario 1 Event file (if not loaded automatically)

Validate that there are no unexpected annunciators or parameters out of band Verify pump running magnets Flag the following:

601.A2 5-7 840.A5 7-1 820.B1 7-2 Have marked up copy of SOP-RHR-SPC for RHR B in Suppression Pool Cooling for the crew at turnover (ensure N/Ad steps are initialed)

NRC Scenario 1 Page 8 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SCHEDULE FILE

1

<ACTION>schedule Schedule\local.sch</ACTION>

</ITEM>

1

<ACTION>Event D:\NRC Scenario Support Files\2017 NRC SC-1.evt</ACTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-FWC011</ACTION>

<DESCRIPTION>STARTUP LVL CNTRLR AUTO DEMAND FAILS AS IS</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOV-CFW043F to FAIL_AS_IS</ACTION>

<DESCRIPTION>RFW-V-109 RFW-HX-6A,6B BYPASS</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction CNH-RCI002B to 0</ACTION>

<DESCRIPTION>RCIC-FIC-600 RCIC FLOW CONTROL OUTPUT</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RWU008</ACTION>

</ITEM>

1

<ACTION>Insert malfunction XMT-RRS106A to 0 in 10 on event 1</ACTION>

<DESCRIPTION>RFW-DPT-4A FIXED OUTPUT REACTOR VESSEL LEVEL A</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction HTX-RCC010F to 1.80000 in 600 on event 2</ACTION>

<DESCRIPTION>RWCU-HX-2 FOULING: RWCU-HX-2 NON-REGENERATIVE HEAT EXCH</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-CRD001 to TRIP on event 3</ACTION>

<DESCRIPTION>CB-CRD-P-1A CRD-P-1A MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

NRC Scenario 1 Page 9 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017

<ACTION>Insert malfunction BKR-CSS001 to FA_CTRL_FUS on event 4</ACTION>

<DESCRIPTION>CB-HPCS-P-1 HPCS-P-1 MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction XMT-RRS036A to 5400.00000 on event 5</ACTION>

<DESCRIPTION>RRC-FT-14A FIXED OUTPUT RECIRC PUMP A FLOW</DESCRIPTION>

</ITEM>

1

<ACTION>Insert override OVR-RRS022D to ON on event 6</ACTION>

<DESCRIPTION>MS-RV-2B SAFETY RELIEF OPEN</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RRS004B to .100000 in 180 on event 7</ACTION>

<DESCRIPTION>RECIRC LINE RUPT- RRC-P-1B SUCT</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RRS004D after 180 to 0.80000 in 300 on event 8</ACTION>

<DESCRIPTION>RECIRC LINE RUPT- RRC-P-1B DISCH</DESCRIPTION>

</ITEM>

<ACTION>Insert malfunction MOV-CFW044F to FAIL_AS_IS on event 9</ACTION>

<DESCRIPTION>RFW-V-112A OUTLET RFW-V-112A/6A HP HTR</DESCRIPTION>

</ITEM>

<ACTION>Insert malfunction MOV-CFW045F to FAIL_AS_IS on event 10</ACTION>

<DESCRIPTION>RFW-V-112B OUTLET RFW-V-112B/6B HP HTR</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-CFW004 to TRIP on event 11</ACTION>

<DESCRIPTION>CB-COND-P-2A COND-P-2A MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-CFW005 to TRIP on event 11</ACTION>

<DESCRIPTION>CB-COND-P-2B COND-P-2B MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-CFW006 to TRIP on event 11</ACTION>

<DESCRIPTION>CB-COND-P-2C COND-P-2C MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

</SCHEDULE>

NRC Scenario 1 Page 10 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT FILE

<EVENT>

</EVENT>

NRC Scenario 1 Page 11 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1

==

Description:==

RFW-DPT-4A fails downscale (Tech Spec)

Event is initiated as directed by the Exam team and is activated using TRIGGER 1.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 1 Time Position Applicants Actions or Behavior ATC Acknowledges annunciator 603.A8.3-7 (RPV LEVEL HIGH/LOW ALERT) and informs CRS Comment:

May recommend entry into ABN-LEVEL and ABN-INSTRUMENTATION Comment:

Examiner Note: Following steps are from ARP 4.603.A8 3-7 (RPV LEVEL HIGH/LOW ALERT)

ATC 1: Checks RFW-LI-606A, B, C

Recognizes and reports that NR A has failed downscale

May select NR B on H13-P603, as A is unreliable (the system automatically selects B for level control since A failed)

Comment:

CRS May enter ABN-LEVEL (only required to be entered on actual RPV level change)

Comment:

Examiner Note: Following step is from ABN-LEVEL CRS 4.1.2: May direct ATC to select B RPV Level channel at H13-P603 Comment:

CRS Enters ABN-INSTRUMENTATION Comment:

Examiner Note: Following steps are from ABN- INSTRUMENTATION CRS 4.3: Directs BOP to check excess flow check valve status indication at H13-P851 (if not already done)

Comment:

NRC Scenario 1 Page 12 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1 (CONTINUED)

Examiner Note: Refer to Simulator Guide Attachments 1, 2 or 3 in reference to ABN Attachments 7.4, 7.5 & 7.6, respectively.

CRS 4.5: Refers to Attachment 7.4 (page 16) or Attachment 7.5 (page 20) and determines associated power supply and Tech Spec reference Comment:

4.6: Refers to Attachment 7.6 (page 22) and determines that detector RFW-DPT-4A feeds level instrument RFW-LI-606A and also notes detector location BOP Checks excess flow check valve status indication at H13-P851 Comment:

Validates that the Feedwater Water Level Control system has automatically shifted RPV level control to channel B Comment:

Dispatches field operator to investigate RFW-DPT-4A in Reactor Building 522 (NW) at instrument rack P004 Comment:

BOOTH ROLEPLAY - If sent to investigate status of RFW-DPT-4A, wait 2 minutes then report Nothing abnormal found with RFW-DPT-4A.

CRS Evaluates Technical Specifications and determines the following action applies:

LCO 3.3.2.2 A.1 - Place affected channel in trip within 7 days Comment:

NRC Scenario 1 Page 13 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2

==

Description:==

RWCU NRHX fouling causes high temperature isolation on RWCU-V-4 Event is initiated after CRS makes Tech Spec call (or as directed by the Exam team) and is activated using TRIGGER 2.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 2 Time Position Applicants Actions or Behavior BOP/ATC Acknowledges annunciator 602.A5 6-8 (CLEANUP FLTR INLET TEMP HIGH) and informs CRS Comment:

Examiner Note: Following steps are from ARP 4.602.A5 6-8 (CLEANUP FLTR INLET TEMP HIGH)

BOP 1: Checks (and continues to monitor) RWCU NRHX Outlet temperature on RWCU-TI-607 (may be done prior to pulling ARP)

Comment:

3: Refers to SOP-RCC-OPS and verifies proper RCC flow to NRHX (requires field support)

Comment:

BOOTH ROLEPLAY - If sent to verify proper RCC flow to NRHX, wait 2 minutes then report RCC flow to NRHX is normal and has not changed during the shift. RCC-V-8 is in the proper throttled position.

BOP 4: Monitors RCC HX Outlet temperature (requires field support)

Comment:

BOOTH ROLEPLAY - If sent to verify proper RCC HX outlet temperature, wait 1 minute then report RCC HX outlet temperature is normal.

BOP 6: Verifies RWCU flow is normal Comment:

7: Dispatches field operators to perform walk-downs and to verify proper RCC and RWCU system alignment Comment:

BOOTH ROLEPLAY - If sent to check for any RCC leakage, wait 10 minutes then report No signs of RCC leakage found.

NRC Scenario 1 Page 14 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2 (CONTINUED)

BOP 8: If temp approaches 140°F and isolation appears imminent (it will):

Stops RWCU-P-1A

Closes RWCU-V-4 (and may close RWCU-V-1)

Verifies closed RWCU-V-44 Comment:

If RWCU automatically isolates (if crew does not respond in time)

Reports RWCU isolation (RWCU-V-4 closed) with failure of RWCU-P-1A to auto trip

Manually trips RWCU-P-1A Comment:

CRS May establish a Key Plant Parameter to isolate RWCU prior to exceeding 140°F on the NRHX outlet Comment:

Directs isolating RWCU per the ARP prior to exceeding 140°F on the NRHX outlet Comment:

May direct RWCU-V-104 be throttled open (per SOP-RWCU-SHUTDOWN, page 7, step 5.1.9)

Comment:

ATC May aid in monitoring temperature while BOP is working through procedures and communicating with field operators Comment:

NRC Scenario 1 Page 15 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3

==

Description:==

CRD-P-1A trips requiring CRD-P-1B to be started Event is initiated after RWCU has been isolated (or as directed by the Exam team) and is activated using TRIGGER 3.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 3 Time Position Applicants Actions or Behavior ATC Acknowledges annunciator 603.A7 3-8 (CRD CHARGE WATER PRESS LOW ) and recognizes that the running CRD pump tripped and informs CRS Comment:

CRS Directs starting CRD-P-1B Comment:

Examiner Note: Following steps are from ARP 4.603.A7 3-8 (CRD CHARGE WATER PRESS LOW )

ATC 1: Checks CRD-PIS-600 (Charging Water Header Pressure at H13-P603) -

will be LT 1300 psig Comment:

2: Determines if either CRD pump is running (neither pump will be running)

Comment:

3a: Places CRD-FC-600 in manual Comment:

3b: Reduces CRD-FC-600 output to zero Comment:

3c: Starts CRD pump 1B Comment:

3d: Nulls CRD-FC-600 Comment:

3e: Transfers CRD-FC-600 to Auto Comment:

NRC Scenario 1 Page 16 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3 (CONTINUED)

ATC 3f: IF necessary, then adjust CRD-V-3 (Drive/Cooling Water Pressure Control) to 255-265 psid on CRD-DPI-602 (Drive HDR/RX P) (may not be necessary)

Comment:

BOP May assist in acknowledging annunciators and monitoring plant parameters while ATC recovers the CRD pump Comment:

Directs field operator to report local control rod accumulator pressures (if directed by CRS)

Comment:

Examiner Note: Depending on the time required to restore CRD flow, one or more control rod HCU accumulator alarms may come in. If asked, local accumulator pressures will be reported to be 980 psig which is below the alarm setpoint but above the LCO 3.1.5 minimum limit of 940 psig required for operability. No Technical Specification actions will be required. Accumulator alarms will clear soon after CRD pump is started.

CRS May direct control rod HCU accumulator pressures be reported from the field Comment:

BOOTH ROLEPLAY - If sent to report control rod HCU accumulator pressures for alarming accumulators, wait 2 minutes then report all accumulator pressures at 980 psig.

CRS Evaluates Technical Specification 3.1.5 (no entry Condition exists)

Comment:

Directs CRD-P-1B to be Protected per PPM 1.3.83 (Protected Equipment Program) Attachment 7.1 (based on CRD-P-1A unavailability)

Comment:

NRC Scenario 1 Page 17 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4

==

Description:==

HPCS-P-1 control power failure (Tech Spec)

Event is initiated after CRD-P-1B started and CRD parameters restored (or as directed by the Exam team) and is activated using TRIGGER 4.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 4 Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 601.A1 6-8 (HPCS OUT OF SERVICE) and recognizes a loss of HPCS pump breaker position indication and informs CRS Comment:

Notes that BISI for CB HPCS OUT OF SERV is lit Comment:

ATC Continues to monitor reactor power, pressure and level Comment:

Examiner Note: Following step is from ARP 4.601.A1 6-8 (HPCS OUT OF SERVICE)

BOP 1: Refers to BISI (CB HPCS OUT OF SERV) (ARP Attachment 1 - Page 4) to determine actions required (see below)

Comment:

Examiner Note: Following steps are from ARP 4.601.A1 6-8 (Attachment 1 - Page 4)

BOP 1: Directs field operator to check status of HPCS Pump breaker (HPCS-CB-P1) and associated breaker control power fuses Comment:

BOOTH ROLEPLAY - If sent to check status of the HPCS Pump breaker, wait 2 minutes then report HPCS Pump breaker is racked in with breaker open. Have loss of local breaker indication.

BOOTH ROLEPLAY - If sent to check status of the HPCS Pump breaker control power fuses, wait 4 minutes then report Both the HPCS breaker close and trip fuses are blown.

BOP 2: Refers CRS to Technical Specification 3.5.1 Comment:

Examiner Note: Following steps are from ARP 4.601.A1 6-8 (HPCS OUT OF SERVICE)

CRS 2: Enter HPCS as Inoperable in the Plant Logging System Comment:

NRC Scenario 1 Page 18 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4 (CONTINUED)

CRS 3: Refers to PPM 1.10.1 (Notifications and Reportable Events) to determine reportability requirements Comment:

Evaluates Technical Specifications and determines the following actions apply:

LCO 3.5.1 B.1 - Immediately verify by administrative means that RCIC is operable Comment:

LCO 3.5.1 B.2 - Restore HPCS system to operable status within 14 days Comment:

LCO 3.5.1 C.1 - Restore RHR-SYS-B or HPCS system to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Comment:

Examiner Note: If HPCS pump control power fuses are replaced they will blow again.

CRS May direct HPCS control power fuses be replaced or removed or request troubleshooting assistance before doing so.

Comment:

BOOTH ROLEPLAY - If directed to replace the HPCS Pump control power fuses, wait 10 minutes then report Replaced the trip and close control power fuses for the HPCS Pump.

Appears the fuses may have blown again.

BOOTH ROLEPLAY - If directed to remove the HPCS Pump control power fuses, wait 5 minutes then report Control power fuses for the HPCS Pump have been removed.

CRS Directs the following systems to be Protected per PPM 1.3.83 (Protected Equipment Program) Attachment 7.1 (based on HPCS unavailability)

RCIC-P-1

LPCS-P-1

DG-SYS-A

DG-SYS-B

ADS-SYS-A

ADS-SYS-B

SW-SYS-A

SW-SYS-B Comment:

NRC Scenario 1 Page 19 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5

==

Description:==

RRC-FT-14A fails low causing APRM-CHS-1 to trip Event is initiated after CRS makes Tech Spec call (or as directed by the Exam team) and is activated using TRIGGER 5.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 5 Time Position Applicants Actions or Behavior ATC Acknowledges annunciator 603.A8 3-6 (FLOW REFERENCE OFF NORMAL) and informs CRS Comment:

Examiner Note: Following steps are from ARP 4.603.A8 3-6 (FLOW REFERENCE OFF NORMAL)

ATC 1: Determine which channel is causing the alarm by checking RBM ODA (H13-P603) or RBM chassis (H13-P608) (BOP will have to check P608)

Comment:

BOP Assists ATC with initial diagnosis at H13-P603 Comment:

May investigates RBM chassis at H13-P608 (as a backup to H13-P603 indications)

Comment:

ATC 2: If CRS directs, bypasses failed channel (APRM A) at H13-P603 -

Annunciator clears Comment:

3: Refers CRS to Technical Specification 3.3.1.1 and LCS 1.3.2.1 Comment:

CRS Directs bypassing APRM A Comment:

Evaluates Technical Specification 3.3.1.1 and LCS 1.3.2.1 and determines the minimum number of required APRMs remain operable and that no Technical Specification or LCS actions apply.

Comment:

NRC Scenario 1 Page 20 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6

==

Description:==

SRV MS-RV-2B inadvertently opens (will close upon fuse removal)

Event is initiated after APRM A is bypassed and Tech Specs have been evaluated (or as directed by the Exam team) and is activated using TRIGGER 6.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 6 Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 601.A2 5-8 (SRV OPEN) and informs CRS Comment:

Examiner Note: Following step is from ARP 4.601.A2 5-8 (SRV OPEN)

BOP 1: Refers CRS to ABN-SRV Comment:

CRS Enters ABN-SRV and directs subsequent actions Comment:

Examiner Note: Following steps are from ABN-SRV CRS 4.1: May establish a Key Plant Parameter of Suppression Pool temperature of less than 110°F (not expected to reach)

Comment:

BOP 4.2: Verifies SRV MS-RV-2B is open by one or more of the following:

Rising tailpipe temperature (H13-P614)

Rising Suppression Pool temperature or level

Reduction in Main Generator output (approx. 70 MWe)

Comment:

NRC Scenario 1 Page 21 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6 (CONTINUED)

Examiner Note: Following three steps are required since reactor power is > 90 percent.

BOP 4.4.1: Places control switch for SRV MS-RV-2B to Open Comment:

ATC 4.4.2: Reduces reactor power to < 90% using RRC flow Comment:

BOP 4.4.3: Places control switch for SRV MS-RV-2B to Off Comment:

Examiner Note: SRV remains open requiring removal of fuses.

CRS 4.6: Directs removal of SRV fuses for SRV MS-RV-2B per Attachment 7.1 Comment:

Examiner Note: BOP should remove badge, rings, and conductive materials and don protective eye-wear (ISPM-20 or ISPM-7 for electrical safety).

Examiner Note: Refer to Simulator Guide Attachments 4 in reference to ABN Attachments 7.1.

BOP May update crew that SRV fuses are going to be pulled for SRV MS-RV-2B Comment:

ATC Monitors for a change in SRV position status while fuses are pulled Comment:

Examiner Note: Applicants are trained to leave fuses on floor just outside the cabinet.

BOP 4.6: Removes fuses (using fusepullers) listed on ABN-SRV (Attachment 7.1) (Fuses BB-F29 and BB-F30 will be removed from Panel H13-P628)

Comment:

CRS May update crew that SRV is closed Comment:

Enters PPM 5.2.1 (Primary Containment Control) if wetwell level exceeds +2 inches (which corresponds to Tech Spec limit of 31 ft 1.75 inches)

Comment:

NRC Scenario 1 Page 22 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6 (CONTINUED)

Examiner Note: Already in Suppression Pool cooling.

CRS Enters (or re-enters) PPM 5.2.1 (Primary Containment Control) if wetwell temperature exceeds 90°F Comment:

Examiner Note: The only applicable Technical Specifications are those related to high wetwell level or high wetwell temperature coincident with EOP entry, if they should occur. There are no applicable TS actions associated with the faulty non-ADS SRV (tracking only).

CRS 4.9: Evaluates Technical Specifications and determines the following actions apply:

SRV MS-RV-2B: NONE Comment:

High wetwell level > 31 feet 1.75 inches: LCO 3.6.2.2 A.1 - Restore Suppression Pool water level to within limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Comment:

High wetwell temperature > 90°F: LCO 3.6.2.1 A.1 - Verify Suppression Pool average temperature is 110°F once per hour AND Restore Suppression Pool average temperature to 90°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Comment:

4.10: May discuss need to perform OSP-CVB/IST-M701 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of SRV opening Comment:

4.11: May discuss need to initiate Condition Report to evaluate reactivity event per PPM 1.3.79 Comment:

NRC Scenario 1 Page 23 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7

==

Description:==

LOCA from RRC-P-1B suction line requiring manual scram Event is initiated after fuses have been removed for SRV MS-RV-2B and associated Tech Specs evaluated (or as directed by the Exam team) and is activated using TRIGGER 7.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 7 Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 601.A3 6-5 (LEAK DETECTION DRYWELL FLOOR DRAIN FLOW HIGH) and observes rising Drywell pressure Comment:

Reports indications of primary leak to CRS Comment:

Pulls up DSIL curve on GDS to check for excessive Drywell pressure for given Drywell temperature (curve slopes to the right)

Comment:

Examiner Note: Following steps are from ARP 4.601.A3 6-5 (LEAK DETECTION DRYWELL FLOOR DRAIN FLOW HIGH)

BOP 1: May check Drywell Floor Drain flow GE 5 GPM as read on EDR-FRS-623 (already have evidence of significant leak)

Comment:

3: Refers CRS to ABN-LEAKAGE Comment:

4: Continues to monitor containment parameters due to RCS leakage Comment:

5: May refer CRS to Technical Specification 3.4.5 Comment:

CRS Enters ABN-LEAKAGE Comment:

NRC Scenario 1 Page 24 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7 (CONTINUED)

Examiner Note: Following steps are from ABN-LEAKAGE Examiner Note: ABN assumes a smaller initial leak rate which can be diagnosed over time.

Only relevant actions will be performed.

BOP 4.1.3: Monitors Containment radiation monitors at RAD Board 22 and 23 (may not get to this)

Comment:

4.1.4: Monitors Drywell temperature and pressure (in progress)

Comment:

CRS 4.1.9: Directs ROs to investigate source of leak and isolate if possible (unisolable)

Comment:

Examiner Note: CRS will direct manual scram before automatic high Drywell pressure scram occurs. If time permits RRC flow may be reduced to 74Mlbm/hr before scram inserted.

CRS Updates crew and directs ATC to scram the reactor Comment:

Examiner Note: Following steps are Immediate Actions from PPM 3.3.1 (Reactor Scram)

ATC 6.1.1: Places Reactor Mode Switch to Shutdown Comment:

6.1.2: Monitors reactor power, pressure and level Comment:

6.1.5: Inserts SRM and IRM monitors (detectors)

Comment:

After above three steps ATC makes scram report to CRS:

Mode switch is in Shutdown

APRMs are downscale

RPV pressure is (value and trend)

RPV level is (value and trend)

EOP entry on low RPV level (and possibly high Drywell pressure)

Comment:

NRC Scenario 1 Page 25 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7 (CONTINUED)

CRS Repeats back scram report Comment:

ATC 6.1.6: After CRS repeat back, reports all control rods are IN Comment:

CRS Enters PPM 5.1.1 (RPV Control) on low RPV level (+13 inches)

Comment:

Enters PPM 5.2.1 (Primary Containment Control) and re-enters PPM 5.1.1 on high Drywell pressure (1.68 psig)

Comment:

Examiner Note: Following steps are Subsequent Actions from PPM 3.3.1 (Reactor Scram)

ATC 6.2.5.a: Verify Recirc pumps have run back to 15 Hz Comment:

6.2.6: Range down on IRMs, as necessary, to follow power decrease Comment:

BOP 6.2.7: Make PA announcement for reactor scram Comment:

Examiner Note: See Event 8 for feedwater actions per SOP-RFW-FCV-QC quick card ATC 6.2.8: Transfers level control to RFW-FCV-10A/B per SOP-RFW-FCV-QC Comment:

BOP 6.2.9: If necessary (with Main Generator load < 50 MWe):

If Main Turbine did not trip - simultaneously depress both Emerg Trip pushbuttons (H13-P820)

If Main Generator did not trip -depress either Unit Emergency Tip pushbutton or Unit Overall Trip pushbutton (H13-P800)

Verify power transfer to Startup Transformer (TR-S)

Comment:

CRS Directs 1.68 psig actuations be verified Comment:

NRC Scenario 1 Page 26 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7 (CONTINUED)

BOP Verifies 1.68 psig actuations - Observes:

All ECCS pumps started (except for HPCS) and min flow valves opened

Both service water pumps (SW-P-1A & B) started

EDG-1 & EDG-2 running

GDS status for containment isolation valve closure (no yellowed border NSSSS groups indicated)

Reports actuations verified to CRS Comment:

CRS Works down the Primary Containment Pressure leg of PPM 5.2.1 (RPV Containment Control) and sets a Key Plant Parameter of 2 psig Wetwell pressure Comment:

BOP Reports when Wetwell pressure reaches 2 psig Examiner Note: With RHR System B already in Suppression Pool Cooling, CRS should use RHR loop B for Wetwell sprays and (later on) RHR loop A for Drywell sprays; otherwise, RHR pump runout may occur with Drywell Sprays and Suppression Pool Cooling operating from the same loop.

CRS Directs Wetwell Spray using RHR B (preferred) or A spray loops Comment:

BOP Refers to SOP-RHR-SPRAY-WW-QC quick card to initiate Wetwell Sprays:

2.1.1: Verify RHR-P-2A(B) running

2.1.2: Verify RHR-V-42A(B) closed (LPCI injection valve)

2.1.3: Open RHR-V-27A(B) (Suppression Pool Spray valve)

2.1.4: Before Wetwell Spray drops below 0.0 psig, or when directed by the CRS, then close RHR-V-27A(B)

Comment:

CRS Directs Wetwell Sprays be secured prior to Wetwell pressure reaching 0.0 psig Comment:

NRC Scenario 1 Page 27 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7 (CONTINUED)

CRS Works down the Primary Containment Pressure leg of PPM 5.2.1 (RPV Containment Control) and sets Key Plant Parameter of 12 psig in the Wetwell Comment:

Works down the Drywell Temperature leg of PPM 5.2.1 (RPV Containment Control) and sets Key Plant Parameter of 285 °F in the Drywell (not expected to be reached during scenario)

Comment:

BOP Reports Wetwell pressure at 12 psig Comment:

CT #1 - Initiate Drywell sprays when Wetwell pressure exceeds 12 psig but prior to exceeding PSP, after verifying Drywell parameters are within DSIL and RHR is NOT required for adequate core cooling.

CRS Directs BOP to verify within DSIL (Drywell Spray Initiation Limit - Fig. E on PPM 5.2.1 (Primary Containment Control)) (May verify themself)

Comment:

BOP Reports Drywell parameters within DSIL Comment:

CRS Verifies RHR-P-2A not currently needed to ensure Adequate Core Cooling Comment:

Directs RRC pumps be verified off and Drywell Cooling fans be secured Comment:

BOP Verifies RRC pumps off and secures the Drywell Cooling fans on back panel (bottom row of containment fans with switches that are not in the brown area on the panel)

Reports completion to CRS Comment:

NRC Scenario 1 Page 28 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7 (CONTINUED)

CRS Directs Drywell Sprays (should be initiated on opposite loop that Wetwell Sprays are on Comment:

BOP Refers to SOP-RHR-SPRAY-DW-QC quick card:

2.1.1: Verify RHR-P-2A(B) is running

2.1.2: Verify RHR-V-42A(B) closed (LPCI injection valve)

2.1.3: Open the following to spray the Drywell:

RHR-V-17A(B) (Drywell Spray Inboard Isolation)

RHR-V-16A(B) (Drywell Spray Outboard Isolation)

2.1.4: Before Drywell pressure drops below 0.0 psig, or when directed by the CRS, then close the following:

RHR-V-16A(B)

RHR-V-17A(B)

Comment:

CRS Directs Drywell Sprays be secured before Drywell pressure drops to zero psig Comment:

BOP/ATC Reports Main Steam Tunnel temperature alarms When MSIVs close: Updates crew that MSIVs are closed and pressure control is with SRVs at 800 to 1050 psig (or current pressure band)

Comment:

CRS May direct RPV pressure reduction to a band of 500 to 600 psig (but not expected in order to conserve inventory)

Comment:

NRC Scenario 1 Page 29 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7 (CONTINUED)

BOP Lowers RPV pressure if directed using SOP-DEH-QC (Main Turbine DEH Operations Quick Card):

2.1.1a: Selects PRESSURE TARGET

2.1.1b: Enters desired pressure

2.1.1c: Selects OK

2.1.1.d: If change in pressure rate is desired:

1: Selects PRESSURE RATE 2: Enters desired PRESSURE RATE 3: Selects OK

2.1.1.e: Selects GO

2.1.1.f: Selects YES

2.1.1.g: Verifies pressure demand and throttle pressure change at the pressure rate.

Comment:

NRC Scenario 1 Page 30 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 8

==

Description:==

RFW-FIC-620 controller failure with RFW-V-109 failing to open and RFW-V-112A & B failing to open once closed Event is activated at the beginning of the scenario and is realized when ability to feed reactor with feedwater system has been lost.

Time Position Applicants Actions or Behavior Examiner Note: Following steps are from SOP-RFW-FCV-QC (Transfer RPV Level Control to RFW-FCV-10A/10B - Quick Card).

Examiner Note: If the ATC operator fails to close either RFW-V-112A or RFW-V-112B below (i.e.

one or both valves remain open) then direct insertion of Trigger 11 to cause a trip of all running Condensate Booster pumps to ensure a total loss of feedwater injection occurs which is needed to support Critical Tasks.

ATC 2.1.1: (2-handed operation) Starts closing RFW-V-112A and RFW-V-112B Comment:

2.1.2: Starts opening RFW-V-118 Comment:

2.1.3: Verifies RFW-V-109 is closed Comment:

2.1.4: (2-handed operation) Verifies RFW-V-117A and RFW-V-117B open Comment:

2.1.5: Verifies RFW-LIC-620 is in Manual (V selected for Valve position demand with 0 output)

Comment:

2.1.6: IF Reactor Feed Pump(s) (RFP) are operating, then performs the following:

2.1.6.a: Verifies RFPs have ramped down in speed

2.1.6.b: Places RFW-P-1B in MDEM mode

2.1.6.c: Places RFW-P-1B in MDEM mode

2.1.6.d: Controls turbine speed as required

2.1.6.e: If desired, then places RFW-FCV-2A(B) in Manual and slowly open to approximately 80%

Comment:

NRC Scenario 1 Page 31 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 8 (CONTINUED)

Examiner Note: RFW-HX-6A & B Discharge to Rx Discharge MOVs (RFW-V-112A & B) will fail to open (if attempted) after being closed below.

ATC 2.1.7: Verifies RFW-V-112A and RFW-V-112B are fully closed Comment:

2.1.8: Verifies RFW-V-118 is fully open Comment:

2.1.9: IF Reactor Feed Pump(s) (RFP) are operating, then adjusts the running RFP speed to establish ~ 200 psid across RFW-FCV-10A & 10B using either Feedwater touch screen (H13-P840)

Comment:

Examiner Note: Controller failure will not allow step 2.1.10 below to be performed.

ATC 2.1.10: Adjusts RFW-LIC-620 manual output to control RPV level - Will be unsuccessful Comment:

Examiner Note: For step below RFW-V-109 fails to open and RFW-V-118 is already fully open.

ATC 2.1.12: If unable to control RPV level with RFW-FCV-10A/B, then considers throttling RFW-V-109 or RFW-V-118 to control RPV level Comment:

Reports to CRS existing faults with feedwater and the inability to feed Comment:

NRC Scenario 1 Page 32 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 9

==

Description:==

RCIC-FIC-600 fails low on startup requiring manual trip of RCIC turbine Event is activated at the beginning of the scenario and is realized when RCIC system is started manually or automatically starts on low RPV level (-50 inches)

Time Position Applicants Actions or Behavior Examiner Note: Indications of RCIC controller failure will be the same whether RCIC is started manually below or RCIC started automatically on low RPV level (- 50 inches)

CRS (If not already running) Directs manual start of RCIC for RPV level control Comment:

BOP Refers to SOP-RCIC-INJECTION-QC quick card:

2.1.1.a: Verifies the RCIC Manual Initiation Pushbutton in Armed

2.1.1.b: Depresses and hold the RCIC Manual Initiation pushbutton

2.1.1.c: When all applicable RCIC valves have repositioned, then releases the RCIC Manual Initiation pushbutton Comment:

Recognizes RCIC turbine speed oscillating below minimum requirement of 2100 RPM (with no RPV injection flow) and that the RCIC controller (RCIC-FIC-600) output is zero Comment:

Shifts RCIC controller (RCIC-FIC-600) to Manual and presses the right OPEN pushbutton in an attempt to raise controller output (RCIC Turbine speed) - Will be unsuccessful Reports RCIC controller problem (and inability to inject with RCIC) to CRS Comment:

NRC Scenario 1 Page 33 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 9 (CONTINUED)

CRS May direct trip of RCIC turbine based on above report Comment:

Examiner Note: ATC may trip RCIC turbine based on direction from CRS or after recommending to CRS in which case RCIC ARPs may not be immediately addressed ATC Acknowledges annunciator 603.A4 1-4 (RCIC TURBINE BEARING OIL PRESSURE LOW) and informs CRS Comment:

Examiner Note: Following steps are from ARP 4.603.A4 1-4 (RCIC TURBINE BEARING OIL PRESSURE LOW )

Examiner Note: In first step below, it is expected that RCIC will still be tripped even if needed for inventory control since using RCIC for inventory control is not currently possible.

BOP 1: If not required for inventory control, then trip RCIC-DT-1 manually (RCIC Turbine)

Comment:

2: Verify RCIC-V-46 is closed Comment:

CRS May inform Security of the unavailability of RCIC system Comment:

BOOTH ROLEPLAY - If sent to investigate status of RCIC system locally, wait until RCIC has been tripped, then report RCIC is not running and nothing abnormal was found.

NRC Scenario 1 Page 34 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 10

==

Description:==

Initiate Emergency Depressurization (ED) on low RPV level and restore RPV level to above TAF Time Position Applicants Actions or Behavior CRS Works down the level leg of PPM 5.1.1 (RPV Control) and recognizes that CRD and SLC are the only high pressure injection sources available Comment:

Directs second CRD pump be started to maximize RPV injection with CRD with a RPV level band of -50 inches to +54 inches Comment:

ATC Direct field operator perform ABN-CRD-MAXFLOW to facilitate starting a second CRD pump Comment:

BOOTH ROLEPLAY - If sent to perform field actions for ABN-CRD-MAXFLOW, insert Trigger 26 and wait 5 minutes, then report Field actions for ABN-CRD-MAXFLOW are complete.

ATC Completes MCR actions per ABN-CRD-MAXFLOW:

4.8.1: Place CRD-FC-600 in Manual

4.8.2: Start the second CRD pump to have both pumps in service

4.8.3: Adjust CRD-FC-600 to throttle open CRD-FCV-2A(2B)

4.8.4: Throttle opens CRD-V-3 to maximize flow to the RPV Comment:

Gives RPV level reports as level continues to lower Comment:

CRS Directs SLC initiation Comment:

Directs ADS be inhibited when ADS Timers initiate Comment:

Expands level band as RPV level drops Comment:

NRC Scenario 1 Page 35 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 10 (Continued)

CRS Directs BOP to verify containment isolations as RPV level lowers to -50 inches and again at -129 inches Comment:

BOP Verifies the following containment isolation valves closed at -50 inches (as seen at the Isolation Control System panel or on GDS):

EDR-V-19 / EDR-V-20

FDR-V-3 / FDR-V-4

RHR-V-49 / RHR-V-40

RHR-V-9 / RHR-V-8

RWCU-V-1 / RWCU-V-4

RRC-V-19 / RRC-V-20

RHR-V-60A / RHR-V-75A

RHR-V-60B / RHR-V-75B

TIP isolation valves BOP Verifies the following ADDITIONAL containment isolation valves closed at

-129 inches (as seen at the Isolation Control System panel or on GDS):

MS-V-22A / MS-V-28A

MS-V-22B / MS-V-28B

MS-V-22C / MS-V-28C

MS-V-22D / MS-V-28D

MS-V-67A

MS-V-67B

MS-V-67C

MS-V-67D

MS-V-16

MS-V-19 Comment:

NRC Scenario 1 Page 36 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 10 ATC Initiates SLC as directed - Refers to SOP-SLC-INJECTION-QC quick card:

2.1: Removes the SLC keylock switch blanks and inserts both keys into the SLC system control switches

2.2: Initiates SLC injection by performing the following (H13-P603):

Places SLC System A control switch to the OPER position Places SLC System B control switch to the OPER position

2.3: Records the following:

SLC flowrate (~43 gpm for one pump or ~86 gpm for both)

Initial tank level Circles RWCU-V-4 status (should be closed)

2.4: Reports one of the following, or similar words, to the CRS as they hand the CRS the procedure:

SLC is injecting normally SLC is partially injecting SLC failed to inject Reports initial tank level of 4800 gallons and that SLC flowrate is 86 gpm Comment:

BOP/ATC When RPV level drops to -129 inches and the ADS Timers initiate, inhibits ADS Comment:

Reports ADS inhibited to CRS Comment:

Reports RPV level as it transitions from Wide Range to Fuel Zone Comment:

Reports RPV level at TAF and trending down Comment:

CRS Determines that Emergency Depressurization (ED) is required when RPV level cannot be maintained > -161 inches Comment:

NRC Scenario 1 Page 37 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 10 (CONTINUED)

CT #2 - Initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS preferred) after RPV water level reaches TAF (-161 inches) and within 10 minutes of level dropping below TAF.

CRS Updates crew and exits the pressure leg of PPM 5.1.1 (RPV Control) via override and enters PPM 5.1.3 (Emergency RPV Depressurization)

Comment:

Determines that with high Drywell pressure signal sealed in, low pressure ECCS systems will be required to maintain Adequate Core Cooling (and therefore will not be stopped and prevented)

Comment:

Determines Wetwell level is above 17 feet Comment:

Directs 7 SRVs be opened (ADS preferred) (ADS SRVs are those with the red stripe on left side of their nameplate)

Comment:

BOP Opens 7 SRVs (ADS preferred) as directed while verifying proper containment response as each is opened and reports completion to CRS Comment:

CRS Directs Wetwell and Drywell sprays and Suppression Pool Cooling be secured to maximize RPV injection Comment:

BOP When directed, refers to SOP-RHR-SPRAY-WW-QC quick card to secure Wetwell Sprays:

2.1.4: Closes RHR-V-27A(B)

Comment:

When directed, refers to SOP-RHR-SPRAY-DW-QC quick card to secure Drywell Sprays:

2.1.4:

Closes RHR-V-16A(B)

Closes RHR-V-17A(B)

Comment:

NRC Scenario 1 Page 38 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 10 (CONTINUED)

BOP When directed secures Suppression Pool Cooling (on RHR B loop) by closing RHR-V-24B Comment:

CT #3 - After ED, and within 10 minutes of RPV pressure lowering to 200 psig, restore and maintain RPV water level above TAF (-161 inches) using Low Pressure ECCS systems.

BOP Allows ECCS injection valves to automatically open at 470 psig Comment:

Reports RPV injection as it occurs, when level is rising, and again when level is restored above TAF (-161 inches)

Comment:

CRS When below TAF, maximizes RPV injection with all available systems (requiring securing of all Sprays and Suppression Pool Cooling)

Comment:

When above TAF, provided enough injection available, directs re-initiation of Wetwell and Drywell sprays and Suppression Pool Cooling with RHR as appropriate (Wetwell Spray initiation if Wetwell pressure reaches 2 psig and Drywell Spray initiation if Wetwell pressure exceeds 12 psig)

Comment:

BOP/ATC Secures injection systems as directed to return RPV level to -50 inches to

+54 inches band Comment:

BOP Reinitiates Wetwell and Drywell Sprays as appropriate using quick cards Comment:

Refers to SOP-RHR-SPRAY-WW-QC quick card to initiate Wetwell Sprays:

2.1.1: Verify RHR-P-2A(B) running

2.1.2: Verify RHR-V-42A(B) closed (LPCI injection valve)

2.1.3: Open RHR-V-27A(B) (Suppression Pool Spray valve)

2.1.4: Before Wetwell Spray drops below 0.0 psig, or when directed by the CRS, then closes RHR-V-27A(B)

Comment:

NRC Scenario 1 Page 39 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 10 (CONTINUED)

CRS Verifies Drywell parameters within DSIL Comment:

Verifies RHR-P-2A(B) not currently needed to ensure Adequate Core Cooling Comment:

Directs Drywell Sprays (should be initiated on opposite loop that Wetwell Sprays are on Comment:

BOP Refers to SOP-RHR-SPRAY-DW-QC quick card:

2.1.1: Verify RHR-P-2A(B) is running

2.1.2: Verify RHR-V-42A(B) closed (LPCI injection valve)

2.1.3: Open the following to spray the Drywell:

RHR-V-17A(B) (Drywell Spray Inboard Isolation)

RHR-V-16A(B) (Drywell Spray Outboard Isolation)

2.1.4: Before Drywell pressure drops below 0.0 psig, or when directed by the CRS, then close the following:

RHR-V-16A(B)

RHR-V-17A(B)

Comment:

CRS Directs Drywell Sprays be secured before Drywell pressure drops to zero psig Comment:

TERMINATION CRITERIA: The scenario will be terminated when Drywell sprays have been initiated, an Emergency Depressurization has been performed and RPV level is being controlled in the prescribed band OR as directed by the Examination Team.

NRC Scenario 1 Page 40 of 45

ABN-INSTRUMENTATION (ATTACHMENT 7.4 - Page 16)

POWER SUPPLY VERSUS CONTROL ROOM RPV LEVEL/PRESSURE INSTRUMENTS Power Supply Instrument Tech Specs E-PP-7AA/1 MS-LI-612 (Indicator power) N/A MS-LR-615 (Recorder) 3.3.3.1.F2b MS-PI-9 3.3.3.1.F1 MS-LR/PR-623A (Recorder) 3.3.3.1.F2a 3.3.3.1.F1 E-PP-7AA/2 MS-LR/PR-623A (transmitter) 3.3.1.1.F3 3.4.12 E-PP-7AA/4 MS-LR-615 (transmitter) 3.3.3.1.F2b RFW-LI-606A (+) 3.3.2.2 RFW-LI-606B (+)

RFW-PI-605 (+) 3.3.3.1.F1 3.3.1.1.F3 3.4.12 MS-PR-609 (transmitter) 3.3.3.1.F1 3.3.1.1.F3 3.4.12 RFW-LR-608 (Recorder) N/A E-PP-US/5/1 RFW-LI-606A (+) 3.3.2.2 RFW-LI-606B (+)

RFW-PI-605 (+) 3.3.3.1.F1 3.3.1.1.F3 3.4.12 E-PP-8AA/6 MS-LI-612 (transmitter) N/A MS-LI-610 3.3.3.1.F2 E-PP-8AA/8 MS-LR/PR-623B 3.3.3.1.F2a (Transmitter and Recorder) 3.3.3.1.F1 3.3.1.1.F3 3.4.12

(+) Powered via a redundant power supply ATTACHMENT 1 Page 41 of 45

ABN-INSTRUMENTATION (ATTACHMENT 7.5 - Page 20)

Power Supply Instrument Function Tech Specs E-DP-S1/HPCS/D7 MS-LIS-100A HPCS, level 8 3.3.5.1.F3c MS-LIS-100B MS-LIS-31A HPCS, level 2 3.3.5.1.F3a MS-LIS-31B MS-LIS-31C MS-LIS-31D E-PP-7AA/4 or RFW-DPT-4A Turbine trip, level 8 3.3.2.2 E-PP-US/5/1 E-DISC-DPS11A/3 RFW-DPT-4C ATTACHMENT 2 Page 42 of 45

ABN-INSTRUMENTATION (ATTACHMENT 7.6 - Page 22)

RFW-DPT-4A RFW-LI-606A NR 'A' level H13-P603 FWLC, RFP/MT L8 TRIP, PP-7AA & 3.3.2.2 FW/MT L8 TRIP P004, RB 522 NW RRC L4 / L3 PP-US/5 RFW-LR-608 NR 'A' level H13-P603 N/A RFW-DPT-4B RFW-LI-606B NR 'B' level H13-P603 FWLC, RFP/MT L8 TRIP, DP-S1-2A 3.3.2.2 FW/MT L8 TRIP P027, RB 522 SW RRC L4 / L3 RFW-LR-608 NR 'B' level H13-P603 N/A RFW-DPT-4C RFW-LI-606C NR 'C' level H13-P603 FWLC, RFP/MT L8 TRIP, DP-S1-1A 3.3.2.2 FW/MT L8 TRIP P005, RB 522 SE RRC L4 / L3 RFW-LI-606D NR 'C' level H13-P602 N/A MS-DPT-32 MS-DPR/FR-613 Core dP/Total H13-P603 None PP-8AZ 3.4.2 Jet Pumps P009, RB 471 NW Flow MS-PT-5 MS-PI-605 RPV Press H13-P603 None PP-7AA N/A P004, RB522 NW MS-LT-26C MS-LI-604 WR Level H13-P603 None PP-8AZ N/A P005, RB 522 SE RFW-DPT-17 RFW-LR-608 UR Level H13-P603 None PP-7AA N/A P027, RB 522 SW MS-PT-808 MS-PR/FR-609 RPV Press H13-P603 None PP-7AA/US/5 N/A P004, RB 522 NW MS-PT-7 MS-PR/FR-609 Turbine Steam H13-P603 None PP-7AA/US/5 N/A IR-10, TB 471 E Flow RFW-DPT-803A RFW-FI-603A MSL FLOW A H13-P603 None PP-7AA/US/5 N/A P015, RB 501 NW RFW-DPT-803B RFW-FI-603B MSL FLOW B H13-P603 None PP-7AA/US/5 N/A P015, RB 501 NW RFW-DPT-803C RFW-FI-603C MSL FLOW C H13-P603 None PP-7AA/US/5 N/A P025, RB 501 SE RFW-DPT-803D RFW-FI-603D MSL FLOW D H13-P603 None PP-7AA/US/5 N/A P025, RB 501 SE MS-LIS-24A None NR Level None RPS L3, NS4 Gp 5&6 RPS-PP-C72 3.3.1.1 RPS P004, RB 522 NW 3.3.6.1 PC ISOL MS-LIS-24B None NR Level None RPS L3, NS4 Gp 5&6, RPS-PP-C72 3.3.1.1 RPS P026, RB 522 NE RCIC L8 3.3.5.2 RCIC INST 3.3.6.1 PC ISOL ATTACHMENT 3 Page 43 of 45

ABN-SRV (ATTACHMENT 7.1 - Page 9)

SRV FUSE LIST SRV SOLENOID FUSE PANEL BB-F35 MS-RV-1A C BB-F36 H13-P628 BB-F27 MS-RV-1B C BB-F28 H13-P628 BB-F17 MS-RV-1C C BB-F18 H13-P628 BB-F37 MS-RV-1D C BB-F38 H13-P628 BB-F19 MS-RV-2A C BB-F20 H13-P628 BB-F29 MS-RV-2B C BB-F30 H13-P628 BB-F25 MS-RV-2C C BB-F26 H13-P628 BB-F23 MS-RV-2D C BB-F24 H13-P628 BB-F21 MS-RV-3A C BB-F22 H13-P628 BB-F33 MS-RV-3B C BB-F34 H13-P628 BB-F31 MS-RV-3C C BB-F32 H13-P628 BB-F15 A BB-F16 H13-P628 BB-F53 C BB-F54 H13-P628 MS-RV-3D AA-F15 B AA-F16 H13-P631 EE-F01 A EE-F02 E-CP-ARS*

BB-F11 A BB-F12 H13-P628 AA-F11 B AA-F12 H13-P631 MS-RV-4A BB-F49 C BB-F50 H13-P628 CC-F29 B CC-F30 C61-P001 ATTACHMENT 4 Page 44 of 45

Appendix D NRC Scenario No. 1 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 TURNOVER Initial Conditions:

Columbia is operating at 100% power

RCIC Operability Test surveillance was just completed to satisfy Post Maintenance Testing (PMT) requirements and has been returned to a Standby status and declared operable

RHR-SYS-B was placed in Suppression Pool Cooling three (3) hours ago to restore Suppression Pool temperature following the testing and to satisfy RHR-P-2B PMT requirements (see marked up procedure)

LCO 3.5.1 A.1, LCO 3.6.1.5 A.1, LCO 3.6.2.3 A.1, and RFO 1.6.1.5 A.1 have been entered for RHR-SYS-B being inoperable Shift Turnover:

Maintain RHR-P-2B in operation for the next three (3) hours to satisfy pump PMT requirements NRC Scenario 1 Page 45 of 45

INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE COLUMBIA GENERATING STATION SIMULATOR EXAMINATION Lower RRC Flow to 90% using Flow (enter GV Sequential Mode); CRD-LESSON TITLE FC-600 Fails High; LPCS-P-2 Trips (TS); MS-PS-23D Fails causing Half Scram (2 Rods Scram but 1 does not Fully Insert - Can Manually Insert)(TS); FPC-P-1B Trip (FPC-P-1A Fails to Auto Start); Trip of E-CB-1/7 with Scram (ATWS) occurring on Auto-Shift to TR-B; Hydraulic ATWS (Lower Level to -140 to -80); Reduced SLC Injection Flow; RWCU-V-4 Fails to Auto Close; Scram-Reset-Scram not Effective in Inserting Rods (Manual Insertion Permitted)

LENGTH OF LESSON 1.5 Hours Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code LO001857 Rev. No. 0 JPM PQD Code Rev. No.

Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/22/16 REVISED BY DATE VALIDATED BY DATE TECHNICAL REVIEW DATE INSTRUCTIONAL REVIEW DATE APPROVED DATE Operations Training Manager NRC Scenario 2 Page 1 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Facility: Columbia Generating Scenario No.: 2 Op Test No.: 1 Station Examiners: Operators:

Columbia is operating at 100% power. Control Rod Drive (CRD) Pump 1B (CRD-P-1B) is out of Initial Conditions:

service for extended Maintenance. CRD-P-1A is Protected.

Lower reactor power to 90% using Reactor Recirculation flow per PPM 3.2.6 (Power Maneuvering) after assuming the shift based on BPA Load Following request. Steps 5.1.1 thru 5.1.6 of PPM 3.2.6 Turnover:

are complete. Proper margin to Pre-Conditioned Status (PCS) exists per PPM 9.3.18. The Reactivity brief has been performed.

Critical Tasks:

During ATWS with power > 5%, terminate and prevent injection with exception of SLC, RCIC, and CRD, into CT-1 the RPV until RPV level is -65 inches to establish a Lowered Level (LL).

Maintain RPV level above -186 inches. Short excursions below -186 inches does not constitute failure of CT CT-2 provided level restored and maintained above -186 inches within 10 minutes of going below -186 inches.

Event Malf. Event Type* Event Description No.

Lower reactor power with Reactor Recirculation (RRC) flow to 90% for R (ATC) 1 N/A load following per PPM 3.2.6 (which includes placing Main Turbine into N (BOP)

Governor Valve Sequential Valve Mode)

CRD Drive Header Flow Control Valve controller (CRD-FC-600) output 2 TRG-2 I (ATC) fails high while in automatic C (BOP,SRO) 3 TRG-3 RHR-SYS-A/LPCS Keep Fill Pump (LPCS-P-2) trips (Tech Spec)

TS (SRO)

Failure of MS-PS-23D which causes a half scram on RPS B side. Two C (ATC,SRO) 4 TRG-4 control rods scram but one does not go full in (must be manually inserted)

TS (SRO)

(Tech Spec) 5 TRG-5 C (BOP) Ground causes FPC-P-1B to spuriously trip (FPC-P-1A fails to auto start)

Trip of E-CB-1/7 with transfer of SM-7 to Backup Transformer resulting in reactor trip signal 6 TRG-6 M (ALL)**

Hydraulic ATWS - Lower RPV Level -80 inches to -140 inches (CT #1)

(CT #2)

SLC-P-1A shaft shears when pump starts and SLC-P-1B develops a 7 N/A N/A discharge flow blockage which limits SLC injection flow.

RWCU-V-4 does not auto close on SLC initiation but can be closed 8 N/A C (ATC) manually.

Scram/Reset/Scram not effective in inserting control rods - Control rods 9 N/A C (BOP) can be manually driven in

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specifications

- Event forms a portion of significant CGS PSA Accident Sequence (TTC044) (Ref: PSA-1-SM-0001 (Rev 7))

NRC Scenario 2 Page 2 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Target Quantitative Attributes Actual Description Reduced SLC injection capability; RWCU-V-4 fails to Malfunctions after EOP entry (1-2) 3 auto close; Scram-reset-scram ineffective CRD-FC-600 failure; LPCS-P-2 shaft seizure; RPS B Abnormal events (2-4) 4 half scram (2 control rods inadvertently scram); FPC-P-1B trip Major transients (1-2) 1 E-CB-1/7 breaker trip leading to hydraulic ATWS EOPs entered/requiring substantive actions (1-2) 1 PPM 5.1.1 (RPV Control)

EOP contingencies requiring substantive actions (0-2) 1 PPM 5.1.2 (RPV Control - ATWS)

EOP based Critical tasks (2-3) 2 See Critical Task Determination table Trigger Evaluator How Purpose Malfunction Numbers (TRG-x) Directed Triggered TRG-2 YES Manually Event Initiator CNH-CRD001E; BST-CRD001F TRG-3 YES Manually Event Initiator PMP-CSS004S BST-RRS067F; MAL-RMC007-3835; MAL-RMC007-1815; TRG-4 YES Manually Event Initiator MAL-RMC005-1815 TRG-5 YES Manually Event Initiator MOT-FPC002G BKR-EPS003; MAL-CRD007A1; MAL-CRD007A2; MAL-CRD007B1; TRG-6 YES Manually Event Initiator MAL-CRD007B2 TRG-7 Manually Field Action BKR-RHR001 TRG-8 Manually Field Action BKR-CSS002 TRG-9 Automatically Malf Trigger BST-CRD001F TRG-10 Automatically Malf Trigger MAL-RMC007-1815; MAL-RMC005-1815 Initial Condition BST-FPC020F Initial Condition PMP-SLC001B Initial Condition BKR-CRD002 Initial Condition PMP-SLC002F Initial Condition MOV-RWU010F NRC Scenario 2 Page 3 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Event 1 The Scenario starts from 100% power with Control Rod Drive (CRD) Pump 1B (CRD-P-1B) out of service for extended maintenance. Once the crew has the shift, the ATC operator lowers reactor power (for load following) using Reactor Recirculation (RRC) flow to 90% per PPM 3.2.6 (Power Maneuvering). The BOP operator takes the Main Turbine out of Governor Valve Optimization mode per SOP-MT-GV/OPTIMIZATION (Section 5.2) prior to the RRC flow reduction.

Event 2 (TRG-2) CRD Drive Header Flow Control Valve controller (CRD-FC-600) output fails high while in automatic which causes 603.A7 5-8 (CRD PUMP SUCTION FLTR D HIGH) annunciator to come in caused by abnormally high system flow. Upon finding the CRD-FC-600 controller output failed high, the ATC operator informs the CRS and shifts the controller to manual and restores CRD system parameters to normal. Annunciator will clear once system parameters restored to normal.

Event 3 (TRG-3) The shaft on RHR-SYS-A/LPCS Keep Fill Pump (LPCS-P-2) seizes causing a trip of the pump. The RHR A PUMP DISCH PRESS HIGH/LOW annunciator alarms shortly after LPCS-P-2 trips. The LPCS PUMP DISCH PRESS HIGH/LOW annunciator will alarm ~13 minutes after LPCS-P-2 trips (unless LPCS pump started before then). Based on system status and ARP direction, the CRS will direct the BOP operator to start the Low Pressure Core Spray (LPCS) Pump (and place into Suppression Pool Mixing per SOP-LPCS-SP) to maintain system availability provided the LPCS PUMP DISCH PRESS HIGH/LOW annunciator is not in alarm. To prevent an inadvertent start of Residual Heat Removal (RHR) Pump 2A (RHR-P-2A) and therefore a potential for water hammer, the CRS will direct control power fuses removed (TRG-7) from the RHR-P-2A starting circuit. If LPCS pump is not started and the LPCS PUMP DISCH PRESS HIGH/LOW annunciator is received, LPCS Pump control power fuses will also be removed (TRG-8). The CRS will refer to ABN-RHR-DEPRESS as time permits to determine system recovery actions.

With RHR-P-2A and LPCS inoperable, the CRS refers to Technical Specifications and Licensee Controlled Specifications and determines the following actions are applicable:

LCO 3.5.1 A.1 (RHR-SYS-A & LPCS are both tracked as inoperable) - Restore respective subsystem to operable status within 7 days

LCO 3.5.1 C.1 - Restore either RHR-SYS-A or LPCS subsystem to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

LCO 3.6.1.5 A.1 - Restore RHR-SYS-A Drywell Spray subsystem to operable status within 7 days

LCO 3.6.2.3 A.1 - Restore RHR-SYS-A Suppression Pool Cooling subsystem to operable status within 7 days

LCO 3.3.3.2 A.1 - Restore required Function (9.d - RHR-SYS-A Loop Pump) to operable status within 30 days

RFO 1.6.1.5 A.1 - Restore RHR-SYS-A Suppression Pool Spray subsystem to operable status within 7 days Note that LCOs 3.4.6, 3.4.9, and 3.6.1.3 are considered but not applicable with the plant in Mode 1.

Event 4 (TRG-4) Main Steam pressure switch 23D (MS-PS-23D) fails high causing Reactor Protection System (RPS) relay K5D (RPS-RLY-K5D) to actuate a RPV Pressure High Trip Scram relay (as evidenced by annunciator 603.A8 2-2 (RPV PRESS HIGH TRIP)). This actuation causes a half scram on the RPS B side with all RPS B white RPS scram lights de-energized. The ATC operator will determine that two control rods (38-35 and 18-15) inadvertently scrammed during the half scram and that control rod 18-15 only partially inserted. The CRS enters ABN-ROD, section 4.2, for inadvertently scrammed rods. The ATC operator reduces RRC flow to 74 Mlbm/hr at 5% per minute. Following flow reduction, an attempt is made to fully insert control rod 18-15 using the CONTINUOUS INSERT pushbutton (which will be successful). The crew diagnoses the instrument failure and determines the half scram cannot be reset.

The CRS refers to Technical Specifications and determines that TS 3.3.1.1 (RPS Instrumentation) Action A.1 or A.2 requires affected channel or affected trip system, respectively, to be placed in TRIP within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In addition, control rod 18-15 is considered inoperable for not fully inserting when inadvertently scrammed. LCO NRC Scenario 2 Page 4 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 3.1.3 (Control Rod Operability) Action C.1 requires rod 18-15 to be fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and its associated CRD (HCU) disarmed within four hours.

Event 5 (TRG-5) Bus 81 ground as sensed on MC-8BB which powers Fuel Pool Cooling Pump 1B (FPC-P-1B) causes FPC-P-1B to trip when power fuses blow. With this power loss, the standby Fuel Pooling Cooling pump (FPC-P-1A) will not auto start. ARP 4.627.FPC2.3-1 (CIRCULATION PUMP B DISCHARGE PRESSURE LOW) directs entry into ABN-FPC-LOSS. The BOP operator will manually start FPC-P-1A to re-establish fuel pool cooling.

Resetting the Bus 81 ground annunciator (TRG-9) will be successful, if attempted, since ground cleared upon the FPC-P-1B power fuses blowing. Since the status of the FPC-P-1B thermal overloads are unknown at this point the BOP operator may place the FPC-P-1B control switch in the IR-69 position to allow reset of associated overloads.

Event 6 (TRG-6) Trip of CB-1/7 (4160V feed from SM-1 to SM-7) results in an automatic transfer of Division 1 AC safety bus (SM-7) to the Backup Transformer (TR-B). The transient results in a trip of the LPCS Pump (previously started) and a loss of RPS Motor Generator A power to RPS A. With a RPS B half scram signal already present, a full scram signal now exists. The ATC operator recognizes a scram should have occurred and that an ATWS condition exists. The ATC operator takes scram actions including pressing all Manual scram pushbuttons and initiating ARI logic. Both trains of SLC are started due to reactor power being > 5%.

The CRS enters PPM 5.1.1 (RPV Control) and transitions into PPM 5.1.2 (RPV Control - ATWS) and directs the BOP operator to inhibit ADS and to take manual control of HPCS. The CRS addresses the level leg first and directs the BOP operator to perform PPM 5.5.6 (Bypassing MSIV Low RPV Level and High Steam Tunnel Temperature interlocks) to allow MSIVs to stay open on subsequent RPV level reduction. PPM 5.5.1 (Overriding ECCS Valve Logic To Allow Throttling ECCS Injection) is also performed. The CRS then directs stopping and preventing all injection into the RPV except for SLC, CRD and RCIC. When level reaches -65 inches, the ATC operator will restart injection into the RPV through the RFW Startup flow control valve to maintain a RPV Level band of -80 to -140 inches. (CT #1) (CT #2) The CRS directs an RPV pressure band of 800 to 1050 psig with the Digital Electro-Hydraulic (DEH) system in automatic. If reactor power is above 25%, the capacity of the RFW Start-up flow line will be exceeded and the ATC operator will have to augment flow by opening RFW-V-109 (Bypass valve for Feedwater Heaters 6A and 6B). The BOP operator performs PPM 5.5.11 (Alternate Control Rod Insertions) in an attempt to insert control rods.

This event forms a portion of significant CGS PSA Accident Sequence (TTC044) (Ref: PSA-1-SM-0001 (Rev 7))

Event 7 Standby Liquid Control (SLC) Pump 1A fails due to a sheared shaft and SLC Pump 1B discharge is partially blocked resulting in a reduced SLC injection flow in the RPV at approximately 24 gpm. This injection rate will cause reactor power to drop slowly but not prior to the crew lowering RPV level to -80 to -140 inches. Reactor Water Cleanup Valve 4 (RWCU-V-4) does not auto close on the SLC initiation but will be closed manually.

Event 8 Reactor Water Cleanup Valve 4 (RWCU-V-4) does not auto close on the SLC initiation but will be closed manually.

Event 9 Control rods insertion will be attempted per PPM 5.5.11 (Alternate Control Rod Insertions). Since hydraulic ATWS occurred (no white RPS scram lights lit), the BOP operator will remove two (2) ARI fuses and bypass (via switch) the Scram Discharge Volume (SDV) High Level trip. CRD-P-1A will be found tripped and will have to be restarted before a re-scram is attempted. The Scram - Reset - Scram method of control rod insertion is not effective requiring the BOP operator to bypass the Rod Worth Minimizer (RWM) and manually insert control rods individually using CRD drive pressure.

NRC Scenario 2 Page 5 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 TERMINATION CRITERIA: The scenario will be terminated when RPV level is being maintained between -

80 inches to -140 inches, one attempt at scram-reset-scram has been completed, and manual insertion of control rods has commenced OR as directed by the Examination Team.

NRC Scenario 2 Page 6 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Critical Task Determination Table Measurable Performance Critical Task Safety Significance Cueing Performance Feedback Indicators CT #1 - During This is a procedural Procedural direction Crew stops and RPV level and ATWS with power requirement of PPM by PPM 5.1.2 Step L- prevents injection reactor power start

> 5%, terminate and 5.1.2 (RPV Control - 6 directs lowering with the exception lowering.

prevent injection with ATWS). Allowing RPV level to < -65 of SLC, RCIC, and exception of SLC, SLC, RCIC and CRD inches by stopping CRD.

RCIC, and CRD, into injection avoids and preventing all the RPV until RPV conflicts with other injection into RPV level is -65 inches to instructions in the except from boron establish a Lowered EOPs such as injections systems, Level (LL). injecting SLC and RCIC and CRD, inserting control rods. defeating interlocks if Stopping other necessary.

injection sources prevents potential fuel damage due to cold water injection.

(Ref: PPM 5.0.10 Rev 21, section 8.3.4.)

CT #2 - Maintain Prevent unnecessary Procedural direction Crew uses Reactor RPV level RPV level above - significant challenge to by PPM 5.1.2 Step L- Feedwater system indication.

186 inches. Short containment or the 12 directs to maintain RPV excursions below - RPV. maintaining RPV level above -186 186 inches does not level from -140 inches.

constitute failure of inches to -80 inches CT provided level (best practice band) (ED required if restored and with outside shroud level cannot be maintained above - injection systems restored and 186 inches within 10 (Table 5). maintained above minutes of going -186 inches) below -186 inches. OI-15 (EOP and EAL Clarifications),

Section 4.3.2.b.)

NRC Scenario 2 Page 7 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 SIMULATOR SETUP Unload simulator (between each scenario)

Verify in ILC load Reload simulator Reset to IC-201 (reset, go to Run, reset again)

Load Scenario 2 Schedule file Load Scenario 2 Event file (if not loaded automatically)

Validate that there are no unexpected annunciators or parameters out of band Place tagout on CRD-P-1B Protect CRD-P-1A Verify pump running magnets Have marked up copy of PPM 3.2.6 marked up through steps 5.1.6 (ensure N/Ad steps are initialed) for the crew at their pre-brief location (outside the simulator)

NRC Scenario 2 Page 8 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 SCHEDULE FILE

1

<ACTION>schedule Schedule\local.sch</ACTION>

</ITEM>

1

<ACTION>Event D:\NRC Scenario Support Files\2017 NRC SC-2.evt</ACTION>

</ITEM>

1

<ACTION>Insert malfunction BST-FPC020F to FAIL_TO_TRIP</ACTION>

</ITEM>

1

<ACTION>Insert malfunction PMP-SLC001B</ACTION>

<DESCRIPTION>SLC-P-1A SLC PUMP 1A SHAFT BREAK</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction PMP-SLC002F to 40.00000</ACTION>

<DESCRIPTION>SLC-P-1B SLC PUMP 1B REDUCED FLOW</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-CRD002 to FA_CTRL_FUS</ACTION>

<DESCRIPTION>CB-CRD-P-1B CRD-P-1B MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOV-RWU010F to FAIL_AUTO_CLOSE</ACTION>

<DESCRIPTION>RWCU-V-4 RWCU SUCTION OUTBOARD ISO</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction CNH-CRD001E to 100.00000 on event 2</ACTION>

<DESCRIPTION>CRD-FC-600 FLOW CONTROL (M/A STATION) AUTO OUTPUT</DESCRIPTION>

</ITEM>

2

<ACTION>Insert malfunction BST-CRD001F to SPURIOUS_TRIP on event 2</ACTION>

<DESCRIPTION>CRD-DPS-15 CRD-P-1A&amp1B SUCTION FILTER DP</DESCRIPTION>

</ITEM>

NRC Scenario 2 Page 9 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017

1

<ACTION>Insert malfunction PMP-CSS004S on event 3</ACTION>

<DESCRIPTION>LPCS-P-2 LPCS/RHR A WATER LEG PUMP SHAFT SEIZURE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BST-RRS067F to SPURIOUS_TRIP on event 4</ACTION>

<DESCRIPTION>MS-PS-23D RX PRESS TRIP LOGIC B2 SCRAM</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC007-3835 on event 4</ACTION>

<DESCRIPTION>ROD 3835 SINGLE ROD SCRAM</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC007-1815 on event 4</ACTION>

<DESCRIPTION>ROD 1815 SINGLE ROD SCRAM</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC005-1815 after 2 on event 4</ACTION>

<DESCRIPTION>ROD 1815 STUCK</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOT-FPC002G to 100.00000 on event 5</ACTION>

<DESCRIPTION>FPC-P-1B FUEL POOL CIRCULATION PUMP B WINDING OVERCUR</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-EPS003 after 10 to TRIP on event 6</ACTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-CRD007A1 to 100.00000 on event 6</ACTION>

<DESCRIPTION>HYDRAULIC ATWS EAST SDV BLOCKAGE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-CRD007A2 to 100 on event 6</ACTION>

<DESCRIPTION>HYDRAULIC ATWS EAST SDV</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-CRD007B1 to 100.00000 on event 6</ACTION>

<DESCRIPTION>HYDRAULIC ATWS WEST SDV BLOCKAGE</DESCRIPTION>

NRC Scenario 2 Page 10 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017

</ITEM>

1

<ACTION>Insert malfunction MAL-CRD007B2 to 100 on event 6</ACTION>

<DESCRIPTION>HYDRAULIC ATWS WEST SDV</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-RHR001 to FA_CTRL_FUS on event 7</ACTION>

<DESCRIPTION>CB-RHR-P-2A RHR-P-2A MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-CSS002 to FA_CTRL_FUS on event 8</ACTION>

<DESCRIPTION>CB-LPCS-P-1 LPCS-P-1 MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BST-CRD001F to FAIL_TO_TRIP on event 9 delete in 5</ACTION>

<DESCRIPTION>CRD-DPS-15 CRD-P-1A&amp1B SUCTION FILTER DP</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC007-1815 on event 10 delete in 3</ACTION>

<DESCRIPTION>ROD 1815 SINGLE ROD SCRAM</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC005-1815 on event 10 delete in 3</ACTION>

<DESCRIPTION>ROD 1815 STUCK</DESCRIPTION>

</ITEM>

<ACTION>Insert override IND-RMC001AXV to ON</ACTION>

<DESCRIPTION>FULL CORE ROD/DETEC COORD &amp CHANNEL DISP SUB PANEL STATUS LAMP (18,8)</DESCRIPTION>

</ITEM>

<ACTION>Insert override IND-RMC001AWU to ON</ACTION>

<DESCRIPTION>FULL CORE ROD/DETEC COORD &amp CHANNEL DISP SUB PANEL STATUS LAMP (18,8)</DESCRIPTION>

</ITEM>

</SCHEDULE>

NRC Scenario 2 Page 11 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT FILE

<EVENT>

<TRIGGER id="10" description="Rod 18-15 unstuck">XSRLA04L &gt 0</TRIGGER>

</EVENT>

NRC Scenario 2 Page 12 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1

==

Description:==

Lower reactor power with Reactor Recirculation (RRC) flow to 90% for load following per PPM 3.2.6 (which includes placing Main Turbine into Governor Valve Sequential Valve Mode).

Event is initiated by the turnover and starts with PPM 3.2.6 step 5.1.7.

Time Position Applicants Actions or Behavior Examiner Note: Following steps are from PPM 3.2.6 (Power Maneuvering) which was previously completed (marked up) through step 5.1.6.

CRS 5.1.7 Records date and time downpower initiated.

Comment:

5.1.8 Directs BOP to enter Sequential Valve Operation per SOP-MT-GV/OPTIMIZATION (Section 5.2).

Comment:

Examiner Note: Following steps are from SOP-MT-GV/OPTIMIZATION (Section 5.2)

BOP Performs the following to enter Sequential Valve Operation:

5.2.1 If VPL DEMAND is not at 100%, then set VPL DEMAND to 100% as follows (Menu, Main Display):

SELECT VPL TARGET.

ENTER 100%.

SELECT OK.

SELECT GO.

SELECT YES.

VERIFY GO illuminated.

VERIFY VPL DEMAND ramps to VPL TARGET value.

Comment:

NRC Scenario 2 Page 13 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1 (CONTINUED)

BOP 5.2.2 Completes entry into Sequential Valve Mode as follows:

SELECT SEQUENTIAL VALVE MODE.

SELECT YES.

VERIFY GV-1 and GV-4 move to their pre-optimization positions (approximately equal).

VERIFY SEQUENTIAL VALVE MODE is illuminated.

Comment:

Examiner Note: Following steps are a continuation of PPM 3.2.6 (Power Maneuvering).

CRS 5.1.10 Assigns an individual to track thermal power changes.

Comment:

Examiner Note: Crew will track change in power as scenario progresses.

CRS 5.1.11 & 5.1.12 If thermal power changes GT 15% in one hour, then notify Chemistry to evaluate the Offgas release rate.

Comment:

Examiner Note: Main Generator output will not be reduced to 1000 MWe as specified in step 5.1.15 since reactor power reduction is only to 90%.

CRS 5.1.15 Directs ATC to lower power with flow to achieve 90% reactor power at a rate not to exceed 1% per minute.

Comment:

NRC Scenario 2 Page 14 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1 (CONTINUED)

Examiner Note: Following steps are from Quick Card SOP-RRC-FLOW-QC.

Examiner Note: The BOP is expected to act as peer checker for this evolution.

ATC Lowers reactor power using RRC Flow per SOP-RRC-FLOW-QC (Section 2.1):

Examiner Note: Sufficient margin to fuel-preconditioning limits exist as specified in turnover.

ATC 2.1.1 Informs CRS to monitor fuel-preconditioning limits (per 9.3.18) while changing reactor power.

Comment:

2.1.2 Verifies both RRC individual flow controllers are in Auto and then lowers RRC flow using RRC-M/A-R675 (Master Control) Lower pushbutton, as necessary, to achieve a 1% per minute power change until 90% power is achieved.

Comment:

2.1.3 Verifies total core flow is LT 105%.

Comment:

2.1.4 Verifies RRC loop A and B is LT 57.5 Mlb/hr.

Comment:

2.1.5 Notifies the CRS when the change in Reactor power is complete.

Comment:

NRC Scenario 2 Page 15 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 2

==

Description:==

CRD Drive Header Valve controller (CRD-FC-600) output fails high while in automatic.

Event is initiated after CRS gets the report that the power reduction is complete (or as directed by the Exam team) and is activated using TRIGGER 2.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 2 Time Position Applicants Actions or Behavior ATC Responds to CRD PUMP SUCTION FLTR P HIGH alarm (P603.A7 5-8).

Comment:

Observes CRD Cooling Header flow at ~70 gpm and Drive Header/Reactor D/P at ~350 psid and informs the CRS before referring to ARP.

Comment:

Examiner Note: Steps below to take manual control are authorized per PPM 1.3.1 (Operating Policies, Programs and Practices) step 4.6.4.

Examiner Note: Steps below may be performed without a procedure as permitted by OI-9 (Operations Standards and Expectations) section 16.3.1.

ATC Observes Flow Control Valve (CRD-V-2B) full open.

Comment:

Observes CRD Flow Controller (CRD-FC-600) red arrow upscale and the signal is near 100% and informs the CRS.

Comment:

Places CRD-FC-600 controller in manual.

Comment:

Depresses the close pushbutton to restore CRD Cooling Header flow to

~62 GPM and Drive Header D/P to ~265 psid.

Comment:

Observes CRD-V-2B dual indication and the red arrow on CRD-FC-600 returning to the green band.

Comment:

NRC Scenario 2 Page 16 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 2 (CONTINUED)

Examiner Note: Following step is from ARP P603.A7 5-8 (CRD Pump Suction Filter D/P HIGH).

Only step 1 applies since controller failure is causing the alarm due to excessive flow.

Examiner Note: BOP may perform below step while ATC performs manipulations.

BOP 1: Checks CRD-dPIS-15 (CRD Pump Suction Filter Differential Pressure)

(CRD-IR-1A).

Comment:

BOOTH ROLEPLAY - If sent to check suction filter D/P, wait 1 minute then report D/P at 9 psid (if suction filter annunciator locked in) or 5 psid (if suction filter annunciator cleared).

BOOTH ROLEPLAY - If sent to investigate, wait 5 minutes then report Nothing abnormal found with CRD system.

CRS Contacts Work Control for assistance in troubleshooting controller failure.

Comment:

NRC Scenario 2 Page 17 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 3

==

Description:==

RHR-SYS-A/LPCS Keep Fill Pump (LPCS-P-2) trip (Tech Spec)

Event is initiated after CRS gets the report that CRD parameters have been returned to normal (or as directed by the Exam team) and is activated using TRIGGER 3.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 3 Time Position Applicants Actions or Behavior Examiner Note: RHR A discharge low pressure alarm comes in shortly after LPCS-P-2 (Keep Fill) pump shaft seizes. Pump continues to run for several seconds before tripping on over current.

BOP Responds to RHR A PUMP DISCH PRESS HIGH/LOW alarm (P601.A4 3-1).

Comment:

Observes RHR Loop A discharge pressure at ~20 psig (Low) and informs the CRS before referring to ARP.

Comment:

Examiner Note: Below alarms/indications come in when the keep fill pump breaker trips open.

Examiner Note: Below Out Of Service alarms along with the illuminated BYPASS AND INOPERABLE STATUS PANEL (BISI) for LPCS-P-2 Power Loss/OL is used to determine the required ARP actions. RHR A and LPCS BISIs both light (and require the same ARP actions) since they have the Keep Fill pump in common. Either ARP may be used.

BOP Several seconds later responds to the RHR A OUT OF SERVICE and LPCS OUT OF SERVICE alarms (P601.A4 6-1 & P601.A3 6-3, respectively) and associated BISIs caused by LPCS-P-2 (keep fill pump) power loss/overload.

Comment:

Observes panel indication lost for LPCS-P-2 (power loss due to breaker trip) and informs CRS before referring to ARP.

Comment:

Examiner Note: CRS may give priority to starting LPCS pump to maintain its availability before other ARP actions are performed. RHR A pump should not be started.

Examiner Note: LPCS low pressure alarm comes in ~13 min after keep fill pump shaft seizes.

CRS has sufficient time to direct LPCS pump started before LPCS discharge low pressure alarm is received. If crew does not start the LPCS Pump then its control power fuses should be removed following receipt of the LPCS discharge low pressure alarm (P601.A3 5-3).

BOOTH OPERATOR - If directed to remove control power fuses for LPCS Pump, wait 3 minutes then ACTIVATE TRIGGER 7. Report LPCS Pump control power fuses have been removed.

BOOTH OPERATOR - If asked, pre-start checks for LPCS and RHR Pump A are complete.

NRC Scenario 2 Page 18 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 3 (CONTINUED)

Examiner Note: Following step is from ARP 4.601.A4 6-1 for RHR A Out of Service (or ARP 4.601.A3 6-3 for LPCS Out of Service). Either will direct actions for LPCS-P-2 PWR LOSS/OL.

BOP 1. Requests permission from CRS to start LPCS-P-1 per SOP-LPCS-SP (LPCS Suppression Pool Mixing) to maintain operability.

Comment:

Examiner Note: Following steps (to start LPCS-P-1) are from SOP-LPCS-SP (LPCS Suppression Pool Mixing) section 5.1.

Examiner Note: It is expected the CRS will allow an auto start of Service Water Pump A.

BOP 5.1.2 Informs CRS to ENTER LPCS-SYS-1 as inoperable, but available, in the Plant Logging System.

Comment:

5.1.3 Starts LPCS-P-1 (should make plant announcement before starting).

Comment:

5.1.4 Verifies LPCS-FCV-11 opens during low flow conditions (approximately 800 gpm) (Minimum Flow Bypass).

Comment:

5.1.5 Throttles open LPCS-V-12 for approximately 6400 gpm (Test Bypass to Suppression Pool).

Comment:

NRC Scenario 2 Page 19 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 3 (CONTINUED)

BOP 5.1.6 Verifies LPCS-FCV-11 closes (approximately 800 gpm).

Comment:

5.1.7 Verifies SW-P-1A running.

Comment:

5.1.8 Notifies HP that radiological conditions may have changed.

Comment:

5.1.9 Monitors Suppression Pool temperatures.

Comment:

Examiner Note: Following step is a continuation of ARP for RHR A (or LPCS) Out of Service (loss of Keep Fill pump)

BOP 2. Refers to ARP 4.601.A4 3-1 (RHR A PUMP DISCH PRESS HIGH/LOW).

Comment:

Examiner Note: Following steps are from ARP 4.601.A4 3-1 (RHR A PUMP DISCH PRESS HIGH/LOW.

BOP 1. Checks RHR Loop A pressure at the following:

RHR-PI-612A (H13-P601)

RHR-PIS-22A (H22-P018, RB 501)

TDAS pt. X155 Comment:

4.a IF not operating RHR per the EOPs, then inhibits RHR-P-2A start by pulling its control power fuses.

Comment:

BOOTH OPERATOR - If directed to report RHR Loop A discharge pressure on instrument rack H22-P018 in RB 501, wait 2 minutes then report Instrument rack H22-P018 pressure indicates ________ psig (refer to soft panel and report to nearest 5 psig increment).

BOOTH OPERATOR - If directed to remove control power fuses for RHR Pump A, wait 3 minutes then ACTIVATE TRIGGER 8. Report RHR Pump 2A control power fuses have been removed.

NRC Scenario 2 Page 20 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 3 (CONTINUED)

BOP 4.b Checks operation of LPCS-P-2 (Water Leg Pump).

Comment:

BOOTH ROLEPLAY - If sent to investigate status of LPCS-P-2 locally at the pump, wait 2 minutes then report LPCS-P-2 is warm to the touch and not running.

4.c Verifies the following valves are closed:

RHR-V-16A (Upper Drywell Spray)

RHR-V-17A (Upper Drywell Spray)

RHR-V-24A (Test Line Isolation)

RHR-V-27A (Suppression Pool Spray)

RHR-V-42A (LPCI Isolation)

Comment:

Examiner Note: CRS may refer to below procedure but no verifiable actions will be performed.

CRS 4.d Refers to ABN-RHR-DEPRESS (Starting RHR Loop A Following Depressurization) due to loss of the Keep Fill system.

Comment:

Examiner Note: Of the Technical Specification referenced below in the ARP, only LCO 3.5.1, 3.6.1.5 and 3.6.2.3 apply. Other LCOs apply which are not listed in the ARP. See end of this event for all applicable CRS Technical Specification actions.

CRS 5. Refers CRS to Technical Specifications 3.4.6, 3.4.9, 3.5.1, 3.6.1.5, 3.6.2.3, and 3.6.1.3 and Licensee Controlled Specifications 1.3.4.6.

Examiner Note: Following step is a continuation of ARP for RHR A (or LPCS) Out of Service (loss of Keep Fill pump)

BOP 3. Checks the status of the breaker, control power fuses, or thermal overloads for LPCS-P-2 at LPCS-42-7B6B.

Comment:

BOOTH ROLEPLAY - If sent to investigate status of LPCS-P-2 at the breaker, wait 2 minutes then report The breaker at LPCS-42-7B6B was found tripped. There is a mild acrid odor near the breaker. (NO FIRE)

BOP 4. Refers CRS to Technical Specifications 3.5.1, 3.4.9, 3.6.1.5 and 3.6.2.3.

Comment:

NRC Scenario 2 Page 21 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 3 (CONTINUED)

CRS Evaluates Technical Specifications and determines the following Required Actions apply:

LCO 3.5.1 A.1 (RHR-SYS-A & LPCS are both tracked as inoperable) -

Restore respective subsystem to operable status within 7 days Comment:

LCO 3.5.1 C.1 - Restore either RHR-SYS-A or LPCS subsystem to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Comment:

LCO 3.6.1.5 A.1 - Restore RHR-SYS-A Drywell Spray subsystem to operable status within 7 days Comment:

LCO 3.6.2.3 A.1 - Restore RHR-SYS-A Suppression Pool Cooling subsystem to operable status within 7 days Comment:

LCO 3.3.3.2 A.1 - Restore required Function (9.d - RHR-SYS-A Loop Pump) to operable status within 30 days Comment:

Evaluates Licensee Controlled Specifications (LCS) and determines the following Required Action applies:

RFO 1.6.1.5 A.1 - Restore RHR-SYS-A Suppression Pool Spray subsystem to operable status within 7 days Comment:

Examiner Note: LCOs 3.4.6, 3.4.9, and 3.6.1.3 are considered but not applicable.

CRS Direct postings for Protected Equipment to include RHR B, RHR C, HPCS, HPCS SW, DG2, DG3, and SW-B Comment:

NRC Scenario 2 Page 22 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 4

==

Description:==

Failure of MS-PS-23D which causes a half scram on RPS B side. Two control rods scram but one does not go full in (must be manually inserted) (Tech Spec)

Event is initiated after LPCS has been placed into Suppression Pool Mixing and the required Tech Spec Actions entered (or as directed by the Exam team) and is activated using TRIGGER 4.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 4 Time Position Applicants Actions or Behavior Examiner Note: The PRV High Pressure Trip causes the half scram on RPS B.

ATC Responds to RPV PRESS HIGH TRIP (P603.A8 2-2) and 2 SCRAM SYSTEM B (P603.A8 3-4) alarms and informs CRS.

Comment:

Validates that a half scram occurred on RPS B (all white RPS B scram lights de-energized) and informs the CRS.

Comment:

Examiner Note: The Rod Accumulator Trouble results from the two rods which scrammed on the half scram.

ATC Responds to ROD ACCUMULATOR TROUBLE (P603.A7 6-7) alarm.

Comment:

Scans the full core display (or observes RWM screen) for drifting and/or scrammed control rods.

Comment:

Recognizes two control rods have blue SCRAM lights lit and flashing ACCUM lights and informs the CRS.

Comment:

Selects control rod 38-35 and observes it full in.

Comment:

Selects control rod 18-15 and observes it partially inserted.

Comment:

Acknowledges Rod Accumulator Trouble alarm from P603 to allow any subsequent Rod Accumulator Trouble inputs to activate alarm.

Comment:

NRC Scenario 2 Page 23 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 4 (CONTINUED)

CRS Refers to Technical Specification 3.1.5 in response to rod 18-15 which failed to fully insert on scram.

Comment:

Examiner Note: Following steps are from ARP 4.603.A8 3-4 (2 SCRAM SYSTEM B).

ATC 2.a. Checks the Full Core Display for individual control rods that may have scrammed (may have been previously performed).

Comment:

CRS 2.b. Enters ABN-ROD (Control Rod Faults).

Comment:

2.c. Stops all maintenance or surveillance testing that has the potential for generating a trip on the unaffected RPS channel (A).

Comment:

Examiner Note: Failed pressure switch (MS-PS-23D) which failed in the TRIP condition caused the half scram which cannot be bypassed (without maintenance support). The half scram cannot be immediately reset.

CRS 4. Refers to Technical Specification 3.3.1.1 for failed RPS instrument.

Comment:

Examiner Note: Following steps are from ABN-ROD (Control Rod Faults) section 4.2 (note that there are no Immediate Actions that are currently applicable per section 3.2).

ATC 4.2.1 Reduce core flow to 74 Mlbm/hr at 5% per minute (core flow is

> 80 Mlbm/hr (on MS-FR-613 at H13-P603)).

Comment:

CRS 4.2.2 If thermal power changes GT 15% in one hour, then notify Chemistry to evaluate the Offgas release rate.

Comment:

ATC 4.2.4.a. Selects the affected control rod(s) and verifies position (may have been previously performed).

Comment:

NRC Scenario 2 Page 24 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 4 (CONTINUED)

Examiner Note: Control rod 18-15 is the partially inserted rod.

ATC 4.2.4.b.1) Selects control rod 18-15 and depresses the CONTINUOUS INSERT Pushbutton at H13-P603.

Comment:

4.2.4.b.2) Drives control rod 18-15 to its FULL IN position.

Comment:

4.2.4.b.3) Releases the CONTINUOUS INSERT Pushbutton.

Comment:

4.2.4.b.4) Verifies control rod 18-15 remains in the FULL IN position.

Comment:

4.2.4.c. If necessary, reset the rod accumulator trouble annunciator using the accumulator trouble acknowledge pushbutton (H13-P603).

Comment:

4.2.4.d. If necessary, reset the control rod drift annunciator using the rod drift reset pushbutton (H13-P603).

Comment:

CRS 4.2.4.e. Refers to Technical Specifications (Reactivity).

Comment:

4.2.4.f. Initiates (or directs) a MON run to verify acceptable thermal limits and preconditioning.

Comment:

Examiner Note: Following steps are from ARP 4.603.A8 2-2 (RPV PRESS HIGH TRIP). Step 2.c.

cannot be performed (RPS B will not reset).

BOP 2.a. Determines cause for half scram by investigating backpanel area and observing that RPS relay (RPS-RLY-K5D) has dropped out.

Comment:

BOOTH ROLEPLAY - If sent to investigate MS-PS-23D and/or B, wait 2 minutes then report Nothing appears abnormal with MS-PS-23D(B).

NRC Scenario 2 Page 25 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 4 (CONTINUED)

Examiner Note: CRS may also declare rod Control rod 38-35 inoperable based on not knowing reason for rod scram.

CRS Evaluates Technical Specifications and determines the following Required Actions apply:

LCO 3.1.3 C.1 - Fully insert control rod 18-15 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Comment:

LCO 3.5.1 C.2 - Disarm CRD for control rod 18-15 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Comment:

LCO 3.3.1.1 A.1 - Place High Pressure trip channel in TRIP -OR-LCO 3.3.1.1 A.2 - Place RPS B trip system in TRIP Comment:

NRC Scenario 2 Page 26 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 5

==

Description:==

Ground causes FPC-P-1B to spuriously trip (FPC-P-1A fails to auto start).

Event is initiated after control rod 18-15 is fully inserted and required Tech Spec Actions entered (or as directed by the Exam team) and is activated using TRIGGER 5.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 5 Time Position Applicants Actions or Behavior Examiner Note: FPC BOARD FPC-2 TROUBLE is an alarm informing the BOP that there is an alarm on backpanel H13-P627 (Fuel Pool Cooling Div 2 panel).

BOP Responds to BUS 81 GROUND (P800.C5 3-5) and FPC BOARD FPC-2 TROUBLE (P851-S2) alarms and informs CRS.

Comment:

Examiner Note: Below Out Of Service alarm along with the illuminated BYPASS AND INOPERABLE STATUS PANEL (BISI) for Fuel Pool Cooling Pump 1B Loss is used to determine the required ARP actions.

BOP Responds to FPC DIV 2 OUT OF SERVICE (P627.FPC2 4-1) alarm and identifies BISI (FPC-P-1B PWR LOSS) as cause. Informs CRS.

Comment:

Examiner Note: Following steps are from ARP 4.627.FPC2 4-1 (FPC DIV 2 OUT OF SERVICE).

Examiner Note: FPC-P-1B power fuses blew which requires a manual start of FPC-P-1A.

BOP Manually starts FPC-P-2A and inform CRS.

Comment:

CRS Enters ABN-FPC-LOSS on entry condition (unplanned loss of FPC).

Comment:

Examiner Note: Following steps are from ABN-FPC-LOSS section 4.1. CRS may only direct steps 4.1.1 and 4.1.2.a be performed based on ground fault on FPC-P-1B.

BOP 4.1.1 Monitor Spent Fuel Pool level and temperature as directed.

Comment:

4.1.2.a. Place FPC-P-1B control switch in the IR-71(69) position.

Comment:

NRC Scenario 2 Page 27 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 5 (CONTINUED)

Examiner Note: Following steps are from ARP 4.800.C5 3-5 (BUS 81 GROUND).

Examiner Note: Ground will be reported to be on Bus E-MC-8BB.

BOP 1. Directs area operator to investigate ground location on the SL-81 Ground Fault Indication Panel.

Comment:

Examiner Note: Alarm in MCR will clear (once locally reset) since power fuses (upon blowing) removed the ground for FPC-P-1B which is powered from E-MC-8BB.

Examiner Note: Although step 3 below directs exit of ARP, CRS may still perform steps 7. & 8.

BOP 2. & 3. Directs area operator to attempt to reset ground alarm locally (ground alarm relay resets).

Comment:

CRS 7. Maintains grounded circuit de-energized by not replacing fuses for FPC-P-1B until troubleshooting plan developed.

Comment:

8. Directs Work Request be generated for repair of grounded circuit.

BOOTH ROLEPLAY - If sent to investigate ground location, wait 2 minutes then report SL-81 ground appears to be on MC-8BB.

BOOTH OPERATOR - If directed to attempt to reset ground alarm locally, wait 1 minutes then ACTIVATE TRIGGER 9. Report ground indication on MC-8BB is cleared.

NRC Scenario 2 Page 28 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 6

==

Description:==

Trip of E-CB-1/7 with transfer of SM-7 to Backup Transformer results in reactor trip signal.

Event is initiated when crew actions for loss of Fuel Pool cooling are complete (or as directed by the Exam team) and is activated using TRIGGER 6.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 6 Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 800.C3 6-1 (BKR 1/7 TRIP) and informs CRS Comment:

Reports that Bus SM-7 momentarily lost power and automatically transferred to the Backup Transformer Comment:

ATC Reports half scram on RPS A with failure to scram (half scram on RPS B already exists)

Comment:

Examiner Note: Following steps are Immediate Actions from PPM 3.3.1 (Reactor Scram)

ATC 6.1.1: Places Reactor Mode Switch to Shutdown Comment:

6.1.2: Monitors reactor power, pressure and level Comment:

6.1.3: (2 handed operation) Since APRMs are not downscale the following is performed:

6.1.3.a: Depresses the Manual Scram pushbuttons

6.1.3.b: Initiates ARI Comment:

6.1.4: Recognizes that reactor power is > 5% and informs the CRS (See Event 7 for SLC actions)

Comment:

CRS Updates crew on EOP entry into PPM 5.1.1, RPV Control, and directs/verifies that the Mode Switch has been placed in SHUTDOWN Comment:

NRC Scenario 2 Page 29 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 6 (CONTINUED)

CRS Updates crew and exits PPM 5.1.1 (RPV Control) and transitions to PPM 5.1.2 (RPV Control - ATWS)

Comment:

Directs BOP to:

Inhibit ADS and take manual control of HPCS

Verify actuations for +13 and -50 as they occur

Directs pressure control with bypass valves in Auto Comment:

BOP Takes both ADS control switches to the INHIBIT position and acknowledges associated alarms (P601.A3 6-1 ADS DIV 1 OUT OF SERVICE and P601.A2 6-8 ADS DIV 2 OUT OF SERVICE)

Comment:

Arms and Depresses the HPCS system initiation pushbutton while holding the control switch for HPCS-P-1 to STOP Comment:

Takes the control switch for HPCS-V-4 to close when it gets fully opened Comment:

Reports to CRS that ADS is inhibited and manual control of HPCS has been established Comment:

CRS Directs PPM 5.5.6 be performed (Bypassing the MSIV Isolation Interlocks on High Tunnel Temperature and low RPV level)

Comment:

BOP Goes to EOP drawer and gets PPM 5.5.6 procedure and equipment bag containing two keys Comment:

Performs PPM 5.5.6:

At H13-P609 places MS-RMS-S84 to BYPASS

At H13-P611 places MS-RMS-S85 to BYPASS Updates crew upon completion Comment:

NRC Scenario 2 Page 30 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 6 (CONTINUED)

BOP Recognizes and reports EOP entry conditions due to Drywell pressure, Drywell temperature and Wetwell level (as they occur)

Comment:

CRS Updates crew and enters PPM 5.2.1 (Secondary Containment Control)

Establishes a key parameter: Wetwell pressure of 2 psig May establish a key parameter of Drywell temperature at 285°F Comment:

BOP Reports when Wetwell pressure reaches 2 psig Comment:

CRS Directs RCIC-V-1 closed (if Main Turbine online)

Comment:

BOP If directed, closes RCIC-V-1 Comment:

CRS Directs performance of PPM 5.5.1 (Overriding ECCS Valve Logic to Allow Throttling RPV Injection)

Comment:

BOP Goes to EOP drawer and pulls PPM 5.5.1 procedure and equipment bag containing 5 keys and performs PPM 5.5.1:

HPCS - Override HPCS-V-4 (HPCS RPV injection valve) automatic logic by placing HPCS-RMS-S25 in the OVERRIDE position (H13-P625)

LPCS - Override LPCS-V-5 (LPCS RPV injection valve) automatic logic by placing LPCS-RMS-S21 in the OVERRIDE position (H13-P629)

RHR Loop A - Override RHR-V-42A (RHR RPV injection valve) automatic logic by placing RHR-RMS-S105 in the OVERRIDE position (H13-P629)

RHR Loop B - Override RHR-V-42B (RHR RPV injection valve) automatic logic by placing RHR-RMS-S106 in the OVERRIDE position (H13-P618)

RHR Loop C - Override RHR-V-42C (RHR RPV injection valve) automatic logic by placing RHR-RMS-S107 in the OVERRIDE position (H13-P618)

Updates crew to completion of PPM 5.5.1, and that the ECCS injection valves are closed and throttleable Comment:

NRC Scenario 2 Page 31 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 6 (CONTINUED)

CT #1 - During ATWS with power > 5%, terminate and prevent injection with exception of SLC, RCIC, and CRD, into the RPV until RPV level is -65 inches to establish a Lowered Level (LL).

CRS Directs the ATC to:

Stop and prevent condensate and feedwater

Lower level to a band less than -65 inches but greater than -186 inches (preferred band is -80 inches to -140 inches)

Commence RPV injection at -65 inches Comment:

ATC Aligns the Feed and Condensate system per SOP-RFW-FCV-QC quick card as follows:

2.1.1: Starts closing RFW-V-112A and RFW-V-112B

2.1.2: Starts opening RFW-V-118

2.1.3: Verifies RFW-V-109 is closed

2.1.4: Verifies RFW-V-117A and RFW-V-117B open

2.1.5: Verifies RFW-LIC-620 is in manual (V selected for Valve position demand) with 0 output 2.1.6: If Reactor Feed Pumps are operating then perform the following:

b. Places RFW-P-1B in MDEM mode

c. Places RFW-P-1A in MDEM mode

d. Controls Turbine speed as required

e. If desired, then places RFW-FCV-2A (B) in manual and slowly open to approximately 80%

2.1.7: Verifies RFW-V-112A and RFW-V-112B are fully closed 2.1.8: Verifies RFW-V-118 is fully open 2.1.9: If Reactor Feed Pumps are operating, then adjusts the running RFP speed to establish ~ 200 psid across RFW-FCV-10A & 10B using either Feedwater touch screen (H13-P840) 2.1.10: Adjusts RFW-LIC-620 manual output to control RPV level Comment:

NRC Scenario 2 Page 32 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 6 (CONTINUED)

CT #2 - Maintain RPV level above -186 inches. Short excursions below -186 inches does not constitute failure of CT provided level restored and maintained above -186 inches within 10 minutes of going below -186 inches.

ATC Reports EOP entry on low RPV water level at +13 Reports Reactor Power as it drops due to lowering level Maintains RPV level between -65 inches and -186 inches as directed

(-80 inches to -140 inches is the preferred band)

Does not commence feeding until RPV level drops below -65 inches Comment:

CRS Directs PPM 5.5.11, ALTERNATE Control Rod Insertions, be performed to insert control rods (see Event 9)

Comment:

BOP Reports trip of LPCS pump (started during Event 3)

Comment:

NRC Scenario 2 Page 33 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 7

==

Description:==

SLC-P-1A shaft shears when pump starts and SLC-P-1B develops a discharge flow blockage which limits SLC injection flow Event is activated at the beginning of the scenario and is realized when SLC system is started.

Time Position Applicants Actions or Behavior ATC When it is recognized that depressing the manual scram pushbuttons and initiating ARI has not inserted the control rods, refers to SOP-SLC-INJECTION-QC quick card and performs the following:

2.1: Removes the SLC keylock switch blanks and insert both keys into the SLC System control switches Comment:

2.2: Initiates SLC injection by performing the following (H13-P603):

Places SLC System A control switch to the OPER position

Places SLC System B control switch to the OPER position Comment:

2.3: Records the following:

SLC Flow rate (~43 gpm for one pump, or 86 gpm for both pumps)

Will record reduced flowrate of ~24 gpm

Initial SLC tank level

Circle RWCU-V-4 status (should be closed but is open)

Comment:

Reports to CRS that SLC is injecting at a reduced flowrate Comment:

Directs field operator to investigate problems with SLC Comment:

BOOTH ROLEPLAY - If directed to investigate SLC, wait 3 minutes and report It appears that SLC Pump A has a broken shaft and that there is a flow restriction with SLC train B.

NRC Scenario 2 Page 34 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 8

==

Description:==

RWCU-V-4 does not auto close on SLC initiation but can be closed manually Event is activated at the beginning of the scenario and is realized when SLC system is started and RWCU-V-4 does not automatically close.

Time Position Applicants Actions or Behavior ATC After starting both SLC pumps, recognizes that RWCU-V-4 did not automatically close Takes manual action to close RWCU-V-4 (Successful)

Reports issue with RWCU-V-4 to CRS with action taken Comment:

NRC Scenario 2 Page 35 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 9

==

Description:==

Scram/Reset/Scram not effective in inserting control rods - Control rods can be manually driven in Event is activated at the beginning of the scenario and is realized when Scram/Reset/Scram proves inneffective Time Position Applicants Actions or Behavior Examiner Note: Refer to Simulator Guide Attachment 1 in reference to PPM 5.5.11.

BOP Goes to EOP drawer and pulls procedure for PPM 5.5.11 and equipment bag to perform PPM 5.5.11:

Performs PPM 5.5.11:

Determines that no RPS scram lights are lit and:

Removes one TB1 ARI fuse (P650 F01, F02, F03 or F04)

Removes one TB2 ARI fuse (P650 F01, F02, F03 or F04)

Observes that some or all blue scram valve lights are lit and determines Tab B should be performed:

Places the SDV HIGH LEVEL TRIP control switch to BYPASS

Ensures both CRD pumps are running - may direct ABN-CRD MAXFLOW be performed

Determines the scram cannot be reset

Overrides RPS trip signals per Attachment 6.1:

At H13-P611 - Installs a jumper between RPS-RLY-K9B terminal stud 2 and RPS-RLY-K12F terminal stud 4 At H13-P611 - Installs a jumper between RPS-RLY-K9D terminal stud 2 and RPS-RLY-K12H terminal stud 4 At H13-P609 - Installs a jumper between RPS-RLY-K9A terminal stud 2 and RPS-RLY-K12E terminal stud 4 At H13-P609 - Installs a jumper between RPS-RLY-K9C terminal stud 2 and RPS-RLY-K12G terminal stud 4 Comment:

BOOTH ROLEPLAY - If directed to perform ABN-CRD-MAXFLOW, wait 2 minutes and activate Trigger 26. Report completion when valves are fully opened.

NRC Scenario 2 Page 36 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event No. 9 (CONTINUED)

BOP Continues with Tab B operator actions:

Resets the scram by depressing reset pushbuttons

Determines that CRD drive header pressure can be established

Places the RWM bypass switch to bypass on H13-P603

Manually starts to drive control rods by starting at 10-43 and inserting every other rod in every other row

Informs CRS having success driving control rods in

When the Scram Discharge Volume has been drained for more than 2 minutes initiates a manual scram by depressing the four red manual scram pushbuttons.

If all rods did not insert, continues scram/reset/scram per Tab B and raises SDV drain time by 2 minutes

Determines no control rod motion do to Scram/Reset/Scram and requests drain time extension Comment:

TERMINATION CRITERIA: The scenario will be terminated when RPV level is being maintained between -80 inches to -140 inches, one attempt at scram-reset-scram has been completed, and manual insertion of control rods has commenced OR as directed by the Examination Team.

NRC Scenario 2 Page 37 of 41

PPM 5.5.11 ALTERNATE ROD INSERTION (Attachment 6.1)

Any control rod not fully inserted The following may be entered at any time based on fault, with CRS permission, C - Scram Individual Control Rods D - Vent Scram Air Header E - Vent CR Over Piston Volumes Any white NO RPS SCRAM light lit YES REMOVE one TB1 ARI fuse:

REMOVE applicable divisional RPS fuses, based on lit white RPS scram lights, as necessary to insert a full scram.: PANEL FUSE EPN Trip System/ P650 F01 PANEL FUSE (contractor) Group P650 F02 P650 F03 P609 LL-F13 C72A-F18A A1 P650 F04 P609 LL-F14 C72A-F18E A2 P609 BB-F12 C72A-F18C A3 P609 BB-F13 C72A-F18G A4 P611 LL-F13 C72A-F18B B1 P611 MM-F21 C72A-F18F B2 REMOVE one TB2 ARI fuse:

P611 BB-F12 C72A-F18D B3 P611 AA-F20 C72A-F18H B4 PANEL FUSE P650 F01 P650 F02 P650 F03 IF SDV Vent and Drain Valves dont close, P650 F04 THEN REMOVE SDV VENT AND DRAIN VALVE fuses, if required:

EPN SDV PILOT PANEL FUSE (contractor) SOLENOIDS P609 MM-F20 C72A-F17A CRD-SPV-9A/182A Some or NO D

P611 MM-F20 C72A-F17B CRD-SPV-9B/182B All Blue SCRAM valve lights lit YES YES All Rods IN INFORM CRS B

NO INSTALL the following fuses removed Perform concurrently above:

RPS SCRAM fuses B C E

SDV VENT AND DRAIN VALVE fuses ATTACHMENT 1 Page 38 of 41

PPM 5.5.11 ALTERNATE ROD INSERTION (Attachment 6.1)

RESET / SCRAM B

RESTORE RPS B per ABN-RPS, if unable continue with driving rods PLACE SDV HIGH LEVEL TRIP control switch to BYPASS ENSURE both CRD pumps are operating Can NO scram be reset OVERRIDE RPS trip signals, YES Attachment 6.1 RESET SCRAM If unable to reset scram, then MANUALLY DRIVE rods.

ADJUST CRD-FC-600 and CRD-V-3 as necessary to establish CRD drive header pressure.

WHEN Can CRD drive SDV drained header NO more than 2 pressure be minutes established CLOSE CRD-V-34 YES INITIATE manual scram PLACE RWM bypass switch to bypass INFORM MANUALLY DRIVE rods as follows:

YES CRS (NOTE: CRS may direct inserting clustered All Rods IN rods (2 or more adjacent rods) first.

1) Starting with 10-43, insert every other rod in every other row, disregard edge rods.

INSTALL the following 2) Starting with 14-47, insert every other rod in NO fuses removed in TAB A: every other row, disregard edge rods.

3) Starting in row 43, insert the remaining rods

TB1 ARI fuses in every row, disregarding edge rods.

TB2 ARI fuses 4) Insert edge rods Further NO attempts desired YES YES All Rods IN RAISE SDV drain time by 2 minutes NO B E ATTACHMENT 1 Page 39 of 41

PPM 5.5.11 ALTERNATE ROD INSERTION (Attachment 6.1)

Overriding RPS Trips

At H13-P611:

INSTALL a jumper between RPS-RLY-K9B, terminal stud 2, and RPS-RLY-K12F, terminal stud 4.

INSTALL a jumper between RPS-RLY-K9D, terminal stud 2, and RPS-RLY-K12H, terminal stud 4.

At H13-P609:

INSTALL a jumper between RPS-RLY-K9A, terminal stud 2, and RPS-RLY-K12E, terminal stud 4.

INSTALL a jumper between RPS-RLY-K9C, terminal stud 2, and RPS-RLY-K12G, terminal stud 4.

RPS - RLY - K9A (B, C, D) 1 5 7 11 HFA Relay (rear view) 3 9 13 RLY - K9 14 4 10 2 6 8 12 RPS - RLY - K12E(F,G,H) 1 5 7 11 "Jumper" 3 9 13 RLY - K12 14 4 10 2 6 8 12 ATTACHMENT 1 Page 40 of 41

Appendix D NRC Scenario No. 2 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Initial Conditions:

Columbia is operating at 100% power

CRD-P-1B is out of service for extended Maintenance

CRD-P-1A is Protected Shift Turnover:

Lower power to 90% using Reactor Recirculation flow per PPM 3.2.6 (Power Maneuvering) after assuming the shift based on BPA Load Following request

Steps 5.1.1 thru 5.1.6 of PPM 3.2.6 are complete

Proper margin to Pre-Conditioned Status (PCS) exists per PPM 9.3.18

The Reactivity brief has been performed ATTACHMENT 1 Page 41 of 41

INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE COLUMBIA GENERATING STATION SIMULATOR EXAMINATION Place RHR-SYS-A in SP Cooling (LPCS/RHR A ADS Permissive fails to annunciate)

(Tech Spec); Rod (26-19) drifts out. Once inserted, control rod to drift out again (Tech LESSON TITLE Spec); SW-P-1A trips which requires RHR-P-2A to be secured; RFP B vibrations rise requiring RRC Flow reduction and manual trip of RFP B; OBE causes steam leak in RCIC Pump Room with Failure of RCIC-V-8 and RCIC-V-63 to fully close; Manual scram inserted; Steam leak develops in the Main Steam Tunnel; MS-V-22A and MS-V-28A through D fail to automatically close (MS-V-28A through D can be closed manually but does not isolate leak); Emergency Depressurization required on two Max Safes LENGTH OF LESSON 1.5 Hours Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code LO001858 Rev. No. 0 JPM PQD Code Rev. No.

Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/22/16 REVISED BY DATE VALIDATED BY DATE TECHNICAL REVIEW DATE INSTRUCTIONAL REVIEW DATE APPROVED DATE NRC Scenario 3 Page 1 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Facility: Columbia Generating Scenario No.: 3 Op Test No.: 1 Station Examiners: Operators:

Columbia is operating at 85% power due to economic dispatch. Safety Relief Valve 2C (MS-RV-2C) is known to be leaking. Suppression Pool high temperature alarms (601.A11.1-3 and 601.A12.1-3)

Initial Conditions:

have just annunciated. Reactor Closed Cooling (RCC) Pump 1B is tagged out for planned maintenance. RCC-P-1A and RCC-P-1C are protected.

After shift turnover place Residual Heat Removal Pump 2A (RHR-P-2A) in Suppression Pool Cooling Turnover: and allow Standby Service Water Pump 1A (SW-P-1A) to auto start. The pre-evolution brief has been completed and operators are stationed near both pumps.

Critical Tasks:

With reactor at power and with primary system discharging into secondary containment, manually scram CT-1 reactor before any area exceeds its maximum safe operating temperature.

With a primary system discharging into secondary containment and area temperature exceeding maximum safe operating level in more than one area, initiate Emergency Depressurization (ED) by opening seven (7)

Safety Relief Valves (ADS preferred) within 10 minutes of second MSOT being exceeded.

CT-2 Note: If the crew properly elects to invoke the EMERG DEPRESS is anticipated override in ppm 5.1.1 (RPV Control) and in doing so, the second maximum safe operating level is not exceeded, this Critical Task is considered to be met.

Event Malf. Event Type* Event Description No.

N (BOP) Place RHR-SYS-A in Suppression Pool Cooling (LPCS/RHR A ADS 1 N/A TS (SRO) Permissive fails to annunciate during pump start) (Tech Spec) **

C (ATC,SRO) Control rod (26-19) drifts out. Once inserted, releasing the continuous 2 TRG-2 insert pushbutton allows the control rod to drift out again, requiring the TS (SRO) control rod to be isolated (Tech Spec)

C (BOP,SRO) Standby Service Water Pump 1A (SW-P-1A) trips which requires 3 TRG-3 Residual Heat Removal Pump 2A (RHR-P-2A) (currently in Suppression TS (SRO) Pool Cooling) to be manually secured (Tech Spec)

Reactor Feed Pump (RFP) B vibrations rise requiring RRC Flow 4 TRG-4 C (ATC,SRO) reduction and manual trip of the B RFP Operating Bases Earthquake causes a steam leak in the RCIC Pump Room with Failure of RCIC-V-8 and RCIC-V-63 to fully close (preventing 5 TRG-5 M (ALL)

RCIC leak isolation). Manual scram inserted before first secondary containment max safe operating temperature is reached (CT #1) 6 N/A M (ALL) Steam leak develops in the Main Steam Tunnel MS-V-22A and MS-V-28A through D fail to automatically close (MS-V-7 N/A C (BOP) 28A through D can be closed manually but does not isolate leak)

Emergency Depressurization (PPM 5.1.3) is performed when two areas 8 N/A ---

exceed their max safe operating temperature (CT #2)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS) Technical Specifications

- Ref: Columbia OE (AR-00049685 - Root Cause Analysis of RHR-PS-19A Isolation Mispositioning Event)

NRC Scenario 3 Page 2 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Target Quantitative Attributes Actual Description Several MSIVs fail to automatically close and one Malfunctions after EOP entry (1-2) 1 cannot be closed ADS Permissive fails on RHR pump A start; Rod 26-Abnormal events (2-4) 3 19 drifts out; RFB B high vibrations RCIC steam leak requiring scram; Main steam line Major transients (1-2) 2 break PPM 5.1.1 (RPV Control); PPM 5.3.1 (Secondary EOPs entered/requiring substantive actions (1-2) 2 Containment Control)

EOP contingencies requiring substantive actions (0-2) 1 PPM 5.1.3 (Emergency RPV Depressurization)

EOP based Critical tasks (2-3) 2 See Critical Task Determination table Trigger Evaluator How Purpose Malfunction Numbers (TRG-x) Directed Triggered TRG-2 YES Manually Event Initiator MAL-RMC004-2619 TRG-3 YES Manually Event Initiator BKR-SSW001 TRG-4 YES Manually Event Initiator ANN-840A1G05; MAL-FPT005B TRG-5 YES Manually Event Initiator MAL-RWB001; MAL-RCI004 TRG-6 Manually Event Initiator MAL-RMC004-2619 TRG-7 Manually Field Action ANN-840A1G05 TRG-8 Automatically Malf Trigger MAL-RRS006A; MAL-RCI004 TRG-9 Automatically Malf Trigger MAL-RRS006A Initial Condition BST-RHR014F Initial Condition AOV-RRS003F Initial Condition MOV-RCI012F Initial Condition MOV-RCI016F Initial Condition RLY-NSF097F Initial Condition BKR-RCC002 NRC Scenario 3 Page 3 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 SCENARIO 3

SUMMARY

Event 1 As part of the turnover, and with annunciators for Suppression Pool high temperature in alarm (601.A11 1-3 and 601.A12 1-3)), the BOP operator will place Residual Heat Removal Loop A (RHR-SYS-A) into Suppression Pool Cooling mode per SOP-RHR-SPC (Suppression Pool Cooling/Spray/Discharge

/Mixing). Standby Service Water Pump (SW-P-1A) will be allowed to auto start as permitted by procedure.

The CRS declares RHR-SYS-A as inoperable but available and refers to Technical Specifications and the Licensee Controlled Specifications and determines that the following applies:

LCO 3.5.1 Action A.1 which requires restoring RHR-SYS-A to operable status within 7 days

LCO 3.6.1.5 Action A.1 which requires restoring RHR-SYS-A drywell spray subsystem to operable status within 7 days

LCO 3.6.2.3 Action A.1 which requires restoring RHR-SYS-A suppression pool cooling subsystem to operable status within 7 days

RFO 1.6.1.5 Action A.1 which requires restoring RHR-SYS-A suppression pool spray subsystem to operable status within 7 days During RHR-P-2A pump start for entering Suppression Pool cooling mode, an isolated pressure switch (RHR-PS-19A) prevents the LPCS/RHR A ADS Permissive alarm from annunciating on P601. The CRS refers to Technical Specifications and determines that LCO 3.3.5.1 (Emergency Core Cooling System (ECCS) Instrumentation) Action A.1 applies which directs entry into the Condition referenced in Table 3.3.5.1-1 for the channel (Function 4.e) immediately (Condition G). ACTION G.2 directs restoring channel to operable status within 8 days.

Previous Columbia OE (Ref: AR-00049685 - Root Cause Analysis of RHR-PS-19A Isolation Mispositioning Event dated 4/1/2007) involved isolation of same pressure switch which was not discovered until RHR-P-2A was started and the ADS Permissive annunciator did not come in as expected.

Event 2 (TRG-2) Control rod 26-19 drifts out of the core. The ATC operator recognizes the rod drift and takes Immediate Actions to fully insert the control rod using the Continuous Insert pushbutton. The CRS enters ABN-ROD. When the Insert pushbutton is released, the control rod begins again to drift out of the core.

The ATC operator re-inserts the control rod full-in (and keeps the Continuous Inset pushbutton pressed) while the crew takes action to isolate the HCU for control rod 26-19 (TRG-6). The CRS declares control rod 26-19 inoperable. The CRS refers to Technical Specifications and determines that LCO 3.1.3 (Control Rod Operability) Action C.1 applies which requires rod 26-19 to be fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and Action C.2 which requires associated CRD (HCU) disarmed within four hours.

Event 3 (TRG-3) Standby Service Water Pump 1A (SW-P-1A) trips on motor winding overcurrent which requires Residual Heat Removal Pump 2A (RHR-P-2A) (currently in Suppression Pool Cooling) to be manually secured per ABN-SW. Standby Service Water System A (SW-SYS-A) not being available requires that the DG1 Diesel Engine Mode Selector be placed MAINT (Maintenance) effectively making DG1 inoperable.

The CRS declares SW-SYS-A and DG1 inoperable and refers to Technical Specifications and determines that the following applies:

NRC Scenario 3 Page 4 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017

LCO 3.7.1 Action B.1 which requires restoring SW-SYS-A to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

LCO 3.8.1 Action B.1 which requires performing SR 3.8.1.1 for operable offsite circuits (OSP-ELEC-W101 (Offsite Station Power Alignment Check )) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter

LCO 3.8.1 Action B.2 which requires declaring required feature(s) supported by DG 1, inoperable when the redundant required feature(s) are inoperable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of DG1 going inoperable concurrent with the inoperability of the redundant required feature(s)

LCO 3.8.1 Action B.3.1 which requires determining operable DGs are not inoperable due to common cause failure within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months - OR - LCO 3.8.1 Action B.3.2 which requires performance of SR 3.8.1.2 for operable DGs within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (if not performed in the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

LCO 3.8.1 Action B.4.1 which requires restoring DG1 to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of DG1 becoming inoperable AND within 6 days of failure to meet LCO (the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is more restrictive in this case) - OR -

LCO 3.8.1 Action B.4.2.1 which requires establishing risk management actions for the alternate AC sources within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND LCO 3.8.1 Action B.4.2.2 which requires DG1 to be restored to operable status within 14 days after being declared inoperable but in no case longer than 17 days from failure to meet LCO Evaluator note: Although several Technical Specification actions are involved, the CRS will only have to refer to LCO 3.7.1 Condition B and LCO 3.8.1 Condition B to find them.

Event 4 (TRG-4) Vibrations start to rise above the ALERT setpoint on Reactor Feed Pump (RFP) B as indicated by annunciator P840.A1.7-5 (Turbine B Vibration Trouble) and validated on (local) vibration instrument RFW-VBI-1B/XS/T1BXY (Turbine Radial Inboard Bearing Vibration). Feed pump bias is adjusted to minimize load on RFP B in an attempt to reduce vibration (which is unsuccessful). Vibration level will exceed the DANGER setpoint requiring Reactor Recirculation flow to be incrementally reduced in 1% to 5% step changes while monitoring vibration level. Vibration level remains above the DANGER setpoint even after Reactor Recirculation (RRC) flow has been reduced to 74 Mlbm/hr. RFP B is manually tripped per ARP direction. The CRS may direct tripping of RFP B before the flow reduction is complete if equipment damage is a concern. Following the trip, the high vibration annunciator will clear if crew attempts a local reset (TRG-7). As RPV level lowers due to the feed pump trip, both Reactor Recirculation (RRC) Pumps will runback to 30 Hz causing reactor power to stabilize at a lower level of

~68% power.

Event 5 (TRG-5) An earthquake (OBE) causes annunciator 851.S-1 5-1 (Operating Basis Earthquake Exceeded) to alarm. ABN-EARTHQUAKE is entered. Concurrently, a steam leak in the RCIC Pump Room develops resulting in RCIC Equipment Area high temperature alarms. PPM 5.3.1 (Secondary Containment Control) and ABN-HELB (Line Break) are entered on Reactor Building (RB) area high temperature. Crew attempts to isolate steam leak as directed by PPM 5.3.1 (Secondary Containment Control). Control Room notifies plant personnel of safety hazard and directs evacuation of affected areas. Neither RCIC-V-63 (RCIC Steam Supply Inboard Isolation) nor RCIC-V-8 (RCIC Turbine Steam Supply Isolation) will automatically close. Manual attempts to shut RCIC-V-63 and RCIC-V-8 are unsuccessful.

CRS enters PPM 5.1.1 (RPV Control) and directs a manual reactor scram before reaching the max safe operating temperature for the RCIC Pump room (CT #1). All control rods fully insert. The CRS may direct a reactor pressure reduction to 500 to 600 psig to reduce leak rate.

Event 6 Three (3) minutes after the scram, Main Steam Line A piping ruptures causing an unisolable steam leak.

The CRS re-enters PPM 5.3.1 (Secondary Containment Control) based on a second unisolable steam leak in Secondary Containment resulting in high Main Steam Tunnel temperature.

Event 7 Following the Main Steam Line A rupture, the outboard MSIVs fail to AUTO close due to failure of a logic relay but can be manually closed. MSIV 22A (MS-V-22A) fails to AUTO close due to mechanical failure. Inability to manually close MS-V-22A results in an unisolable leak into secondary containment.

NRC Scenario 3 Page 5 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Event 8 The CRS directs entry into PPM 5.1.3 (Emergency RPV Depressurization) once Main Steam Tunnel Temperature exceeds its max safe operating value of 330°F based on two secondary containment areas greater than max safe operating value. With a primary system discharging into secondary containment and area temperature exceeding maximum safe operating level in more than one area, Emergency Depressurization (ED) is initiated by opening seven (7) Safety Relief Valves (ADS preferred) within 10 minutes of second MSOT being exceeded. (CT #2) RPV level will be restored using Condensate Booster Pumps following Emergency Depressurization.

TERMINATION CRITERIA: The scenario will be terminated when an Emergency Depressurization has been performed and RPV level is being controlled in the prescribed band OR as directed by the Examination Team.

NRC Scenario 3 Page 6 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Critical Task Determination Measurable Performance Critical Task Safety Significance Cueing Performance Feedback Indicators CT #1 - With reactor If secondary Procedural direction The operator will All control rods will at power and with containment by PPM 5.3.1 (EOP manually scram fully insert.

primary system temperature exceeds for Secondary reactor by placing discharging into its maximum safe Containment Reactor Mode secondary operating value, Control) Step SC-14 Switch in containment, adequate core directs entering Shutdown.

manually scram cooling, containment PPM 5.1.1 (which reactor before any integrity, safety of requires placing area exceeds its personnel, or Reactor Mode maximum safe continued operability Switch in Shutdown) operating of equipment required before any area temperature. to perform EOP exceeds its flowchart actions can maximum safe no longer be assured. operating temperature.

(Ref: PPM 5.0.10 Rev 21, section 8.9.3 k.1))

CT #2 - With a The criteria of "2 or Procedural direction The operator will The valve light primary system more areas" identifies by PPM 5.3.1 (EOP manually open 7 indications for each discharging into the increase in for Secondary Safety Relief of the 7 Safety secondary parameter trend Containment Valves (ADS Relief Valves will containment and area as a wide spread Control) Step SC-15 preferred) to change from Green temperature problem which may directs Emergency emergency lit to Red lit when exceeding maximum pose a direct and Depressurizing depressurize the control switch is safe operating level in immediate threat to reactor when a RPV. taken to Open.

more than one area, secondary primary system initiate Emergency containment integrity, (RCIC) is Reactor pressure Depressurization (ED) equipment located in discharging into will lower in by opening seven (7) the secondary secondary response.

Safety Relief Valves containment, containment and (ADS preferred) within continued safe two or more area 10 minutes of second operation of the plant, temperatures are MSOT being and personnel both on exceeding their exceeded. and off site. maximum safe operating level.

Note: If the crew (Ref: PPM 5.0.10 Rev properly elects to 21, section 8.9.3 k.3))

invoke the EMERG DEPRESS is anticipated override in ppm 5.1.1 (RPV Control) and in doing so, the second maximum safe operating level is not exceeded, this Critical Task is considered to be met.

NRC Scenario 3 Page 7 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SIMULATOR SETUP Unload simulator (between each scenario)

Verify in ILC load Reload simulator Reset to IC-204 (reset, go to Run, reset again)

Load Scenario 3 Schedule file Load Scenario 3 Event file (if not loaded automatically)

Validate that there are no unexpected annunciators or parameters out of band Verify pump running magnets Verify keys REMOVED from RCIC-V-8 AND RCIC-V-64 Flag the following:

601.A11 1-3 601.A12 1-3 Place tagout on RCC-P-1B Protect the following:

RCC-P-1A and RCC-P-1C Have marked up copy of SOP-RHR-SPC for RHR A in Suppression Pool Cooling (with N/Ad steps initialed) for the crew at their pre-brief location (outside the simulator)

EQ machine tested at correct volume for OBE event (hardware volume knob at 100% with Windows volume at 85%)

NRC Scenario 3 Page 8 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SCHEDULE FILE

1

<ACTION>schedule Schedule\local.sch</ACTION>

</ITEM>

1

<ACTION>Event D:\NRC Scenario Support Files\2017 NRC SC-3.evt</ACTION>

</ITEM>

1

<ACTION>Insert malfunction BST-RHR014F to FAIL_TO_TRIP</ACTION>

<DESCRIPTION>RHR-PS-19A RHR LP A DISPRESS-ADS PERMISSIVE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOV-RCI012F to FAIL_AS_IS</ACTION>

<DESCRIPTION>RCIC-V-63 STM SUPPLY LINE INBOARD ISOL</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOV-RCI016F to FAIL_AS_IS</ACTION>

<DESCRIPTION>RCIC-V-8 STEAM SUPPLY LINE OUTBOARD I</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction RLY-NSF097F to FAIL_TO_TRIP</ACTION>

<DESCRIPTION>MS-RLY-K16 MS-V-28A,B,C,D TRIP ISOL LOGIC</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-RCC002 to FA_CTRL_FUS</ACTION>

<DESCRIPTION>CB-RCC-P-1B RCC-P-1B MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction AOV-RRS003F to FAIL_AS_IS</ACTION>

<DESCRIPTION>MS-V-22A &quotA&quot INBOARD MSIV</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC004-2619 to OUT on event 2</ACTION>

<DESCRIPTION>ROD 2619 DRIFTS</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RMC004-2619 to IN on event 6 delete in 5</ACTION>

NRC Scenario 3 Page 9 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017

<DESCRIPTION>ROD 2619 DRIFTS</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-SSW001 to TRIP on event 3</ACTION>

<DESCRIPTION>CB-SW-P-1A SW-P-1A MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction ANN-840A1G05 to ON on event 4</ACTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-FPT005B to 10.00000 on event 4</ACTION>

<DESCRIPTION>RFPT-B HIGH VIBRATION</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction ANN-840A1G05 to OFF on event 7</ACTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RWB001 to 0.16500 on event 5 delete in 20</ACTION>

<DESCRIPTION>EARTHQUAKE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RCI004 to 3000.00000 on event 5</ACTION>

<DESCRIPTION>RUPT STM LIN UPSTRM OF RCIC-V-45</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RRS006A after 180 to 70000.00000 in 500 on event 8</ACTION>

<DESCRIPTION>MSL-A BREAK IN THE STEAM TUNNEL</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RCI004 to 8000.00000 on event 8</ACTION>

<DESCRIPTION>RUPT STM LIN UPSTRM OF RCIC-V-45</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RRS006A to 140000.00000 in 300 on event 9</ACTION>

<DESCRIPTION>MSL-A BREAK IN THE STEAM TUNNEL</DESCRIPTION>

</ITEM>

</SCHEDULE>

NRC Scenario 3 Page 10 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT FILE

<EVENT>

</EVENT>

NRC Scenario 3 Page 11 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1

==

Description:==

Place RHR-SYS-A in Suppression Pool Cooling (LPCS/RHR A ADS Permissive fails to annunciate during pump start) (Tech Spec)

Event is initiated by the CRS as part of the shift turnover.

Time Position Applicants Actions or Behavior Examiner Note: Below evolution was pre-briefed by the crew before entering the simulator.

Steps 5.1.1 through 5.1.3 were previously completed.

CRS Directs BOP to place RHR-SYS-A into Suppression Pool Cooling mode using SOP-RHR-SPC (section 5.1 starting with step 5.1.4)

Comment:

Examiner Note: Following steps are from SOP-RHR-SPC (starting with step 5.1.4)

CRS 5.1.4: If RHR-SYS-A is required to be operable, then enter RHR-SYS-A as inoperable, but available, in the Plant Logging System (see below)

Comment:

Evaluates Technical Specifications and the LCS and determines the following actions apply:

LCO 3.5.1 Action A.1 which requires restoring RHR-SYS-A to operable status within 7 days Comment:

LCO 3.6.1.5 Action A.1 which requires restoring RHR-SYS-A drywell spray subsystem to operable status within 7 days Comment:

LCO 3.6.2.3 Action A.1 which requires restoring RHR-SYS-A suppression pool cooling subsystem to operable status within 7 days Comment:

RFO 1.6.1.5 Action A.1 which requires restoring RHR-SYS-A suppression pool spray subsystem to operable status within 7 days Comment:

NRC Scenario 3 Page 12 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1 (CONTINUED)

Examiner Note: Annunciator 601.A3 5-1 (ADS LPCS/RHR A Pump Permissive) will fail to alarm when RHR-P-2A started. May take crew a minute or so to validate proper pump starting response.

BOP 5.1.5: Starts RHR-P-2A (and verifies proper pump starting indications)

Breaker closed red indication above pump control switch

Pump current spikes then returns to normal

Verifies annunciator 601.A3 5-1 (ADS LPCS/RHR A Pump Permissive) alarms Notes alarm does not come in and informs the CRS Comment:

Examiner Note: With RHR pump running on min flow, it is expected the CRS will direct BOP to continue evolution while referring to Technical Specifications.

CRS Acknowledges report and directs BOP to continue evolution Comment:

BOP 5.1.6: Verifies RHR-FCV-64A opens during low flow conditions (approximately 800 gpm) (Minimum Flow Bypass) (H13-P601)

Comment:

NRC Scenario 3 Page 13 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1 (CONTINUED) 5.1.7: (2-handed operation) Throttles open RHR-V-24A to between 2500 and 7000 gpm, as determined by the CRS (Suppression Pool Cooling/Test Return) (H13-P601)

Comment:

5.1.8: Verifies RHR-FCV-64A closes (approximately 800 gpm)

Comment:

5.1.9: Verifies SW-P-1A running Comment:

5.1.10: If maximum cooling is desired, then closes RHR-V-48A (RHR-HX-1A Shell Side Bypass) (H13-P601)

Comment:

5.1.11: If minimum cooling is desired, then performs the following:

5.1.11.a: Throttles open RHR-V-48A (RHR-HX-1A Shell Side Bypass) (H13-P601)

5.1.11.b: Throttles closed RHR-V-3A (RHR-HX-1A Outlet)

(H13-P601)

Comment:

5.1.12: Maintains Suppression Pool temperature between 55°F and 90°F Comment:

5.1.13: Notifies HP that radiological conditions may have changed Comment:

NRC Scenario 3 Page 14 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1 (CONTINUED)

BOP Refers to ARP 601.A3 5-1 (ADS LPCS/RHR A Pump Permissive)

Comment:

Examiner Note: Following step is from ARP 4.601.A3 5-1 (ADS LPCS/RHR A Pump Permissive)

Examiner Note: Refer to the following:

The SOURCE (as shown) on the ARP page insinuates that RHR-PS-19A OR RHR-PS-16A is needed to the cause the alarm while in actuality, both are needed to cause the alarm. In this case, RHR-PS-19A will be found to be isolated (prior CGS OE) thereby preventing the alarm.

In any case, the CRS should refer to the applicable Technical Specification.

BOP 1. BOP refers CRS to Technical Specification 3.3.5.1 Comment:

CRS/BOP Dispatches operator or calls Work Week Manager to investigate the status of RHR pressure switches (RHR-PS-16A and 19A)

Comment:

BOOTH NOTE: If directed to investigate both pressure switches at once then, wait 2 minutes and make both reports at once.

BOOTH ROLEPLAY - If directed to investigate anything abnormal with RHR-PS-16A, wait 2 minutes then report Nothing abnormal found with RHR-PS-16A.

BOOTH ROLEPLAY - If directed to investigate anything abnormal with RHR-PS-19A, wait 2 minutes then report RHR-PS-16A was found isolated.

CRS Evaluates Technical Specification 3.3.5.1 and determines the following action applies:

LCO 3.3.5.1 Action G.2 - Restore channel (RHR-PS-19A) to operable status within 8 days Comment:

NRC Scenario 3 Page 15 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2

==

Description:==

Control rod (26-19) drifts out. Once inserted, releasing the continuous insert pushbutton allows the control rod to drift out again, requiring the control rod to be isolated (Tech Spec)

Event is initiated after RHR A is in Suppression Pool Cooling and associated Tech Spec call made (or as directed by the Exam team) and is activated using TRIGGER 2.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 2 Time Position Applicants Actions or Behavior BOP Responds to H13-P603, peer checks what rod is drifting, and acknowledges and resets the ROD DRIFT alarm (603.A7 5-7)

Comment:

Examiner Note: Following Immediate Action steps are from ABN-ROD (section 3.1)

ATC 3.1.2: Selects the drifting control rod (26-19)

Comment:

3.1.3: Performs the following:

3.1.3.a: Depresses the Continuous Insert pushbutton

3.1.3.b: Drives the control rod to its FULL IN position

3.1.3.c: Releases the Continuous Insert pushbutton

3.1.3.d: If the control starts to drift back out, then performs the following:

3.1.3.d.1): Depresses and Holds the Continuous Insert pushbutton 3.1.3.d.2): Informs CRS that control rod 26-19 needs to be isolated at its HCU Comment:

NRC Scenario 3 Page 16 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2 (CONTINUED)

CRS Enters ABN-ROD Comment:

BOP Directs field operator to hydraulically isolate control rod 26-19 per ABN-ROD step 4.1.2.a Comment:

BOOTH ROLEPLAY - If directed to hydraulically isolate control rod 26-19, wait 2 minutes then insert Trigger 1, report Control rod 26-19 hydraulically isolated per ABN-ROD, step 4.1.2.a.

Examiner Note: Following Subsequent Action steps are from ABN-ROD (section 4.1)

ATC 4.1.2.b: Once report received control rod 26-19 hydraulically isolated, releases the Continuous Insert pushbutton Comment:

4.1.3: Resets the Control Rod Drift annunciator using ROD DRIFT RESET pushbutton on H13-P603 (may be already reset by BOP)

Comment:

CRS 4.1.5: Notifies the SNE Comment:

4.1.6: Initiates (or direcs) a MON Run to verify acceptible thermal limits and preconditioning Comment:

4.1.7: Determines if the problem is generic in nature (CRS will call SNEs and station management to make this determination)

Comment:

4.1.8: Refers to Technical Specification 3.1.3 (see next page)

Comment:

4.1.9 & 4.1.10: Performed by calling for help external to the Main Control Room (Event 3 may occur prior to the CRS making these notifications).

Comment:

NRC Scenario 3 Page 17 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2 (CONTINUED)

Examiner Note: Management expectation is to declare the control rod INOP (even though it is not considered INOP per Technical Specifications).

CRS Evaluates Technical Specifications and determines the following actions apply:

LCO 3.1.3 C.1 - Insert control rod 26-19 within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Comment:

LCO 3.1.3 C.2 - Disarm control rod 26-19 HCU within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Comment:

NRC Scenario 3 Page 18 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3

==

Description:==

Standby Service Water Pump 1A (SW-P-1A) trips which requires RHR Pump 2A (RHR-P-2A) (currently in Suppression Pool Cooling) to be manually secured (Tech Spec)

Event is initiated after control rod 26-19 HCU has been isolated and associated Tech Spec call made (or as directed by the Exam team) and is activated using TRIGGER 3.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 3 Examiner Note: This event starts with several BISIs (Bypass and Inoperable Status Indicators) common with SW-P-1A to illuminate which causes several annunciators to alarm. The main annunciator the BOP should pursue is at H13-P840 (840.A5 2-2 (SW Pump A Motor OL/ Gnd))

Time Position Applicants Actions or Behavior BOP Amongst all annunciators in alarm, recognizes that a trip of Service Water Pump 1A has occurred Comment:

Silences lower priority annunciators and refers to ARP 840.A5 2-2 (SW Pump A Motor OL/ Gnd)

Comment:

Examiner Note: Following steps are from ARP 840.A5 2-2 (SW Pump A Motor OL/ Gnd)

BOP 1: If SW-P-1A tripped then perform the following:

1.d: Refers CRS to ABN-SW

1.e: Informs CRS the DG1 Diesel Engine Mode Selector needs to be placed in MAINT (effectively make DG inoperable)

1.f: Informs CRS to complete OSP-ELEC-W101 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Comment:

4: Refers CRS to Technical Specification 3.7.1 Comment:

CRS Enters ABN-SW Comment:

Directs DG1 Diesel Engine Mode Selector to be placed in MAINT Comment:

BOOTH ROLEPLAY - If directed to place DG 1 Diesel Engine Mode Selector in MAINT, wait 2 minutes then insert Trigger 8, report DG 1 Diesel Engine Mode Selector is in MAINT.

BOOTH ROLEPLAY - If directed to investigate why SW-P-1A tripped, wait 5 minutes then report SW-P-2A is very hot to the touch. Its breaker was found tripped.

NRC Scenario 3 Page 19 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3 (CONTINUED)

Examiner Note: Following steps are from ABN-SW (section 4.2)

CRS 4.2.1: Places DG1 in MAINT (may have been previously performed)

Comment:

Examiner Note: When CRS discusses need to complete OSP-ELEC-W101 (Offsite Station Power Alignment Check ) inform them that it will be performed by another RO.

CRS 4.2.3: Directs OSP-ELEC-W101 completed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of DG1 being declared inoperable Comment:

BOP 4.2.5: IF SW A flow is lost (non LOCA), and Adequate Core Cooling and Containment Integrity is assured, then secures the following operating pump:

RHR-P-2A Comment:

Examiner Note: CRS may direct the BOP to exit the Suppression Pool Cooling lineup on RHR Loop A.

Examiner Note: Following steps are from SOP-RHR-SPC (section 5.2)

BOP 5.2.1: Notifies HP that the actions to stop Suppression Pool Cooling may potentially change radiological conditions.

Comment:

5.2.2: Verifies RHR-V-3A open Comment:

(Two handed operation) 5.2.3: Verifies RHR-V-24A open Comment:

(Two handed operation) 5.2.4: Closes RHR-V-24A Comment:

NRC Scenario 3 Page 20 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3 (CONTINUED)

BOP 5.2.5: Stops RHR-P-2A (may already be stopped)

Comment:

5.2.6: Verifies RHR-V-64A closed Comment:

Examiner Note: Crew will not have time to complete step below.

BOP 5.2.7: Verifies RHR Loop A is in Standby Status per SOP-RHR-STBY Comment:

Examiner Note: Continuing steps from ABN-SW (section 4.2)

CRS 4.2.12: Enters SW-SYS-A and DG-SYS-A as inoperable in the Plant Logging System (see below)

Comment:

Evaluates Technical Specifications and the LCS and determines the following actions apply:

LCO 3.7.1 Action B.1 which requires restoring SW-SYS-A to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Comment:

LCO 3.8.1 Action B.1 which requires performing SR 3.8.1.1 for operable offsite circuits (OSP-ELEC-W101 (Offsite Station Power Alignment Check ))

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter (discussed earlier)

Comment:

LCO 3.8.1 Action B.2 which requires declaring required feature(s) supported by DG 1, inoperable when the redundant required feature(s) are inoperable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of DG1 going inoperable concurrent with the inoperability of the redundant required feature(s)

Comment:

LCO 3.8.1 Action B.3.1 which requires determining operable DGs are not inoperable due to common cause failure within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months - OR -

LCO 3.8.1 Action B.3.2 which requires performance of SR 3.8.1.2 for operable DGs within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (if not performed in the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

Comment:

Examiner Note: Technical Specification entries continued on next page.

NRC Scenario 3 Page 21 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3 (CONTINUED)

CRS LCO 3.8.1 Action B.4.1 which requires restoring DG1 to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of DG1 becoming inoperable AND within 6 days of failure to meet LCO (the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is more restrictive in this case) - OR -

LCO 3.8.1 Action B.4.2.1 which requires establishing risk management actions for the alternate AC sources within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND LCO 3.8.1 Action B.4.2.2 which requires DG1 to be restored to operable status within 14 days after being declared inoperable but in no case longer than 17 days from failure to meet LCO Comment:

Directs the following systems to be Protected per PPM 1.3.83 (Protected Equipment Program) Attachment 7.1 (based on SW Pump A and DG1 unavailability)

DG-SYS-B

SW-SYS-B

HPCS (P-1, DG, & SW)

RHR-SYS-B

RHR-SYS-C

E-TR-S

E-TR-B

ADS-SYS-B

H13-P800 Bd. C Control and Indication areas Comment:

NRC Scenario 3 Page 22 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4

==

Description:==

Reactor Feed Pump (RFP) B vibrations rise requiring RRC Flow reduction and manual trip of the B RFP Event is initiated after RHR Pump A is secured and associated Tech Spec call made (or as directed by the Exam team) and is activated using TRIGGER 4.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 4 Time Position Applicants Actions or Behavior ATC Acknowledges annunciator 804.A1 7-5 (TURB B VIB TROUBLE) and informs CRS Comment:

Examiner Note: Vibration levels for RFP B Turbine Radial Inboard Bearing (see ARP page below) are considered at the ALERT setpoint when it reaches 3 mls and at the DANGER setpoint when it reaches 4.5 mls.

Examiner Note: Following steps are from ARP 804.A1 7-5 (TURB B VIB TROUBLE)

BOP 1: Directs field operator to investigate source of the vibration using RFW-VMP-1 on TB 441 Elev Comment:

BOOTH ROLEPLAY - If sent to investigate high vibrations on vibration panel, wait 2 minutes then report RFP B Turbine Radial Inboard Bearing reads 3.1 mls up slow. All other bearing vibration levels are normal.

BOOTH ROLEPLAY - If sent to investigate high vibrations locally at turbine, wait 2 minutes then report RFP B Turbine sounds slightly different than what Im used to hearing.

BOP 2: Verifies reported vibration is above is above the applicable alarm setpoint (3 mls)

Comment:

NRC Scenario 3 Page 23 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4 (CONTINUED)

Examiner Note: Bearing vibration will continue to rise above the DANGER level (4.5 mls) even after ARP step 4 is completed below.

CRS Directs BRO to adjust lead Feed Pump bias to minimize load on RFP B (as described below)

BOP 4: If any value is above the ALERT setpoint, but below the DANGER setpoint, then adjust the lead Feed Pump bias to minimize load on the affected Reactor Feedwater Pump/Turbine (B) as follows:

Raise RFW-P-1A Speed using RFT-COMP-1or RFT-COMP-2 (Pump Control Screen)

Lower RFW-P-1B Speed using RFT-COMP-1or RFT-COMP-2 (Pump Control Screen)

Verify feed pump speed controllers are stable and not hunting Comment:

Directs field operator to report current vibrations on the RFP B Turbine Radial Inboard Bearing Comment:

BOOTH ROLEPLAY - If directed to report current vibrations on vibration panel BEFORE BIAS ADJUSTMENT MADE, then report RFP B Turbine Radial Inboard Bearing reads 3.6 mls up slow.

BOOTH ROLEPLAY - If directed to report current vibrations on vibration panel AFTER BIAS ADJUSTMENT MADE, then report RFP B Turbine Radial Inboard Bearing reads 4.7 mls up slow.

BOOTH ROLEPLAY - If directed to report current vibrations on any other bearing report All other bearing vibration levels are normal.

BOOTH ROLEPLAY - If directed to report when bearing vibration level reaches the DANGER setpoint of 4.5 mls then only report AFTER BIAS ADJUSTMENT MADE, that the RFP B Turbine Radial Inboard Bearing reads 4.5 mls up slow.

ATC 5: If any indicated value is sustained at or above the DANGER setpoint following feed pump load reduction via bias adjustment, then reduce Reactor Power with Reactor Recirculation flow incrementally, within the capacity of one Reactor Feed Pump, to reduce vibrations (Consider 1-5%

step changes in Reactor Power while monitoring vibrations)

Comment:

SRO Directs ATC to reduce reactor power with RRC flow using (1-5%) step changes while monitoring bearing vibrations Comment:

NRC Scenario 3 Page 24 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4 (CONTINUED)

Examiner Note: Intent is to report bearing vibration level above the DANGER (4.5 mls) setpoint throughout the flow reduction. Vibration level will continue to trend up slowly but once power has been reduced approximately 5%, a call will come in from the field that the B RFP is vibrating excessively and should be tripped immediately.

Examiner Note: For credit (as written) the ATC operator has to be the one to trip the B RFP CRS 6: If any indicated value is sustained at or above the DANGER setpoint following feed pump load reduction via bias adjustment, then reduce Reactor Power with Reactor Recirculation flow incrementally, within the capacity of one Reactor Feed Pump, to reduce vibrations (Consider 1-5%

step changes in Reactor Power while monitoring vibrations)

CRS directs power reduction Comment:

ATC Reduces reactor power with RRC flow as follows:

Notes reactor power and/or Main Generator output (MWe)

Refers to SOP-RRC-FLOW-QC quick card and performs the following per section 2.1 (Reactor Power change with RRC Flow controllers in Auto):

2.1.1: Monitors fuel pre-conditioning limits (per PPM 9.3.18) while changing reactor power

2.1.2: Lowers RRC flow using RTC-M/A-R675 (Master controller) as necessary (below sub-steps are good practice steps)

Observes lowering frequency on both RRC pumps Verifies reactor power lowers and RFPs respond to maintain RPV level

2.1.3: Verifies total core flow is less than 105%

2.1.4: Verifies RRC Loop A and B is less than 57.5 Mlbm/hr

2.1.5: Notifies CRS when change in power is complete Comment:

BOP May peer check ATC reactivity manipulation Maintains communication with field operator to monitor vibration levels Comment:

BOOTH ROLEPLAY - If directed to report current vibrations during power reduction, then report RFP B Turbine Radial Inboard Bearing reads 5.2 mls up slow.

BOOTH ROLEPLAY - Once reactor power has been lowered approximately 5%, make the following report (make it sound urgent): The B RFP is vibrating excessively and should be tripped immediately.

NRC Scenario 3 Page 25 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4 (CONTINUED)

CRS Directs tripping RFW-P-1B per step 6 of ARP (840.A1 7-5) or out of concerns for equipment safety Comment:

ATC Trips the B Feed Turbine Comment:

Monitors for RRC Runback to 30Hz (both pumps)

Comment:

Verifies RFP A responds properly to transient in controlling RPV level Comment:

When plant stabilizes, provides reactor power, pressure and level to CRS Comment:

BOP Make plant announcement concerning reactor power and RFP status Comment:

Follows up with ARP 840.A1 1-5 (TURB B TRIP) (as time permits)

1: Verifies proper RRC Runback occurred

2: Verifies MS-V-172B closed (RFW-P-1B High Press Stop Valve)

3: Verifies BS-V-60B closed (RFW-P-1B Low Press Stop Supply)

4: Verifies the following open:

BS-V-44B (BS-V-60B Body Drain)

BS-V-45B (RFW-DT-1B Stage Drain)

MS-V-142B (RFW-P-1B HP Stop Above Seat Drain)

5: Verifies RFW-FCV-2B is closed (Pump Minimum Flow)

6: When RFW-DT-1B slows to less than 1 rpm, and lube oil is available, then place RFW-DT-1B Turning Gear Control to Auto Engage Comment:

NRC Scenario 3 Page 26 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4 (CONTINUED)

BOP May direct field operator to reset vibration panel alarms (to clear MCR annunciator)

Comment:

BOOTH ROLEPLAY - If directed to reset local vibration panel alarm, then activate TRG-7 (MCR annunciator will clear)

NRC Scenario 3 Page 27 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5

==

Description:==

Operating Bases Earthquake causes a steam leak in the RCIC Pump Room with Failure of RCIC-V-8 and RCIC-V-63 to fully close (preventing RCIC leak isolation). Manual scram inserted before first secondary containment max safe operating temperature is reached.

Event is initiated after the B RFP has been tripped and the plant is stabilized (or as directed by the Exam team) and is activated using TRIGGER 5.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 5 Examiner Note: First annunciator (601.A3 5-7 (LEAK DET RCIC EQUIP AREA DT HIGH))

indicative of RCIC steam leak does occur not until about 3 minutes after the OBE.

Examiner Note: RCIC maximum safe operating temperature (1st Max Safe) will not be reached for at least the next 15 minutes.

Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 851.S1 5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) and informs CRS Comment:

Examiner Note: Following steps are from ARP 4.851.S1 5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED)

BOP 1: Identifies alarm on H13-P823 (Board L) (BOP goes to the back to check Board L indications) - Reports all red and all amber shock lights illuminated (indication of seismic strength) 2: Refers CRS to ABN-EARTHQUAKE Comment:

CRS Enters ABN-EARTHQUAKE Comment:

Examiner Note: Following steps are from ABN-EARTHQUAKE (due to higher plant priorities only certain actions will be listed here - Crew may not get to all of them)

CRS 4.2: Verify adequate systems are available for safe shutdown and cooldown of reactor (will verify equipment operability against turnover sheet)

Comment:

4.4: Discusses need to initiate controlled reactor shutdown per PPM 3.2.1 Comment:

BOP 4.7: Makes announcement per ABN-EARTHQUAKE step 4.7 Comment:

NRC Scenario 3 Page 28 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

BOP 4.8: Directs SAS (Secondary Alarm Station) to repeat above announcement on the Alternate Security/ Area Wide and Security radio channels Comment:

BOOTH ROLEPLAY - If directed to repeat announcement as SAS, then repeat back direction (BOP does this by talking over the chain to the Booth Operator or calling the Booth)

CRS 4.10: Directs crew to check for any indications of RCS leakage or any other equipment issues Comment:

BOP 4.11: Directs field operator to check the Spent Fuel Pool for damage Comment:

BOOTH ROLEPLAY - If directed to check Spent Fuel Pool for damage, wait 3 minutes then report There are no signs of damage to the Spend Fuel Pool.

ATC Actively monitors reactor power, pressure and level for abnormalities Comment:

4.1.14 Checks neutron monitoring system for proper operation and changes Comment:

CRS 4.1.15: Directs initial plant inspection Comment:

Examiner Note: RCIC steam leak starts. Annunciator 601.A3 5-7 (LEAK DET RCIC EQUIP AREA DT HIGH) comes in first quickly followed by annunciators 601.A3 1-4 & 601.A2 1-2 (LEAK DET RCIC EQUIP AREA TEMP HI-HI). BOP should address the higher priority alarms.

BOP Acknowledges annunciators 601.A3 1-4 / 601.A-2 1-2 (Leak Detection RCIC Equip Area Hi-Hi) and informs CRS Comment:

Examiner Note: Following steps from ARPs 601.A3 1-4 / 601.A2 1-2 (Leak Detection RCIC Equip Area Hi-Hi)

BOP 1: Identifies alarming point(s) on LD-MON-1A on H13-P632 Comment:

2: Compares alarming point(s) on LD-MON-1B on H13-P642 Comment:

NRC Scenario 3 Page 29 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

BOP 3: Informs CRS of alarming points and trend (RCIC Pump Room DT >

50°F) (which is a PPM 5.3.1 (Secondary Containment Control) entry condition)

Comment:

CRS Enters PPM 5.3.1 (Secondary Containment Control) on RCIC Pump Room DT > 50°F (Table 22)

Comment:

BOP 4 & 5: Determines the status of the following RCIC components:

RCIC-V-63 (should be closed but remains intermediate)

RCIC-V-76 closed (already closed)

RCIC-V-8 (should be closed but remains intermediate)

RCIC Turbine (should be tripped and is tripped)

Reports to CRS that RCIC-V-63 and RC-V-8 did not fully close (indicate intermediate) and that RCIC did not isolate Comment:

6: May direct field operator to SAFELY investigate possible steam leak in the RCIC Pump Room (before becoming too large)

Comment:

BOOTH ROLEPLAY - If directed to investigate RCIC Pump Room leak, wait 2 minutes then report The RCIC Pump Room appears unsafe to enter based on high temperature and humidity.

BOP 8 & 10: Refers CRS to ABN-HELB and to Technical Specification 3.3.6.1 Comment:

Examiner Note: Following steps are from ABN-HELB (Line Break) (section 4.2)

CRS 4.2.1: Directs evacuation of all non-emergency personnel from the Reactor Building This will be directed to the BOP operator who will refer to the blue sheets and PA binder by the PA speaker Comment:

BOP Makes evacuation announcement as directed Comment:

NRC Scenario 3 Page 30 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

CRS Directs the BOP to obtain the keys for RCIC-V-8 and RC-V-63 and attempt to manually shut them (May also be directed from 5.3.1, SC-9, below)

Comment:

BOP With keys in hand, inserts one key into keylock switch for RCIC-V-63 and takes it to close Inserts second key into keylock switch for RCIC-V-8 and takes it to close Reports to CRS that RCIC could not be manually isolated (RCIC-V-63 and RCIC-V-8 did not close)

Comment:

CRS Requests assistance in getting RCIC-V-8 closed (more accessible than RCIC-V-63) although any attempt will be unsuccessful Comment:

BOOTH ROLEPLAY - If directed to close RCIC-V-8 locally, wait 20 minutes then report Im here with maintenance. We could not close RCIC-V-8. It appears mechanically bound.

CRS Establishes a Key Plant Parameter for RCIC Pump Room temperature below the Max Safe value of 200°F (see below)

Comment:

NRC Scenario 3 Page 31 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

BOP Trends RCIC Pump Room temperature as Key Plant Parameter and notifies CRS when value reached Comment:

CT #1 - With reactor at power and with primary system discharging into secondary containment, manually scram reactor before any area exceeds its maximum safe operating temperature.

CRS When notified Key Plant Parameter has been reaced, updates the crew on plant conditions then enters PPM 5.1.1 (RPV Control)

Comment:

Directs ATC to scram the reactor Comment:

Examiner Note: Following steps are Immediate Actions from PPM 3.3.1 (Reactor Scram)

ATC 6.1.1: Places Reactor Mode Switch to Shutdown Comment:

6.1.2: Monitors reactor power, pressure and level Comment:

6.1.5: Inserts SRM and IRM monitors (detectors)

Comment:

After above three steps ATC makes scram report to CRS:

Mode switch is in Shutdown

APRMs are downscale

RPV pressure is (value and trend)

RPV level is (value and trend)

EOP entry on low RPV level (and possibly high Drywell pressure)

Comment:

CRS Repeats back scram report Comment:

ATC 6.1.6: After CRS repeat back, reports all control rods are IN Comment:

NRC Scenario 3 Page 32 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

CRS Enters PPM 5.1.1 (RPV Control) on low RPV level (+13 inches)

Comment:

CRS Directs BOP to verify containment isolations occurred at +13 inches Comment:

BOP Verifies +13 inch containment isolation valves closed on the Isolation Control panel:

RHR-V-8, RHR-V-9

RHR-V-40, RHR-V-49

RHR-V-60A, RHR-V-60B

RHR-V-75A, RHR-V-75B Comment:

Examiner Note: Following steps are Subsequent Actions from PPM 3.3.1 (Reactor Scram)

ATC 6.2.5.a: Verify Recirc pumps have run back to 15 Hz Comment:

6.2.6: Range down on IRMs, as necessary, to follow power decrease Comment:

BOP 6.2.7: Make PA announcement for reactor scram Comment:

ATC 6.2.8: Transfers level control to RFW-FCV-10A/B per SOP-RFW-FCV-QC quick card Comment:

BOP 6.2.9: If necessary (with Main Generator load < 50 MWe):

If Main Turbine did not trip - simultaneously depress both Emerg Trip pushbuttons (H13-P820)

If Main Generator did not trip -depress either Unit Emergency Tip pushbutton or Unit Overall Trip pushbutton (H13-P800)

Verify power transfer to Startup Transformer (TR-S)

Comment:

NRC Scenario 3 Page 33 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

Examiner Note: Following steps are from SOP-RFW-FCV-QC (Transfer RPV Level Control to RFW-FCV-10A/10B - Quick Card).

ATC 2.1.1: (2-handed operation) Starts closing RFW-V-112A and RFW-V-112B Comment:

2.1.2: Starts opening RFW-V-118 Comment:

2.1.3: Verifies RFW-V-109 is closed Comment:

2.1.4: (2-handed operation) Verifies RFW-V-117A and RFW-V-117B open Comment:

2.1.5: Verifies RFW-LIC-620 is in Manual (V selected for Valve position demand with 0 output)

Comment:

2.1.6: IF Reactor Feed Pump(s) (RFP) are operating, then performs the following:

2.1.6.a: Verifies RFPs have ramped down in speed

2.1.6.b: Places RFW-P-1B in MDEM mode

2.1.6.c: Places RFW-P-1B in MDEM mode

2.1.6.d: Controls turbine speed as required

2.1.6.e: If desired, then places RFW-FCV-2A(B) in Manual and slowly open to approximately 80%

Comment:

2.1.7: Verifies RFW-V-112A and RFW-V-112B are fully closed Comment:

2.1.8: Verifies RFW-V-118 is fully open Comment:

NRC Scenario 3 Page 34 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

ATC 2.1.9: IF Reactor Feed Pump(s) (RFP) are operating, then adjusts the running RFP speed to establish ~ 200 psid across RFW-FCV-10A & 10B using either Feedwater touch screen (H13-P840)

Comment:

2.1.10: Adjusts RFW-LIC-620 manual output to control RPV level Comment:

2.1.12: If unable to control RPV level with RFW-FCV-10A/B, then considers throttling RFW-V-109 or RFW-V-118 to control RPV level Comment:

CRS Directs BOP to maintain RPV pressure band from 800 to 1050 psig using DEH in automatic (may direct BOP to establish a new pressure band of 500-600 psig with DEH in automatic to reduce the driving head of the leak into secondary containment)

Comment:

BOP Lowers RPV pressure if directed using SOP-DEH-QC (Main Turbine DEH Operations Quick Card):

2.1.1a: Selects PRESSURE TARGET

2.1.1b: Enters desired pressure

2.1.1c: Selects OK

2.1.1.d: If change in pressure rate is desired:

1: Selects PRESSURE RATE 2: Enters desired PRESSURE RATE 3: Selects OK

2.1.1.e: Selects GO

2.1.1.f: Selects YES

2.1.1.g: Verifies pressure demand and throttle pressure change at the pressure rate.

Comment:

NRC Scenario 3 Page 35 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6

==

Description:==

Steam leak develops in the Main Steam Tunnel Event is activated at the beginning of the scenario and is realized 3 minutes after the Reactor Mode Switch is taken to Shutdown.

Examiner Note: First annunciator (601.A3 3-8 (LEAK DET MSL TUNNEL DT HIGH)) indicative of a Main Steam Line break does not occur until about 6 minutes after the scram.

Examiner Note: Annunciator 601.A3 3-8 (LEAK DET MSL TUNNEL DT HIGH) comes in first quickly followed by annunciators 601.A3 1-7 & 601.A2 3-1 (LEAK DET MSL TUNNEL TEMP HIGH). BOP should address the higher priority alarms.

BOP Acknowledges annunciators 601.A3 1-7 / 601.A2 3-1 (LEAK DET MSL TUNNEL TEMP HIGH) and informs CRS Comment:

Examiner Note: Following steps from ARPs 601.A3 1-7 / 601.A2 3-1 (LEAK DET MSL TUNNEL TEMP HIGH)

BOP 1: Identifies alarming point(s) on LD-MON-2A on H13-P632 Comment:

2: Compares alarming point(s) to temperatures on LD-MON-2B on H13-P642 and recognizes steam leak appears to be on MSL A Comment:

3: Informs CRS of alarming points and trend on MSL A (MSL Tunnel >

80°F) (which is a PPM 5.3.1 (Secondary Containment Control) entry condition)

Comment:

CRS Re-enters PPM 5.3.1 (Secondary Containment Control) on MSL Tunnel DT

> 80°F (Table 22)

Comment:

BOP 4 & 5: Determines the status of the NSSSS Group 1 isolation which should have occurred - See Event 7 (next page)

Comment:

NRC Scenario 3 Page 36 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7

==

Description:==

MS-V-22A and MS-V-28A through D fail to automatically close (MS-V-28A through D can be closed manually but does not isolate leak)

Event is activated at the beginning of the scenario and is realized when the MSIVs do not close as expected.

BOP Recognizes that MS-V-22A and MS-V-28A through D failed to automatically close based on Group 1 isolation signal Comment:

Attempts to close MS-V-22A and MS-V-28A through D and notes that all valves closed except for MS-V-22A Comment:

Informs CRS of the failure of MS-V-22A and MS-V-28A through D to auto close and that after manual close attempt all valves closed with exception of MS-V-22A Comment:

Takes MSIV switches for those MSIVs that automatically shut to the Closed position Comment:

Informs crew that pressure control is with SRVs (using previously provided band)

Comment:

Reports that Main Steam Tunnel temperature continues to rise Comment:

NRC Scenario 3 Page 37 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 8

==

Description:==

Emergency Depressurization (PPM 5.1.3) is performed when two areas exceed their max safe operating temperature CRS Establishes a Key Plant Parameter for Main Steam Tunnel temperature of 320°F (Max Safe value) (see below)

Comment:

BOP Trends MSL Tunnel temperature as Key Plant Parameter and notifies CRS when value reached Comment:

CRS Directs second operator verify max safe temperature in two areas has been exceeded Comment:

ATC Verifies max safe temperature in two areas has been exceeded Comment:

NRC Scenario 3 Page 38 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 8 (CONTINUED)

CT #2 - With a primary system discharging into secondary containment and area temperature exceeding maximum safe operating level in more than one area, initiate Emergency Depressurization (ED) by opening seven (7) Safety Relief Valves (ADS preferred) within 10 minutes of second MSOT being exceeded.

Note: If the crew properly elects to invoke the EMERG DEPRESS is anticipated override in ppm 5.1.1 (RPV Control) and in doing so, the second maximum safe operating level is not exceeded, this Critical Task is considered to be met.

CRS When verified Key Plant Parameter has been reached, updates the crew on plant conditions, exits the pressure leg of PPM 5.1.1 (RPV Control) via override then enters PPM 5.3.1 (Emergency RPV Depressurization)

Comment:

Determines a high Drywell pressure signal is not sealed in Comment:

Determines Wetwell level is > 17 feet Comment:

Directs 7 SRVs be opened (ADS preferred) (ADS SRVs are those with the red stripe on left side of their nameplate)

Comment:

BOP Opens 7 SRVs (ADS preferred) as directed while verifying proper containment response as each is opened and reports completion to CRS Comment:

CRS Directs pumps not required for Adequate Core Cooling be stopped from injecting Comment:

Directs RPV level band of -50 to +54 inches Comment:

ATC Maintains RPV level as required to maintain RPV level band Comment:

TERMINATION CRITERIA: The scenario will be terminated when an Emergency Depressurization has been performed and RPV level is being controlled in the prescribed band OR as directed by the Examination Team.

NRC Scenario 3 Page 39 of 40

Appendix D NRC Scenario No. 3 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 TURNOVER Initial Conditions:

Columbia is operating at 85% power due to economic dispatch

Safety Relief Valve 2C (MS-RV-2C) is known to be leaking

Suppression Pool high temperature alarms (601.A11.1-3 and 601.A12.1-3) have just annunciated

Reactor Closed Cooling (RCC) Pump 1B is tagged out for planned maintenance

RCC-P-1A and RCC-P-1C are protected Shift Turnover:

After shift turnover place RHR-P-2A in Suppression Pool Cooling and allow SW-P-1A to auto start per SOP-RHR-SPC (section 5.1) - Steps 5.1.1 through 5.1.3 are complete

The pre-evolution brief has been completed and operators are stationed near both pumps

Maintain RHR-P-2B in operation for the next three (3) hours to satisfy pump PMT requirements NRC Scenario 3 Page 40 of 40

INSTRUCTIONAL COVER SHEET PROGRAM TITLE OPERATIONS TRAINING COURSE TITLE COLUMBIA GENERATING STATION SIMULATOR EXAMINATION Withdraw Control Rods during Startup; REA-FN-1B Trip requiring PPM LESSON TITLE 5.3.1 Entry and SGTS Start (TS); IRM A Upscale Failure with Half Scram; Loss of SL-11 (Re-energized from Alternate Source); RCIC-P-1 Coupling Found Broken; RHR-P-2B Suction Rupture (Lowering WW Level); SW-V-29 Fails to Auto Open; FDR-V-607 Fails to Close; Manual Scram on Low WW Level (Mode Switch Failure - Scram Pushbuttons Successful); ED performed on Low WW Level (One ADS Valve Fails to Open)

LENGTH OF LESSON 1.5 Hours Lesson Plan PQD Code Rev. No.

Simulator Guide PQD Code LO001859 Rev. No. 0 JPM PQD Code Rev. No.

Exam PQD Code Rev. No.

DIVISION TITLE Nuclear Training DEPARTMENT Operations Training PREPARED BY Dave E. Crawford DATE 12/22/16 REVISED BY DATE VALIDATED BY DATE TECHNICAL REVIEW DATE INSTRUCTIONAL REVIEW DATE APPROVED DATE Operations Training Manager NRC Scenario 4 (Spare)

Page 1 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Facility: Columbia Generating Scenario No.: 4 Op Test No.: 1 Station Examiners: Operators:

The reactor is in Mode 2 during reactor startup. Reactor is critical at 3% power with RPV pressure at Initial Conditions: approximately 300 psig. Reactor Building Exhaust Fan 1A (REA-FN-1A) is out of service for extended maintenance.

Withdraw control rods as required to establish and maintain Bypass Valves approximately 20% open.

Turnover:

Continuous rod withdrawal permitted.

Critical Tasks:

CT-1 Manually scram the reactor before wetwell level drops below 19 feet 2 inches.

When wetwell level cannot be maintained above 19 feet 2 inches, initiate emergency depressurization by CT-2 opening seven (7) Safety Relief Valves (ADS preferred) within 10 minutes of wetwell level lowering to 19 feet 2 inches.

Malf.

Event No. Event Type* Event Description No.

Withdraw control rods as required to establish and maintain the bypass 1 N/A R (ATC) valves approximately 20% open C (BOP,SRO) Trip of REA-FN-1B results in a high reactor building pressure and entry 2 TRG-2 TS (SRO) into PPM 5.3.1 (EOP - Secondary Containment Control) (Tech Spec) 3 TRG-3 I (ATC,SRO) IRM A fails upscale resulting in a half scram Differential current lockout of transformer (TR-1/11) results in a loss of 4 TRG-4 C (BOP,SRO)

SL-11 which requires bus to be re-energized from alternate source C (ATC**,SRO) 5 N/A RCIC-P-1 coupling discovered broken (Tech Spec)

TS (SRO)

M (ALL) Failure of the RHR-P-2A suction line results in lowering wetwell level 6 TRG-6 FDR-V-607 fails to auto close due to a failed level switch (which allows flooding to continue into RCIC Pump Room). Cannot be closed manually SW-V-29 fails to auto open when HPCS-P-2 is started for wetwell 7 N/A C (BOP) makeup Reactor mode switch fails to scram reactor, requiring use of manual 8 N/A C (ATC) scram pushbuttons to scram reactor prior to wetwell level lowering to 19 feet 2 inches (CT #1)

Prior to wetwell level going below 19 feet 2 inches, the crew determines

--- that wetwell level cannot be maintained 19 feet 2 inches and initiates 9 N/A RPV Emergency Depressurization (ED) with 7 SRVs opened (CT #2)

One ADS SRV (MS-RV-4D) fails to open requiring manually opening one C (BOP) non-ADS SRV (CT #2)

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (TS)Technical Specifications

- Normally assigned to BOP. NRC Evaluator will have to direct CRS to use ATC.

NRC Scenario 4 (Spare)

Page 2 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-1 Columbia Generating Station ILC NRC Exam - February, 2017 Target Quantitative Attributes Actual Description SW-V-29 fails to auto open; Mode switch failure; ADS Malfunctions after EOP entry (1-2) 3 SRV fails to open Fan REA-FN-1B trip; IRM A trip with half scram; Loss Abnormal events (2-4) 3 of SL-11 Major transients (1-2) 2 Primary containment failure; Manual scram PPM 5.1.1 (RPV Control); PPM 5.2.1 (Primary EOPs entered/requiring substantive actions (1-2) 3 Containment Control); PPM 5.3.1 (Secondary Containment Control);

EOP contingencies requiring substantive actions (0-2) 1 PPM 5.1.3 (Emergency RPV Depressurization)

EOP-based Critical Tasks (2-3) 2 See Critical Task Determination table Trigger Evaluator How Purpose Malfunction Numbers (TRG-x) Directed Triggered TRG-2 YES Manually Event Initiator PMP-SCN010S TRG-3 YES Manually Event Initiator MAL-NIS002A TRG-4 YES Manually Event Initiator ANN-800C3A02; BKR-EPS001; BKR-EPS004 MAL-RHR001; XMT-PCN006A; XMT-PCN007A; XMT-TRG-6 YES Manually Event Initiator PCN003A; XMT-PCN004A TRG-7 Manually Event Initiator BKR-RHR001 Initial Condition BKR-SCN001 Initial Condition MOV-SSW009F Initial Condition MOV-RHR029F Initial Condition SRV-RRS016C Initial Condition AOV-SCN014F Initial Condition OVR-RPS001A NRC Scenario 4 (Spare)

Page 3 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SCENARIO 4

SUMMARY

Event 1 With reactor power at ~3% and reactor pressure at ~300 psig, the ATC operator withdraws control rods to restore and maintain Main Turbine Bypass Valves (BPVs) approximately 20% open as directed by PPM 3.1.2 (Startup Flowchart), Attachment 7.3, step Q32.

Event 2 (TRG-2) Trip of Reactor Building Exhaust Fan 1B (REA-FN-1B) results in a high Reactor Building pressure and entry into PPM 5.3.1 (EOP - Secondary Containment Control). Secondary containment becomes inoperable. ARP 4.812.R2 9-1 (REACTOR BUILDING EXHAUST FAN B TRIP) directs starting REA-FN-1A which cannot be started (out-of-service). Subsequent ARP direction requires the BOP operator to isolate Reactor Building HVAC and starting the Standby Gas Treatment system to return Reactor Building pressure to within the TS limit ( 0.25 inch of vacuum water gauge). The CRS refers to Technical Specifications and determines that TS 3.6.4.1 (Secondary Containment), Action A.1 applies which requires restoring secondary containment to operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Event 3 (TRG-3) IRM A fails upscale resulting in an IRM upscale trip and Neutron Monitor System trip annunciators and a half scram. Per the ARP and when directed by the CRS, the ATC operator bypasses IRM A and resets the half-scram. The CRS refers to Technical Specifications 3.3.1.1 (RPS Instrumentation) and determines that the minimum number of IRM instruments required remains operable and that no TS actions are required.

Event 4 (TRG-4) Differential current lockout of transformer (TR-1/11) supplying 480V Bus SL-11 occurs which de-energizes the bus. After accessing what caused the lockout, and when directed, the BOP operator repowers SL-11 from SL-21 using the Quick Card (SOP-ELEC-480V-OPS-QC).

Event 5 Call comes into the Control Room reporting RCIC turbine coupling to the RCIC pump was found broken.

CRS will direct the RCIC turbine to be tripped. The CRS refers to Technical Specifications and determines that TS 3.5.3 (RCIC System), Action A.1 applies which immediately requires verifying that HPCS is operable by administrative means AND Action A.2 which requires restoring RCIC system to operable status within 14 days.

Event 6 (TRG-6) A break on the Residual Heat Removal Pump 2A (RHR-P-2A) suction line causes wetwell level to lower. ABN-FLOODING is entered. When attempting to close the RHR-P-2A Motor-Operated suction valve (RHR-V-4A), the valve fails open. The CRS enters PPM 5.2.1 (EOP - Primary Containment Control) on Suppression Pool low level. Crew should direct removal of control power fuses (TRG-7) for RHR-P-2A as time permits.

FDR-V-607, the cross-connect valve between the RHR-SYS-A and Reactor Core Isolation Cooling (RCIC) rooms fails to auto close due to a failed level switch (which allows flooding to continue into RCIC Pump Room). The valve cannot be manually closed. The CRS re-enters PPM 5.3.1 (EOP - Secondary Containment Control) due high RHR-SYS-A and RCIC room levels. The leak from the Suppression Pool is not considered a Primary System discharging into Secondary Containment and therefore a controlled reactor shutdown is required for high RCIC room water level (6 inches above floor).

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Event 7 The crew takes actions to restore wetwell level using the High Pressure Core Spray (HPCS) pump (HPCS-P-1) per PPM 5.5.23 (Emergency Suppression Pool Makeup). During this lineup, the HPCS Standby Service Water Pump (HPCS-P-2) discharge valve (SW-V-29) fails to auto open when HPCS-P-2 is started, requiring the BOP operator to manually open the valve. HPCS is ineffective is restoring Suppression Pool level.

Event 8 The CRS enters PPM 5.1.1 (EOP - RPV Control) and directs manually scramming the reactor prior to wetwell level reaching 19 feet 2 inches. (CT #1) The reactor will not scram when the mode switch is taken to SHUTDOWN. The ATC operator identifies the failure to scram and takes actions per PPM 3.3.1 (Reactor Scram) to scram the reactor. The Manual Scram Pushbuttons are effective in inserting all control rods.

Event 9 Prior to wetwell level going below 19 feet 2 inches, the CRS determines that wetwell level cannot be maintained 19 feet 2 inches and directs Emergency Depressurization (ED) per PPM 5.1.3 by opening seven (7) Safety Relief Valves (ADS preferred) within 10 minutes of wetwell level lowering to 19 feet 2 inches. (CT #2)

One Automatic Depressurization System (ADS) Safety Relief Valve (MS-RV-4D) fails to open during the ED requiring the BOP operator to manually open a non-ADS SRV. (CT #2)

TERMINATION CRITERIA: The scenario will be terminated when emergency depressurization has commenced (7 SRVs open) and RPV level is being controlled in the prescribed band OR as directed by the Examination Team.

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 Measurable Safety Performance Critical Task Cueing Performance Significance Feedback Indicators CT #1 - Manually Ensures reactor is Procedural direction by The operator will All control rods fully scram the reactor scrammed and PPM 5.2.1 (EOP for manually scram insert.

before wetwell level shutdown before Primary Containment reactor by placing drops below 19 feet requirement to Control) Step L-5 Reactor Mode 2 inches. Emergency directs entering PPM Switch in Depressurize (ED) is 5.1.1 (which requires Shutdown (and reached. placing Reactor Mode follow up with all Switch in Shutdown) Manual Scram If ED is anticipated once it is determined pushbuttons when (see PPM 5.1.1 P-1 that wetwell level RMS fails to scram override), dumping cannot be maintained the reactor).

steam to main above 19 feet 2 inches.

condenser via Main Turbine bypass valves may be used to reduce reactor pressure before the requirement to ED occurs. ED would still be performed if required by EOPs.

(Ref: PPM 5.0.10 Rev 21, 8.8.2 f))

CT #2 - When Suppression of Procedural direction by The operator will The valve light wetwell level cannot pressure from PPM 5.2.1 (EOP for manually open 7 indications for each be maintained blowdown Primary Containment Safety Relief of the 7 Safety above 19 feet 2 (Emergency Control) Step L-6 Valves (ADS Relief Valves will inches, initiate Depressurization) directs Emergency preferred) to change from Green emergency through the Depressurizing reactor emergency lit to Red lit when depressurization by downcomers when Wetwell water depressurize the control switch is opening seven (7) cannot be assured for level cannot be RPV. taken to Open.

Safety Relief Valves water levels below 19 maintained above 19 (ADS preferred) feet 2 inches. feet 2 inches. Reactor pressure within 10 minutes of will lower in wetwell level (Ref: PPM 5.0.10 response.

lowering to 19 feet Rev 21, 7.12.3) 2 inches.

NRC Scenario 4 (Spare)

Page 6 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SIMULATOR SETUP

Unload simulator (between each scenario)

Verify in ILC load

Ensure Startup Sequence batch file loaded

Reload simulator

Reset to IC-202 (reset, go to Run, reset again)

Load Scenario 4 Schedule file

Load Scenario 4 Event file (if not loaded automatically)

Validate that there are no unexpected annunciators or parameters out of band

Verify pump running magnets

Ensure proper Startup Sequence shows at H13-P603

Place tagout on REA-FN-1A

Have marked up copy of the following for crew (with N/Ad steps initialed) for the crew at their pre-brief location (outside the simulator):

o PPM 3.1.2 (Reactor Startup) o Rod Pull Sheets o SOP-CR-MOVEMENT o Large Startup Flowchart o ABN-ROD NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 SCHEDULE FILE

1

<ACTION>schedule Schedule\local.sch</ACTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-SCN001 to FA_CTRL_FUS</ACTION>

<DESCRIPTION>CB-REA-FN-1A REA-FN-1A MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOV-SSW009F to FAIL_AUTO_OPEN</ACTION>

<DESCRIPTION>SW-V-29 SERVICE WATER PUMP DISCHARGE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MOV-RHR029F to FAIL_AS_IS</ACTION>

</ITEM>

1

<ACTION>Insert malfunction AOV-SCN014F to FAIL_AS_IS</ACTION>

<DESCRIPTION>FDR-V-607 RCIC FLOOR DR - FDR SUMP-R1 INLT</DESCRIPTION>

</ITEM>

1

<ACTION>Insert override OVR-RPS001A to OFF</ACTION>

<DESCRIPTION>RPS-RMS-S1 REACTOR MODE SHUTDOWN</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction SRV-RRS016C to CLOSE</ACTION>

<DESCRIPTION>MS-RV-4D MS-RV-4D SAFETY RELIEF OPEN/CLOSE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction PMP-SCN010S on event 2</ACTION>

<DESCRIPTION>REA-FN-1B REACTOR BLDG EXHAUST FAN SHAFT SEIZURE</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-NIS002A to HIGH on event 3</ACTION>

<DESCRIPTION>IRM INSTR FAILURE A</DESCRIPTION>

</ITEM>

NRC Scenario 4 (Spare)

Page 8 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017

<ACTION>Insert malfunction ANN-800C3A02 to ON on event 4</ACTION>

<DESCRIPTION>XFMR TR-1/11 DIFF LOCKOUT</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-EPS001 after 10 to TRIP on event 4</ACTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-EPS004 after 10 to TRIP on event 4</ACTION>

</ITEM>

1

<ACTION>Insert malfunction MAL-RHR001 to 8400.00000 in 180 on event 6</ACTION>

<DESCRIPTION>LINE BREAK AT RHR-P-2A SUCTION</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction XMT-PCN006A after 180 to 16.00000 in 900 on event 6</ACTION>

<DESCRIPTION>CMS-LT-6A FIXED OUTPUT WETWELL WIDE RNG LEVEL DIV 1</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction XMT-PCN007A after 180 to 16.50000 in 900 on event 6</ACTION>

<DESCRIPTION>CMS-LT-6B FIXED OUTPUT WETWELL WIDE RNG LEVEL DIV 2</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction XMT-PCN003A after 120 to -25 in 240 on event 6</ACTION>

<DESCRIPTION>CMS-LT-1 FIXED OUTPUT SUPP CHAMBER LEVEL</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction XMT-PCN004A after 120 to -25 in 240 on event 6</ACTION>

<DESCRIPTION>CMS-LT-2 FIXED OUTPUT SUPP POOL LEVEL</DESCRIPTION>

</ITEM>

1

<ACTION>Insert malfunction BKR-RHR001 to FA_CTRL_FUS on event 7</ACTION>

<DESCRIPTION>CB-RHR-P-2A RHR-P-2A MOTOR SUPPLY BREAKER</DESCRIPTION>

</ITEM>

</SCHEDULE>

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 1

==

Description:==

Withdraw control rods as required to establish and maintain the bypass valves approximately 20% open Event is initiated by Turnover.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 1 Time Position Applicants Actions or Behavior CRS Directs ATC to Directs ATC to withdraw control rods to maintain the BPVs approximately 20% open (per startup flowchart step Q32)

Comment:

ATC References the Startup Rod Withdrawal sequence sheets to identify next rod to withdrawal Comment:

BOP May peer check ATC during manipulations Comment:

ATC Withdraws rod as follows:

Selects control rod

Uses continuous withdraw up to 2 notches prior to final position (unless withdrawing to notch 48)

Single notch withdraws the final 2 notches Comment:

BOP Periodically reports Bypass Valve position to ATC Comment:

ATC Reports to CRS when Bypass Valves are approximately 20% open Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2

==

Description:==

Trip of REA-FN-1B results in a high reactor building pressure and entry into PPM 5.3.1 (EOP - Secondary Containment Control) (Tech Spec)

Event is initiated after rod withdrawal has opened Bypass Valves approximately 20% open (or as directed by the Exam team) and is activated using TRIGGER 2.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 2 Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 812.R2 9-1 (RX BLDG EXH FAN B TRIP)

Comment:

Determines that fan REA-FN-1B has tripped Comment:

Reports annunciator and status of fan REA-FN-1B to the CRS Comment:

Examiner Note: Following steps are from ARP 812.R2 9-1 (RX BLDG EXH FAN B TRIP)

BOP 1: Verifies REA-FN-1B tripped (may have been previously completed)

Comment:

2: Notes that REA-FN-1A cannot be started (undergoing maintenance) and informs CRS that a Standby Gas Treatment train will have to be started Comment:

CRS 3.a: Directs BOP to start either Standby Gas Treatment train 1A or 1B per the SOP-SGT-START-DIV1(2)-QC (Standby Gas Treatment Start - Quick Card)

Comment:

Calls for assistance in getting REA-FN-1A or REA-FN-1B back Comment:

BOP/ATC Acknowledges annunciator 602.A5 2-8 (SEC PRESS DP HIGH) when it comes in and informs CRS Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2 (CONTINUED)

Examiner Note: Following steps are from ARP 602.A5 2-8 (SEC PRESS DP HIGH)

BOP/ATC 1 & 2: Checks REA-DPR-1A(B) for RB Pressure (already known to be near zero) and refer to CRS to ppm 3.8.1 (Secondary Containment Control)

Comment:

3: Refers CRS to Technical Specification 6.3.4.1 Comment:

Refers to annunciator 812.R1 7-3 (SEC PRESS CONTR A P HIGH/LOW) and 812.R2 7-1 (SEC PRESS CONTR B P HIGH/LOW) and notes they are expected for the plant condition (no RB HVAC)

Comment:

CRS Enters ppm 3.8.1 (Secondary Containment Control) on low RB differential pressure Comment:

Evaluates Technical Specifications and determines the following action applies:

LCO 6.3.4.1 A.1 - Restore secondary containment to operable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Comment:

Examiner Note: Following steps are from SOP-SGT-START-DIV1-QC (Standby Gas Treatment Start - Quick Card) assuming Div 1 is started (Div 2 is similar)

BOP 3.1.1: (2 handed operation) Places the following fans to PTL (Pull to Lock)

ROA-FN-1A

ROA-FN-1B

REA-FN-1A

REA-FN-1B Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 2 (CONTINUED)

BOP 3.1.2: Closes the following valves:

ROA-V-1

ROA-V-2

REA-V-1

REA-V-2 Comment:

3.1.3: Momentarily turns SGT-FN-1A1 fan control switch from Auto to PTL SYS.START Comment:

3.1.4: Verifies the following items:

Main Heaters energize as indicated by the Main Heater ON light and A1 amp meters

SGT-V-5A1 opens (Exhaust to Stack)

SGT-FN-1A1 starts (within 10 seconds)

Comment:

Examiner Note: Following are steps from ARP 812.R2 9-1 (RX BLDG EXH FAN B TRIP)

BOP 3.b: Notifies Chemistry to monitor Reactor Building ventilation per ODCM 6.1.2.1 and LCS 1.3.3.1 3.c: Refers CRS to ODCM 6.1.2.1 and LCS 1.3.3.1 Comment:

4: Refers CRS to ABN-HVAC (no actionable items)

Comment:

BOP Monitors secondary containment D/P with a Standby Gas Treatment train running and informs CRS when secondary containment D/P has been restored Comment:

CRS Validates restoration of secondary containment integrity - exits LCO 3.6.4.1 Evaluates exiting of PPM 5.3.1 (Secondary Containment Control)

Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3

==

Description:==

IRM A fails upscale resulting in a half scram Event is initiated after secondary containment D/P has been restored with a Standby Gas Treatment train running (or as directed by the Exam team) and is activated using TRIGGER 3.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 3 Time Position Applicants Actions or Behavior ATC Acknowledges annunciators 603.A7 1-5 (IRM ACEG UPSCL TRIP OR INOP) and 603.A7 3-4 (1/2 SCRAM SYSTEM A)

Comment:

Checks for control rod motion Comment:

Reports to CRS that a half scram occurred (RPS A white RPS scram lights de-energized) due to IRM upscale and that no rod motion has occurred Comment:

BOP Makes PA announcement Half Scram system A. Stop all maintenance and surveillance testing on RPS system B.

Comment:

CRS Calls Work Week Manager or Operations Management for assistance Comment:

Examiner Note: Following are steps from ARP 603.A7 3-4 (1/2 SCRAM SYSTEM A)

ATC 2.a: Checks for scrammed rods (may have already been performed)

Comment:

BOP 2.c: Make announcement to stop work (may have already been made)

Comment:

CRS 2.d: Directs Bypassing IRM A Reports when IRM Upscale or INOP annunciator clears Comment:

ATC Positions IRM Bypass Switch to bypass IRM A Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 3 (CONTINUED)

CRS Directs ATC to reset the half scram Comment:

ATC 3: Resets half scram by doing the following:

3.a: Depresses RPS-RMS-S5A (RPS Logic A1/B1 Reset pushbutton ) (H13-P603).

3.b: Depresses RPS-RMS-S5B (RPS Logic A2/B2 Reset Pushbutton) (H13-P603)

3.c: Verifies the Scram group solenoid lights for Groups 1, 2, 3 and 4 are illuminated (H13-P609 & H13-P603)

3.d: Verifies the Backup Scram System lights have extinguished (H13-P603)

Comment:

4: Refers the CRS to Technical Specification 3.3.1.1 and LCS 1.3.2.1 Comment:

CRS Evaluates Technical Specification 3.3.1.1 and LCS 1.3.2.1 and determines that the minimum number of channels required for IRM operability exists and that no T.S. action statements need be entered Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4

==

Description:==

Differential current lockout of transformer (TR-1/11) results in a loss of SL-11 which requires bus to be re-energized from alternate source Event is initiated after IRM A has been bypassed and a Tech Spec evaluation has been made (or as directed by the Exam team) and is activated using TRIGGER 4.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 4 BOP Acknowledges annunciator 800.C3 1-2 (XFMR TR-1/11 DIFF LOCKOUT) and notes 480 VAC Bus SL-11 has de-energized - Reports annunciator and bus status to CRS CRS Enters ABN-ELEC-SM1/SM7 Comment:

Comment:

Examiner Note: Following are steps from ARP 800.C3 1-2 (XFMR TR-1/11 DIFF LOCKOUT)

BOP 1: Verifies that both CB-1/11 & CB 11/1 feeder breakers tripped open (as expected based on alarm) but that CB 21/11 did not close Comment:

CRS 3: Requests plant assistance for cause of transformer lockout Comment:

Examiner Note: Following are steps from ABN-ELEC-SM1/SM7 (section 4.7)

BOP 4.7.1: Verifies DEH-P-1B is running Comment:

4.7.3: Verifies SL-11 lockout is reset Comment:

CRS 4.7.4: When the E-SL-11 problems have been corrected, then restore SL-11 to service per SOP-ELEC-480V-OPS-QC (there are not issues with SL-11 itself)

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 4 (CONTINUED)

CRS 4.7.5: When SL-11 is returned to service then restore SL-11 loads to service per SOP-ELEC-SM1-MAINT (the CRS will make note of this)

Comment:

BOP 4: Determines that normal source to Bus SL-11 unavailable but energizes SL-11 using SOP-ELEC-480V-OPS-QC quick card Comment:

Examiner Note: Following are steps from SOP-ELEC-480V-OPS-QC quick card (section 2.4)

BOP 2.4.1: Verifies CB-21/2 closed Comment:

2.4.2: Verifies CB-11/1 green light illuminated and green flag displayed Comment:

2.4.3: Verifies CB-21/11 green light illuminated and green flag displayed Comment:

2.4.4: Closes CB-21/11 Comment:

2.4.5: Verifies SL-11 voltage is approximately 480 (432-528) volts Comment:

2.4.6: Verifies (and maintains) E-TR-1/11 load 277 amps Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5

==

Description:==

RCIC-P-1 coupling discovered broken (Tech Spec)

Event is initiated after Bus SL-11 has been restored (or as directed by the Exam team) and commences with a call to the Main Control Room.

BOOTH OPERATOR - Call the Main Control Room and report the following as OPS-2:

I discovered several pieces of RCIC pump coupling on the floor in the RCIC Pump room. The pump looks detached from the turbine.

CRS Directs ATC to trip RCIC turbine Comment:

Examiner Note: For ATC credit the CRS will have to be directed to ensure ATC gets assigned task while BOP monitors reactor parameters.

ATC Trips RCIC turbine as directed by pressing RCIC Manual trip pushbutton Comment:

Refers to ARP 601.A4 1-5 (RCIC TURBINE TRIP) for follow up actions Comment:

Examiner Note: Following are steps from ARP 601.A4 1-5 (RCIC TURBINE TRIP)

ATC 1: Verifies RCIC-V-1 is closed (RCIC Turbine Trip and Throttle Valve)

(H13-P601)

Comment:

2: Verifies RCIC-V-46 is closed Comment:

5: Refers CRS to Technical Specification 3.5.3 Comment:

CRS 6: Informs Security to take compensatory actions for RCIC out of service Comment:

Enters RCIC as inoperable in the Plant Logging system - Evaluates Technical Specifications and determines the following actions apply:

LCO 3.5.1 Action A.1 applies which immediately requires verifying that HPCS is operable by administrative means (it is)

Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 5 (CONTINUED)

CRS LCO 3.5.1 Action A.2 which requires restoring RCIC system to operable status within 14 days Comment:

Request assistance on RCIC investigation and unplanned unavailability Comment:

Protects HPCS-P-1, HPCS DG and HPCS Service Water systems Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6

==

Description:==

Failure of the RHR-P-2A suction line results in lowering wetwell level (unisolable)

Event is initiated after RCIC system is tripped and Technical Specification call is made (or as directed by the Exam team) and is activated using TRIGGER 6.

BOOTH OPERATOR - As briefed or when directed activate TRIGGER 6 Time Position Applicants Actions or Behavior BOP Acknowledges annunciator 602.A13 2-1 (REACTOR BLDG FLOOR SUMP R1 LEVEL HI-HI) and informs CRS Comment:

Examiner Note: Following are steps from ARP 602.A13 2-1 (REACTOR BLDG FLOOR SUMP R1 LEVEL HI-HI)

BOP 1: Determines Sump Pump status by calling Radwaste Control Room to ensure that either FDR-P-1A or 1B is running Comment:

BOOTH ROLEPLAY - If directed to report status of FDR-P-1A and 1B sump pumps, report both floor drain sump pumps are running.

BOP 2: Sends field operator to investigate RI Sump level (and possible flooding) in RHR A pump room Ensures they understand that this is a potentially hazardous situation and that they need to take the appropriate precaution NOTE: This step is also directed from ABN-FLOODING step 4.1.1 Comment:

BOOTH ROLEPLAY - If directed to investigate possible flooding in RHR A pump room, wait 1 minute and:

If alarm 601.A4 5-3 has not come in report I hear a big inrush of water in the RHR A pump room. The Sump is overflowing with several inches of water on the floor.

If alarm 601.A4 5-3 has already come in report I hear a big inrush of water in the RHR A pump room with about a foot of water of water on the floor and rising. Im leaving do to safety concerns.

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6 (CONTINUED)

BOP 3: Notes that FDR-V-607 (RCIC Floor Drain Sump FDR Sump R1 Inlet) did not automatically close and attempts to manually close it on H13-P632 Reports to CRS that FDR-V607 did not auto close and could not be closed manually Comment:

Acknowledges annunciator 601.A4 5-3 (RHR A PUMP ROOM WATER LEVEL HIGH) and informs CRS Refers CRS to ppm 5.3.1 (Secondary Containment Control) (per ARP)

Comment:

Acknowledges annunciator 601.A12 2-3 or 601.A11 2-3 (SUPP POOL LEVEL HIGH/LOW)

Observes lowering level in the Suppression Pool (Wetwell) and provides crew update Comment:

CRS Enters ABN-FLOODING Comment:

Enters 5.3.1 (Secondary Containment Control) on RB water level above alarm setpoint of 6 inches (RHR Pump Room A)

Comment:

Enters ppm 5.2.1 (Primary Containment Control) based on low suppression level (-2 inches)

Comment:

Directs BOP to verify RHR-P-2A secured and then to shut RHR-V-4A (Pump Suction from Supp Pool) in an attempt to isolate the leak (as directed by ppm 5.3.1 (Secondary Containment Control) step SC-9)

Comment:

BOP Reports after isolation attempt that RHR-V-4A did not close and that Suppression Pool level continues to lower Comment:

NRC Scenario 4 (Spare)

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Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 6 (CONTINUED)

CRS Directs BOP to make announcement per ABN-FLOODING step 4.2 Comment:

BOP Since flooding was confirmed by report/plant indications:

4.2.1: Sounds Alert Tone for 5 to 10 seconds

4.2.2: Alert station personnel to flooding in the affected room(s)

4.2.3: Evacuate all unnecessary personnel (may already be done)

4.2.4: Refer to PPM 13.5.1 for localized evacuation Comment:

CRS Directs BOP to emergency makeup to the Suppression Pool per 5.5.23 (Emergency Suppression Pool Makeup) (See Event 7 for actions)

Comment:

Sets a Key Plant Parameter for Suppression Pool (Wetwell) level sufficiently above 19 feet 2 inches (to allow margin for actions needed to ED later on before reaching 19 feet 2 inches)

Comment:

ATC/BOP 4.1.2: Trends Key Plant Parameter for Suppression Pool (Wetwell) level Comment:

ATC Directs field operator to remove trip and close (control power) fuses for RHR-P-2A (per ABN-Flooding Attachment 7.1 (section 7.1.1)

Comment:

BOOTH ROLEPLAY - If directed to pull the control power fuses for RHR-P-2A, wait 3 minutes then activate TRIGGER 7 then report The trip and close fuses have been removed for RHR-P-2A.

NRC Scenario 4 (Spare)

Page 22 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 7

==

Description:==

SW-V-29 fails to auto open when HPCS-P-2 is started for wetwell makeup Event is activated at the beginning of the scenario and is realized when SW-V-29 fails to auto open.

Time Position Applicants Actions or Behavior BOP When directed to perform ppm 5.5.23 (Emergency Suppression Pool Makeup) the following actions are taken (per section 4)

4.1: Verifies HPCS-V-1 is open (Pump Suction from CST)

4.2: Starts HPCS-P-1

4.3: Verifies HPCS-V-12 opens (HPCS-P-1 Minimum Flow Bypass)

4.4: Starts HPCS-P-2 (HPCS Service Water Pump)

4.5: Verifies SW-V-29 opens (Service Water Pump Discharge)

Observes that SW-V-29 did NOT automatically open and therefore attempts to open it manually Reports to the CRS that SW-V-29 did not automatically open but was able to be opened manually

4.6 is N/A (no HPCS auto initiation signal present)

4.7: Throttles open HPCS-V-23 (Test Bypass To Suppression Pool)

4.8: Adjusts flow as necessary to a maximum of 7175 GPM to fill the Suppression Pool

4.9: Verifies HPCS-V-12 closes

4.10: Monitors Suppression Pool level Reports to CRS that HPCS is making up to Suppression Pool but Suppression Pool level continues to lower Comment:

NRC Scenario 4 (Spare)

Page 23 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 8

==

Description:==

Reactor mode switch fails to scram reactor, requiring use of manual scram pushbuttons to scram reactor prior to wetwell level lowering to 19 feet 2 inches Event is activated at the beginning of the scenario and is realized when the Mode Switch is positioned to Shutdown and no scram occurs.

Time Position Applicants Actions or Behavior CT #1 - Manually scram the reactor before wetwell level drops below 19 feet 2 inches.

ATC/BOP Reports when Key Plant Parameter met for Suppression Pool (Wetwell) low level Comment:

CRS When notified Key Plant Parameter has been reached for Wetwell level, updates the crew on plant conditions then enters PPM 5.1.1 (RPV Control)

Comment:

Directs ATC to scram the reactor Comment:

Examiner Note: Following steps are Immediate Actions from PPM 3.3.1 (Reactor Scram)

Examiner Note: The reactor scram procedure (PPM 3.3.1) does not provide direct procedural guidance for the situation where the Mode Switch is taken to Shutdown at < 5% power (APRM downscale) and the rods fail to insert. The ATC may automatically use the Manual Scram pushbuttons (as a backup to the Mode switch) or the CRS may direct use of the Manual Scram pushbuttons as authorized by PPM 1.3.1 (Operating Policies, Programs, and Expectations). As an alternative, when PPM 5.2.1 (Primary Containment Control) directs entry into PPM 5.1.1 (RPV Control - Step L-5), the CRS may consider the failure to scram with the Mode Switch justification for entering PPM 5.1.2 (RPV Control ATWS) via the RC-2 override in PPM 5.3.1.

After Inhibiting ADS and taking manual control of HPCS , the Reactor Power leg in PPM 5.2.1 would then direct use of ARI (via pushbuttons) to shut down the reactor.

ATC 6.1.1: Places Reactor Mode Switch to Shutdown

Observes that the control rods DID NOT go in Comment:

6.1.2: Monitors reactor power, pressure and level (no change)

Comment:

6.1.5: Inserts SRM and IRM monitors (detectors) (some are not fully inserted during this point in the startup)

Comment:

NRC Scenario 4 (Spare)

Page 24 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 8 (CONTINUED)

ATC After above three steps ATC makes scram report to CRS:

Mode switch is in Shutdown

APRMs are downscale

RPV pressure is (value and trend)

RPV level is (value and trend)

No EOP entry (reactor power is < 5%)

Comment:

CRS Repeats back scram report Comment:

ATC 6.1.6: After CRS repeat back, reports all control rods are NOT IN

If ATC automatically presses the Manual Scram pushbuttons before reporting rod status, all rods will be IN Comment:

CRS May direct ATC to use Manual Scram pushbuttons (if not used already) to insert all control rods or may direct after entering PPM 5.1.2 (RPV Control -

ATWS) initiation of ARI (via switches on H13-P603) (see examiner note previous page)

Comment:

Examiner Note: Following steps are Subsequent Actions from PPM 3.3.1 (Reactor Scram)

ATC 6.2.6: Range down on IRMs, as necessary, to follow power decrease Comment:

BOP 6.2.7: Make PA announcement for reactor scram Comment:

ATC 6.2.8: Transfers level control to RFW-FCV-10A/B per SOP-RFW-FCV-QC quick card (No action - already on startup level controller in Auto)

Comment:

CRS Sets a Key Plant Parameter for Suppression Pool (Wetwell) level sufficiently above 19 feet 2 inches to allow a controlled Emergency Depressurization Comment:

NRC Scenario 4 (Spare)

Page 25 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 EVENT No. 9

==

Description:==

Prior to wetwell level going below 19 feet 2 inches, the crew determines that wetwell level cannot be maintained 19 feet 2 inches and initiates RPV Emergency Depressurization (ED) with 7 SRVs opened (SRV MS-RV-4D) fails to open requiring manually opening one non-ADS SRV)

Time Position Applicants Actions or Behavior CT #2 - When wetwell level cannot be maintained above 19 feet 2 inches, initiate emergency depressurization by opening seven (7) Safety Relief Valves (ADS preferred) within 10 minutes of wetwell level lowering to 19 feet 2 inches.

CRS When notified Key Plant Parameter has been reached for Wetwell level, updates the crew on plant conditions, exits the pressure leg of PPM 5.1.1 (RPV Control) via override then enters PPM 5.3.1 (Emerg Depressurization)

Comment:

Determines a high Drywell pressure signal is not sealed in Comment:

Determines Wetwell level is > 17 feet Comment:

Directs 7 SRVs be opened (ADS preferred) (ADS SRVs are those with the red stripe on left side of their nameplate)

Comment:

Examiner Note: Proper containment response (comparing Wetwell and Drywell pressures as each SRV is opened to detect tailpipe failure) will be difficult at an already low RPV pressure.

BOP Opens 7 SRVs (ADS preferred) as directed while verifying proper containment response as each is opened and reports completion to CRS

Observes that SRV (MS-RV-4D) did not open

Opens one other non-ADS SRV Reports 7 SRV opened and that SRV 4D failed to open requiring the opening of another SRV Comment:

CRS Directs RPV level band of +13 to +54 inches Comment:

ATC Maintains RPV level as required to maintain RPV level band Comment:

NRC Scenario 4 (Spare)

Page 26 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 TERMINATION CRITERIA: The scenario will be terminated when emergency depressurization has commenced (7 SRVs open) and RPV level is being controlled in the prescribed band OR as directed by the Examination Team.

NRC Scenario 4 (Spare)

Page 27 of 28

Appendix D NRC Scenario No. 4 FORM ES-D-2 Columbia Generating Station ILC NRC Exam - February, 2017 TURNOVER Initial Conditions:

The reactor is in Mode 2 during reactor startup

Reactor is critical at 3% power with RPV pressure at approximately 300 psig

Reactor Building Exhaust Fan 1A (REA-FN-1A) is out of service for extended maintenance Shift Turnover:

Withdraw control rods as required to establish and maintain Bypass Valves approximately 20% open

Next control rod movement is Rod 16-43 (Group 30-13) going from notch 12 to 48

Continuous rod withdrawal permitted NRC Scenario 4 (Spare)

Page 28 of 28

ML17081A541

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