Text

Retake Admin JPMs
	ML22146A090
Person / Time
Site:	Perry
Issue date:	07/27/2017
From:	Randy Baker; NRC/RGN-III/DRS/OLB
To:	;
	Baker R
Shared Package
ML17142A146	List: ML22146A090; ML22146A095; ML22146A098; ML22146A100;
References
	Download: ML22146A090 (91)
	v • d • e

OT-3701-ADM_025SRO Rev 0 JOB PERFORMANCE MEASURE SETUP SHEET System: Administrative Time Critical: No Applicability: SRO only Administrative Topic: N/A - Refueling Validated: 28 minutes Setting: Classroom / Simulator

References:

FTI-D09 Rev. 19 & SOI-F15 Rev 22 Required Material: FTI-D09 - Use of the Fuel Movement Checklist & JPM Support Material Tasks: 341-029-03-02 Supervise Refueling Operations as Refueling Supervisor 341-624-01-02 Verify Compliance with Technical Specifications, Procedures, and Instructions During Refueling and Fuel Handling Task Standard: Verify proper placement of in-core components during Refuel Operations IAW the Fuel Movement Checklist.

K/A Data: 2.1.35 Knowledge of the fuel-handling responsibilities of SROs.

Importance Rating: RO 2.2 SRO 3.9.

2.1.42 Knowledge of new and spent fuel movement procedures.

Importance Rating: RO 2.5 SRO 3.4

1. Instructions: None

2. Location / Method: Simulator or Control Room / Administrative performance.

3. Initial Condition: Plant is in Mode 5 with refueling operations in progress. The ROV cameras had been out of service for the past several hours, but are now back in service.

You are the on-coming Refuel Bridge SRO. The off-going Refuel Bridge SRO asks you to verify Page 65 of the attached Fuel Movement Checklist. The Fuel Movement Checklist Cover Sheets, Core Map, and ROV pictures of the applicable core cells are available.

4. Initiating Cue: As the Refueling SRO, verify the in-core placement of all components on Page 65 of the attached Fuel Movement Checklist.

Start Time End Time Operator OT-3701-ADM_025SRO Rev 0 JPM BODY SHEET Standard: Performer obtains or simulates obtaining all materials, procedures, tools, keys, radios, etc before performing task.

Standard: Performer follows management expectations with regards to safety and communication standards.

Step 1 Determine INCORRECT loading of Cell 54-15.

The operator reviews the Fuel Movement Checklist and compares the cell pictures to the FMC.

Critical Step: The Operator reviews the FMC Page 65 Step 295 and determines that bundle 12P400 is in the incorrect location. Additionally, bundle 12P499 for Step 296 is in the incorrect location.

Instructor Cue: Acknowledge errors and advise to continue to perform peer check.

Notes: Provide a copy of FTI-D09.

Bundle 12P400 should be in location 53-16 and bundle 12P499 should be in location 55-14.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_025SRO Rev 0 Step 2 Determine CORRECT loading of Cell 46-35.

Standard: The Operator reviews the FMC Page 65 Steps 306, 306.5, & 307 and determines that they were performed correctly.

Instructor Cue: Advise to continue to perform peer check.

Notes: None SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 3 Determine INCORRECT loading of Cell 42-19.

Critical Step: The Operator reviews the FMC Page 65 Step 311 and determines that bundle 13P907 is in the correct location and correct orientation. However, bundle 13P911 for Step 312 has the incorrect orientation in the correct location.

Instructor Cue: If Operator has not referenced FTI-D009, question operator on what is required when an error is discovered.

Acknowledge errors and advise to continue to perform peer check.

Notes: None SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_025SRO Rev 0 Step 4 Determine INCORRECT orientation (loading) of Cell 38-55.

Standard: The Operator reviews the FMC Page 65 Step 315 and determines that the Full Blade Guide was placed in the incorrect locations.

Instructor Cue: Ask Operator if this is a fuel movement error. He should respond no.

Notes: FBG is in locations 39-56 & 37-54.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 5 Question to the Candidate: What is the required action if a fuel movement error is discovered?

FTI-D09 - Use of the Fuel Movement Checklist 4.7 Recovery from Fuel Movement Errors

A fuel movement error is any condition where a fuel assembly is not as defined in the Fuel Movement Checklist.

Perform the following steps to recover from a fuel movement error:

1. Notify the Unit Supervisor immediately following the discovery of any error.

NOTE Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position. A mispositioned bundle within the reactor does not constitute a safe conservative position until the impact on shutdown margin is determined.

3. Suspend core alterations for any mispositioning error discovered in a reactor location (RX1-).

JPM Step 5 continued next page.

OT-3701-ADM_025SRO Rev 0 JPM Step 5 continued:

4. Relocate all single mispositioning and misorientation errors in the reactor to a vacant non-reactor location. The Control Room Unit Supervisor shall be apprised of an intended non-reactor location. If multiple errors are discovered, contact the Reactor Engineer to evaluate the impact of the errors and to recommend the appropriate recovery actions.

5. Contact the Reactor Engineer to evaluate the impact of the fuel movement error and to recommend the appropriate recovery actions. The effect on Shutdown Margin shall be addressed, as required.

Standard: The Operator reviews FTI-D009 and determines the required action if an error is discovered.

Instructor Cue: Ask the Candidate, What is the required action if an error is discovered?

Notes: Actions for recovery form a fuel movement error are in FTI-D09.

Terminate the JPM SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Terminating Cue: Operator determines that bundles 12P400 and 12P499 are in the incorrect locations and bundle 13P911 has the incorrect orientation in the correct location.

Evaluation Results: SAT_____ UNSAT_____

End Time OT-3701-ADM_025SRO Rev 0 JPM CUE SHEET INITIAL CONDITIONS:

Plant is in Mode 5 with refueling operations in progress.

The ROV cameras had been out of service for the past several hours, but are now back in service.

You are the on-coming Refuel Bridge SRO.

The off-going Refuel Bridge SRO asks you to verify Page 65 of the attached Fuel Movement Checklist.

The Fuel Movement Checklist Cover Sheets, Core Map, and ROV pictures of the applicable core cells are available.

INITIATING CUE: As the Refueling SRO, verify the in-core placement of all components on Page 65 of the attached Fuel Movement Checklist.

OT-3701-ADM_025SRO Rev 0 FUEL MOVEMENT CHECKLIST REFUEL FLOOR CHECKLIST PAGES 1 TO 182 STEPS 2 TO 1860 14-MAR-2017 15:01 The purpose of this FMC is to support refueling activities in 1R16. It shall be executed in accordance with FTI-D0009, SOI-F0015, SOI-F0042, and 101-0009.

The Maintenance Window for control rod blades and drive mechanisms, and SRM and IRM visual inspections is reached when Step 495 is completed and stays open through Step 1071.

Steps 1072 through 1074 (which move only single blade guides) may be completed any time after the SRM and IRM dry tube inspections have been completed but must be completed before the LPRM dry tube replacements begin.

Phase 2 commences with Step 1075 and ends with Step 1845.

Steps 1854 through 1860 configure the half blade guides in RPI. Since these half blade guides are not used during the outage, these steps may be completed at any convenient time.

In Phase 1, if any SRM becomes inoperable, contact Reactor Engineering to determine if fuel moves may be continued in another quadrant, and any limitations to changing quadrants. No changes are permitted in Phase 2.

Phase 2 peripheral camera inspections for uncontrolled bundle seating are noted on the FMC. Seating checks are performed after step has been completed.

The quadrants for I RI 3 are defined by the North/South line along coordinate 30 and the East/West line along coordinate 31. This makes control rod 30-3 1 the center of the core. This is a change from 1 RI 2.

If a planned location in the containment pool (RP1) is not usable, any unused location in RP1 rows C through F, columns -01 through -03, may be used for a FREE MOVE. Note that the furthest East column is designated 00 (not 01).

Page 1 of 2 OT-3701-ADM_025SRO Rev 0 FUEL MOVEMENT CHECKLIST REFUEL FLOOR CHECKLIST PAGES 1 TO 182 STEPS 2 TO 1860 14-MAR-2017 15:01 Notify Reactor Engineering any time a Free Move is used and cleared.

Orientation in RP1 is in accordance with the direction of the bridge supervisor.

Transfer Tube steps may be performed any time the appropriate component is loaded in the carriage.

Fuel movement does not involve invessel fuel moves.

Fuel movement only removes fuel from core.

Fuel movement only reloads core to BOC pattern.

Fuel movement involves a shuffle through intermediate loading patterns.

Minimum SDM = _1.0__% dk/k Reference Tech Spec SR 3.1.1.1 concerning the above requirements.

PREPARED BY: Chuck R. Enginer REVIEWED BY: Pat B. Peer APPROVED BY: P. J. Supe DATE: 3/19/17 FMC COMPLETE: ______________________________ DATE: __________

CHANGE HISTORY:

Page 2 of 2 OT-3701-ADM_025SRO Rev 0 FENOC - - PNPP NUCLEAR FUEL ACCOUNTING FUEL MOVEMENT CHECKLIST - Refuel Floor FMC 14-Mar-2017 15:01 Page 65 FROM FREE MOVE TO STEP COMPONENT LOCATION INIT/DATE LOCATION ORIENT INIT DATE 295 12P400 RX1 31-30 RX1 53-16 SE HD LEFT - EZ RIGHT / LBJ / 4/4/17 296 12P499 RX1 05-14 RX1 55-14 NW EZ LEFT - EZ RIGHT / LBJ / 4/4/17 306 14P046 FTT U1-E RX1 45-36 SE

/ LBJ / 4/4/17 306.5 TRANSFER FUEL CARRIAGE DOWN TO FHB LBJ / 4/4/17 307 14P038 RP1 B-01 RX1 47-34 NW

/ LBJ / 4/4/17 311 13P907 RX1 43-30 RX1 43-20 SW HD LEFT - EZ RIGHT / LBJ / 4/4/17 312 13P911 RX1 49-30 RX1 41-18 NE EZ LEFT - EZ LEFT / LBJ / 4/4/17 315 FBG RX1 41-20 RX1 37-56 RX1 43-18 RX1 39-54 LBJ / 4/4/17 316 14P051 RX1 59-24 FTT U1-W

/ LBJ / 4/4/17 317 14P049 FTT U1-E RX1 47-48 SE

/ /

OT-3701-ADM_025SRO Rev 0 N

Control Cell 54-15

N Control Cell 46-35

N Control Cell 42-19

N Control Cell 38-55

OT-3701-ADM_031_SRO Rev 0 JOB PERFORMANCE MEASURE SETUP SHEET System: Admin Time Critical: No Alternate Path: No Applicability: SRO only Safety Function: Conduct of Operations Validated Time: 30 Minutes

References:

NOP-OP-1002 Rev 12 Required Material NOP-OP-1002- Conduct of Operations 10CFR50.54 Conditions of licenses Task: 343-598-03-02 Contact an Individual in Situations that Require Being Called-in During Non-Scheduled Work Hours 343-501-03-03 Verify that the Shift is Manned Properly Task Standard: Identify positions that are not staffed properly for the on-coming shift.

K/A: 2.1.5 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. Importance: SRO 3.9.

1. Simulator Setup Instructions: N/A

2. Location / Method: Classroom / Performance

3. Initial Condition: Plant is in Cold Shutdown for a forced outage. A severe snow storm has made travel to the plant difficult. Attached is a list of the oncoming crew personnel that are currently on site. The report was developed from Career Map. The Security Shift Supervisor reports that Site Protection is fully staffed.

4. Initiating Cue: As the off going Shift Manager, determine if the oncoming crew has sufficient personnel to meet minimum FENOC shift staffing requirements. Document the results of your determination on the JPM CUE SHEET.

Start Time End Time Operator 1

OT-3701-ADM_031_SRO Rev 0 JPM BODY SHEET Standard: Performer obtains or simulates obtaining all materials, procedures, tools, keys, radios, etc before performing task.

Standard: Performer follows management expectations with regards to safety and communication standards.

Step 1 NOP-OP-1002 Conduct of Operations 4.1.12 Fire Brigade

1. Site specific staffing requirements are contained in Attachment 4.

4.1.13 Shift Staffing

1. A summary of Operations staffing requirements at each site is contained in Attachment 4.

With information provided, candidate reviews NOP-OP-1002 and/or 10CFR50.54 to determine if minimum FENOC staffing requirements are met.

Standard: Operator reviews NOP-OP-1002 and the Career Map printout to determine if sufficient personnel on site to man the shift.

Instructor Cue: If requested, provide copies of NOP-OP-1002 and/or 10CFR50.54.

If asked about the Column Names, Complete means Qualified.

Notes: Perry Shift Staffing Requirements are contained on Attachment 4 of NOP-OP-1002, page 101.

Minimum Safe Shutdown staffing in Mode 4 is 3 people. However, Minimum EP staffing is 18 people (14 from Operations).

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

2

OT-3701-ADM_031_SRO Rev 0 Step 2 With information provided, determine if minimum FENOC staffing requirements are met for Shift Manager, Unit Supervisor and Shift Engineer.

Critical Step: Candidate determines that minimum EP staffing requirements for Shift Manager, Unit Supervisor and Shift Engineer cannot all be met. One of the three will not be filled.

Instructor Cue: After Candidate determines that all positions cannot be filled, ask for explanation of which spots cant be filled.

Notes: Safe Shutdown requirements can be met, but not EP Staffing requirements.

Only Tony E. can fill the Shift Manager spot.

Only Adam C. can fill the Shift Engineer spot.

Either Tony E or Adam C can fill the US spot.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 3 With information provided, determine if minimum FENOC staffing requirements are met for Reactor Operator and Fire Brigade Leader.

Critical Step: Candidate determines that minimum staffing requirements for Reactor Operator and Fire Brigade Leader cannot be met. Either one RO or the Fire Brigade Leader cannot be filled.

Instructor Cue: None Notes: Safe Shutdown requirements can be met, but not EP Staffing requirements.

ROs or SROs can be qualified Fire Brigade Leaders.

Only Henry J. and Jacob P. can fill the 2 RO spots.

Only Henry J. can fill the FBL spot SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

3

OT-3701-ADM_031_SRO Rev 0 Step 4 With information provided, determine if minimum FENOC staffing requirements are met for Non-Licensed Operator and Fire Brigade Member.

Critical Step: Candidate determines that minimum staffing requirements for Non-Licensed Operator and Fire Brigade Member cannot be met. Either one NLO or one Fire Brigade member cannot be filled.

Instructor Cue: None Notes: Safe Shutdown requirements can be met, but not EP Staffing requirements.

POs or PPAs can be qualified Fire Brigade Members.

Only Amanda L., Andrew J., Brian J., and David P. can fill the 2 NLO spots.

Only Amanda L., Andrew J., Colin J. and David P. can fill the 4 Fire Brigade Member spots.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 5 With information provided, determine if minimum FENOC staffing requirements are met for Radwaste Technician.

Standard: Candidate determines that minimum staffing requirements for Radwaste Technician are met.

Instructor Cue: None Notes: Radwaste Operations Supervisor is the Radwaste Technician.

Radwaste Technician can be filled by either Alan K. or Robert J.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

4

OT-3701-ADM_031_SRO Rev 0 Step 6 With information provided, determine if minimum FENOC staffing requirements are met for Health Physics Technician.

Standard: Candidate determines that minimum staffing requirements for Health Physics Technician are met.

Instructor Cue: None Notes: The RP Shift Tech is the Health Physics Technician Health Physics Technician positions (2) are filled by both Alan D. and Nick A.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 7 With information provided, determine if minimum FENOC staffing requirements are met for Chemistry Technician.

Standard: Candidate determines that minimum staffing requirements for Chemistry Technician are met.

Instructor Cue: None Notes: Chemistry Technician can only be filled by Elizabeth A.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

5

OT-3701-ADM_031_SRO Rev 0 Step 8 With information provided, determine if minimum FENOC staffing requirements are met for I&C Technician.

Standard: Candidate determines that minimum staffing requirements for I&C Technician are met.

Instructor Cue: None Notes: I&C Technician can be filled by either Bret B., David R., or Mason M.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 9 With information provided, determine if minimum FENOC staffing requirements are met for NRC Phone Talker.

Standard: Candidate determines that minimum staffing requirements for NRC Phone Talker are met.

Instructor Cue: None Notes: Even though Michael C is not qualified as Rx operator, he is licensed and has requisite knowledge to fill this position. NRC Phone Talker can only be filled by Michael C. unless an additional RO, NLO or FBM position is left open.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Terminating Cue: Three positions in the oncoming shift staffing are identified as not being filled.

Evaluation Results: SAT_____ UNSAT_____

End time 6

OT-3701-ADM_031_SRO Rev 0 Personnel qualification report from Career Map (partial)

Item Days Completion First Name Last Name Title Complete Remaining Status Description Elizabeth A ******* Chemistry Technician Yes 65 Credit Gary L ******* Chemistry Technician No Adam C ******* Field Supervisor Yes 123 Credit Henry J ******* Field Supervisor Yes 5 Credit Jacob E ******* Field Supervisor Yes 265 Credit Michael C ******* Field Supervisor No Tony E ******* Field Supervisor Yes 44 Credit Adam J ******* Fire Brigade Member No Amanda L ******* Fire Brigade Member Yes 67 Credit Andrew J ******* Fire Brigade Member Yes 152 Credit Brian J ******* Fire Brigade Member No Colin J ******* Fire Brigade Member Yes 301 Credit David P ******* Fire Brigade Member Yes 28 Credit Henry J ******* Fire Brigade Member Yes 92 Credit Michael C ******* Fire Brigade Member No Bret B ******* I & C Technician Yes 31 Credit Chaz A ******* I & C Technician No David R ******* I & C Technician Yes 4 Credit Mason M ******* I & C Technician Yes 166 Credit Henry J ******* Perry Fire Brigade Leader Yes 92 Credit Michael C ******* Perry Fire Brigade Leader No Adam J ******* Perry Plant Attendant Yes 101 Credit Amanda L ******* Perry Plant Attendant Yes 109 Credit Andrew J ******* Perry Plant Attendant Yes 84 Credit Brian J ******* Perry Plant Attendant Yes 94 Credit Colin J ******* Perry Plant Attendant Yes 154 Credit David P ******* Perry Plant Attendant Yes 108 Credit Adam J ******* Plant Operator No Amanda L ******* Plant Operator Yes 109 Credit Andrew J ******* Plant Operator Yes 152 Credit Brian J ******* Plant Operator Yes 94 Credit Colin J ******* Plant Operator No David P ******* Plant Operator Yes 108 Credit Alan K ******* Radwaste Operations Supervisor Yes 154 Credit Robert J ******* Radwaste Operations Supervisor Yes 154 Credit 7

OT-3701-ADM_031_SRO Rev 0 Item Days Completion First Name Last Name Title Complete Remaining Status Description Henry J ******* Reactor Operator Yes 138 Credit Jacob E ******* Reactor Operator Yes 260 Credit Michael C ******* Reactor Operator No Alan D ******* RP Shift Tech Yes 154 Credit Nick A ******* RP Shift Tech Yes 93 Credit Adam C ******* Shift Engineer Yes 120 Credit Michael C ******* Shift Engineer No Adam C ******* Shift Manager No Tony E ******* Shift Manager Yes 104 Credit Adam C ******* Unit Supervisor Yes 120 Credit Jacob E ******* Unit Supervisor N0 Tony E ******* Unit Supervisor Yes 104 Credit 8

OT-3701-ADM_031_SRO Rev 0 JPM CUE SHEET INITIAL

Plant is in Cold Shutdown for a forced outage.

CONDITIONS:

A severe snow storm has made travel to the plant difficult.

Attached is a list of the oncoming crew personnel that are currently on site.

The report was developed from Career Map.

The Security Shift Supervisor reports that Site Protection is fully staffed.

INITIATING CUE:

As the off going Shift Manager, determine if the oncoming crew has sufficient personnel to meet minimum FENOC shift staffing requirements.

Document the results of your determination on the JPM CUE SHEET.

9

OT-3701-ADM_032SRO Rev 0 JOB PERFORMANCE MEASURE SETUP SHEET System: Administrative Time Critical: No Applicability: SRO Safety Function: Administrative Setting: Classroom / Simulator Validated: 20 minutes

References:

Master Setpoint List, SOI-D17A Rev 20, PDB-R0001 Rev 32, TS 3.7.2, EAL Classification Matrix Rev 1/3/17, PSI-0019, Rev 19, PYRM-ERS-5003 Rev 6, and NOP-LP-5004, Rev 1 Required Material: SOI-D17A, Process Radiation Monitoring System, PDB-R0001, Operational Requirements Manual, Technical Specifications, EAL Classification Matrix, PSI-0019, Emergency Action Level (EAL) Bases Document, PYRM-ERS-5003, Equipment Important To Emergency Response and NOP-LP-5004, Equipment Important to Emergency Response SVI-D17-T8057, Offgas Pretreatment Radiation Monitor Channel Functional for 1D17-K612 Tasks: 343-690-03-02 Determine the Operability Requirements for Tech Spec-Related Process Radiation Monitors IAW SOI-D17A Task Standard: Determine requirements for Offgas Pretreat rad monitor inoperability and simultaneous Offgas Pretreat and one Post Treat rad monitor inoperability.

K/A Data: 2.3.15 Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. SRO 3.1 OT-3701-ADM_032SRO Rev 0

1. Setup Instructions: None

2. Location / Method: Classroom / Administrative

3. Initial Condition 1: The plant has been at rated power for 40 days. During the performance of Tech Spec Rounds, the Operator found a large increase in the Offgas Pretreatment Radiation Monitor, 1D17-K612 reading with no alarm present. No changes to Plant radiation levels have been identified. Attached is the Tech Spec Rounds sheet and the Master Setpoint List printout for 1D17-K612.

Initial Condition 2: I&C is unable to adjust Offgas Pretreatment Radiation Monitor 1D17-K612 to within the allowable value. Additionally, 1D17-K601A, Off-Gas Post-Treatment Radiation Monitor Channel A just lost power.

4. Initiating Cue 1: As the Unit Supervisor, determine any required actions for the 1D17-K612 rad monitor based on readings from Tech Spec Rounds. Document your findings on the JPM CUE SHEET.

Initiating Cue 2: Evaluate any additional required actions based on the above additional conditions. Document your findings on the JPM CUE SHEET.

Start Time: End Time:

Candidate:

OT-3701-ADM_032SRO Rev 0 JPM BODY SHEET Standard: Performer obtains or simulates obtaining all materials, procedures, tools, keys, radios, etc before performing task.

Standard: Performer follows management expectations with regards to safety and communication standards.

Step 1 Review TS Rounds, Setpoint List, SOI-D17A, Tech Specs, and the ORM.

Standard: Candidate reviews TS Rounds, Setpoint List, SOI-D17A, Tech Specs, and the ORM and identifies >50% change in Offgas Pretreat reading.

Instructor Cue: None Notes: If asked, there is no known inoperability of the Offgas Post-Treat rad monitor.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 2 Determine applicability of TS 3.7.5.

Critical Step: Candidate evaluates Tech Spec 3.7.5, Main Condenser Offgas and contacts Chemistry to perform and analyze sample within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Instructor Cue: None Notes: SR 3.7.5.1 must be done within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per TS Bases SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_032SRO Rev 0 Step 3 Determine applicability of ORM 6.2.6.

Critical Step: Candidate evaluates ORM 6.2.6, Radiation Monitoring Instrumentation and determines that ORM 6.2.6 ACTION a is applicable since the Offgas treatment system is operating and the Rad Monitor alarm setpoint is not set in accordance with Technical Specification 3.7.5.

Instructor Cue: None Notes: Action a. - With a radiation monitoring instrumentation channel alarm/trip setpoint exceeding the value shown in Table 6.2.6-1, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

Since OG Pretreat reading is so high, Operator may elect to declare the rad monitor INOP instead of having I&C try to adjust.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 4 Review EAL Classification Chart, PSI-19, PYRM-ERS-5003, and NOP-LP-1004.

Standard: Candidate reviews procedures to determine that with D17-K612 failing non-conservatively, entry into the E-plan for EAL SU4.1 may be delayed and the Shift Manager will direct station personnel to implement compensatory measures.

Instructor Cue: After Candidate completes the above evaluations, read Initial Conditions/Initiating Cue #2 Notes: None SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_032SRO Rev 0 Step 5 Determine required actions for Offgas Pretreat Rad Monitor inoperability.

Critical Step: Candidate determines the following ORM 6.2.6, ACTION c is required:

Instructor Cue: None Notes: If Action a cannot be met, then Action c will be entered - Release via this pathway may continue for up to 30 days provided:

1. The Offgas System is not bypassed, and

2. The Offgas Post-treatment monitor is OPERABLE, and

3. Grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Otherwise, be in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 6 Determine any required actions for simultaneous inoperability of one channel of Offgas Post-treat rad monitor and Offgas Pretreat Rad Monitor.

Critical Step: Candidate evaluates ORM 6.2.6 Table 6.2.6-1 and determines that only one channel of Offgas Post-treat rad monitor is required with Offgas treatment system operating and no additional ORM 6.2.6, ACTIONS are required.

Instructor Cue: If asked, there is no known inoperability of the Offgas Post-treat rad monitor Channel B.

Notes: Terminate the JPM SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_032SRO Rev 0 Terminating Cue: Candidate determined the applicability and required Actions for inoperability of Offgas Pretreat rad monitor and determined no additional actions for simultaneous inoperability of one channel of Offgas Post-treat rad monitor.

Evaluation Results: SAT_____ UNSAT_____

End Time:

OT-3701-ADM_032SRO Rev 0 JPM CUE SHEET -2 INITIAL

I&C is unable to adjust Offgas Pretreatment Radiation Monitor 1D17-CONDITIONS: K612 to within the allowable value.

Additionally, 1D17-K601A, Off-Gas Post-Treatment Radiation Monitor Channel A just lost power.

INITIATING CUE:

Evaluate any additional required actions based on the above additional conditions.

Document your findings on the JPM CUE SHEET.

OT-3701-ADM_032SRO Rev 0 JPM CUE SHEET INITIAL

The plant has been at rated power for 40 days.

CONDITIONS:

During the performance of Tech Spec Rounds, the Operator found a large increase in the Offgas Pretreatment Radiation Monitor, 1D17-K612 reading with no alarm present.

No changes to Plant radiation levels have been identified.

Attached is the Tech Spec Rounds sheet and the Master Setpoint List printout for 1D17-K612.

INITIATING CUE:

As the Unit Supervisor, determine any required actions for the 1D17-K612 rad monitor based on readings from Tech Spec Rounds.

Document your findings on the JPM CUE SHEET.

OT-3701-ADM_032SRO Rev 0 Perry Nuclear Power Plant SETPOINTS PNPP No. 10038 ASSET LABEL: TYPE: REVISION DOCUMENT:

1D17K0612 NON RELAY 02-0288 SOURCE DOCUMENT#

CHI-71 N/A N/A SAFETY CLASS:

NON DESCRIPTION:

OFF-GAS PRETREAT RADIATION MONITOR INSTRUMENT TYPE: UNITS OF MEASURE:

RAD MON MR/HR CT/PT RATIO:

SETPOINT ALLOWABLE LIMIT ANALYTICAL LIMIT LEAVE AS IS ZONE RESET VALUE 14000 HI 28900 N/A +6000/-4000 N/A 1.5 LOW N/A N/A +/-0.5 N/A N/A N/A N/A N/A SETTINGS:

NOTES:

OFF-GAS PRETREATMENT I/T UNIT PLANT RADIATION MONITORING SP1/2-HIGH/LOW PLANT COMPUTER POINT ID D17EA007 DATA ENTRY NO.

INPUT DATE VERFIFIED DATE REV MADE CHKD APPVD DATE OT-3701-ADM_032SRO Rev 0 OT-3701-ADM_321_SRO Rev 0 CurrentJOB PERFORMANCE MEASURE SETUP SHEET System: E-plan Time Critical: Yes Alternate Path: No Applicability: SRO only Safety Function: Emergency Procedures / Plan Validated Time: 15 Minutes

References:

EPI-A1 Rev 27, EPI-A2 Rev 20, EPI-B1 Rev 26, & EAL Classification Matrix Rev 1/3/17 Required Material EPI-A2 - Emergency Actions Based On Event Classification EPI-B1 - Emergency Notification System EPI-A1 - Emergency Action Levels EAL Classification Matrix Forms 7983A, 7794, & 9100 Task: 344-533-05-02 Prepare Emergency Plan Follow-up Notification Task Standard: Perform Event Classification, complete Follow-Up Notification form, and deliver form to Communicator within 18 minutes.

K/A: 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation. SRO 4.5

1. Instructions: Setup ICS clock to 2140. Info to be given to the Operator include the completed forms and remaining blank forms from an E-Plan Packet

2. Location / Method: Remote Shutdown / Performance

3. Initial Condition: Plant was operating at 100% power with a B.5.b shuffle in progress in the Fuel Handling Building. At 2100, an Alert was declared based on a fuel handling accident in the FHB (RA2.2). A minor increase from the UNIT 1 Plant vent was identified. At 2110, the Initial Notification and Pager Message forms were completed. At 2112, Off-Site Notifications were completed. At 2120, NRC Form NOP-OP-1015 was completed and given to the Communicator for the ALERT Classification for the FHB Accident. At 2130 an increase in DW pressure and temperature was identified. At 2135, a manual Reactor SCRAM was successfully inserted. The current time is 2140 and DW pressure is 2.2 psig and stable, DW temperature is 170°F and stable.

4. Initiating Cue: With the information provided, complete the required E-Plan paperwork. Task is Time Critical Start Time End Time Operator OT-3701-ADM_321_SRO Rev 0 JPM BODY SHEET Standard: Performer obtains or simulates obtaining all materials, procedures, tools, keys, radios, etc before performing task.

Standard: Performer follows management expectations with regards to safety and communication standards.

Step 1 EPI-A2 - Emergency Actions Based On Event Classification 5.2 Follow-Up Actions 5.2.1 Emergency Coordinator:

1. Perform the Follow-Up Actions specified on Page 2 of 2 to the Event Classification Checklist (PNPP No. 7983A, Attachment 1) which include:

NOTE Follow-up notifications need to include all subsequent EAL entry conditions that are met but do not change the current emergency classification level.

NOTE The completed Follow-up Notification form (PNPP No. 7795) should be approved and forwarded to facility communicator(s) within 50 minutes of the event declaration or decision to revise PAR.

f. Direct a follow-up notification to the State of Ohio, Counties of Ashtabula, Geauga and Lake within 60 minutes of event classification, reclassification, or decision to revise offsite PAR per <EPI-B1>.

JPM Step 1 continued next page OT-3701-ADM_321_SRO Rev 0 JPM Step 1 continued With information provided, candidate reviews EAL Classification Matrix to determine if additional EALs are met.

Critical Step: Determines additional EAL of FA1.1 is met.

Instructor Cue:

If Operator requests a new E-Plan Packet, provide it to him when he identifies where to obtain.

The I&C Communicator is standing by.

Notes: An additional ALERT classification does not require an Initial Notification, but must be documented on the Follow-up Notification.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_321_SRO Rev 0 Step 2 Review Event Classification Checklist (Page 2) for Follow-Up Actions:

4. Direct that a dose projection be performed to verify offsite doses and to determine the need to generate or change offsite PARs per EPI-B8.

Standard: Operator reviews Event Classification Checklist and determines that a Dose Projection (MIDAS run) must be completed.

Instructor Cue: When directed, inform Operator you will perform Dose Projection (MIDAS).

Notes: Operator may direct either the Shift Engineer or Chemistry Tech to perform Dose Projection.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-ADM_321_SRO Rev 0 Step 3 Review Event Classification Checklist (Page 2) for Follow-Up Actions:

5. Direct completion of a Follow-up Notification form (PNPP No. 7795), approve, and forward to Communicators within 50 minutes of event classification or PAR revision.

Critical Step: Operator completes a Follow-Up Notification to the State and Local Counties within 18 minutes of reading Initiating Cue.

Instructor Cue: If asked the TSC is not yet operational.

When asked, provide Operator with MIDAS printout.

Notes: Perry Station goal is to hand Follow-Up Notification to I&C Communicator within 50 minutes of Event Declaration. However, the I&C Communicator must contact the first off-site entity within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of Event Declaration.

The Critical information on the Follow-Up Notification include:

All check blocks checked as on Answer Key.

Correct Date and Time in Block 3

Correct Date, Time, and EAL in Block 4 (May also list 2nd EAL here)

Wording containing the information from Answer Key in Block 5.c.

Approval Signature SAT ___ UNSAT ___ Time Completed _________

Comment(s): ________________________________________________________________

OT-3701-ADM_321_SRO Rev 0 Step 4 EVENT CLASSIFICATION CHECKLIST

6. OSC IN THE PROCESS OF BEING ACTIVATED.

7. TSC IN THE PROCESS OF BEING ACTIVATED.

8. Verify proper event classification using EPI A1.

9. [ALERT OR ABOVE ONLY] Direct the periodic notification of INPO and ANI using an Industry Event Notification form (PNPP No. 9596).

10. Verify completion of first follow up notification, and establish a schedule for subsequent periodic notifications.

11. Perform a 10CFR50.72(c) follow-up notification to the NRC using the Reactor Plant Event Notification Worksheet Form (NOP OP 1015 01) for further degradations in plant safety not requiring classification, results of evaluations/assessments of plant conditions, effectiveness of response or protective measures taken, information related to plant behavior that is not understood.

Standard: Candidate identifies will be commencing Steps 6 through 11 of Event Classification Checklist.

Instructor Cue: None Notes: Actual completion of remaining steps of Event Classification Checklist not required for this JPM.

Terminate the JPM.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Terminating Cue: Follow-Up Notification completed within 18 minutes and additional EAL identified.

Evaluation Results: SAT_____ UNSAT_____

End time OT-3701-ADM_321_SRO Rev 0 FENOC NUCLEAR POWER PLANT USE ONLY FOR UPDATING STATUS OF STATE / COUNTY USE ONLY CURRENT CLASSIFICATION. DO NOT USE FOLLOW-UP NOTIFICATION FORM FOR CHANGING CLASSIFICATIONS OR DATE: TIME:

Perry CHANGING PROTECTIVE ACTION RECOMMENDATIONS. Part A MESSAGE NO:

PNPP No. 7795 Rev. 9/17/12 EPI-B1

1. This is the Perry Nuclear Power Plant

2. This is: An Actual Emergency X A Drill

3. The following data represent the most current and accurate information, projections, and/or prognosis available as of: Time: Current time Date: Todays Date

4. The emergency classification remains at a(n):

GENERAL EMERGENCY SITE AREA EMERGENCY X ALERT UNUSUAL EVENT Declared at: Time: 2100 Date: Todays Date EAL RA2.2

5. a. Reactor is: at power decreasing power X in hot shutdown in cold shutdown in refueling mode

b. Prognosis is: X stable improving degrading

c. Additional pertinent information: Reactor was scrammed due to steam leak in Drywell.

Additional EAL FA1.1 (Alert) was met due to DW pressure > 1.68 psig.

6. General Information: (mark all applicable) X NA

a. Evacuation of non-essential personnel has been initiated. Completed at (time).

b. Offsite assembly for monitoring and decontamination purposes required.

c. Fire department has been requested. Yes No Currently on-site: Yes No

d. Ambulance has been requested. Yes No Currently on-site: Yes No

7. The radiological conditions are:

X a. A non-routine release of radioactive material, as a result of this event, is in progress.

The release is: X Airborne Liquid X Part B Attached

b. The release of radioactive material associated with this event has been terminated.

c. NO Radiological Release is in progress as a result of this event.

8. Recommended Protective Actions:

(a) Evacuation of people as follows:

Subareas: 1 2 3 4 5 6 7 Lake (circle)

And recommend administering Potassium Iodide (KI) in accordance with State procedures and the general public in unaffected areas should be advised to go indoors and monitor EAS broadcasts.

(b) Sheltering of people as follows:

Subareas: 1 2 3 (circle)

And recommend administering Potassium Iodide (KI) in accordance with State procedures and the general public in unaffected areas should be advised to go indoors and monitor EAS broadcasts.

X (c) None For Utility Use Only Approved: Candidates Signature OT-3701-ADM_321_SRO Rev 0 FENOC NUCLEAR POWER PLANT State/ County USE ONLY FOLLOW-UP NOTIFICATION FORM Part B Page 1 of 1 DATE: _____________ TIME: _____________

MESSAGE NO: _________________________

Dose Assessment Data Form SITE: PERRY ACTUAL MENU: H Summary For Calculation: (Raw Met Data for Today - 5 minutes ago) DATE PRINTED: Today - Now Current Input Data Adjusted Measured Precipitation Meteorological Data: Level Wind Direction from Wind Speed Wind Speed inch/15 min Stability Class 10 meters 250 degrees 6.0 mph 6.0 mph No Rain D Effective Release Monitor Conversion Concentration Flow Release Rate Monitor ID Point Reading Status* Units to uCi/cc (uCi/cc) Reading Status* Units (uCi/sec)

U1 VNT L 1 9.1E+01 0 cpm 5.3E-08 5.5E-06 9.2E+04 0 CFM 1.1E+02 U2 VNT L 1 8.3E+01 0 cpm 5.2E-08 5.3E-06 5.0E+04 0 CFM 1.0E+02 OGVENT L 1 4.8E+01 0 cpm 4.5E-08 2.2E-06 1.7E+04 0 CFM 1.8E+01 TB VNT L 1 5.7E+01 0 cpm 3.5E-08 2.0E-06 2.7E+05 0 CFM 2.6E+02

Status: 0=Good, 1=Questionable, 2=Bad, 7=Manual Override of Raw Data Release Characterization: Accident Type: Mix ID # 5: CONTAINMENT GAP RELEASE Type Noble Gases Iodines Particulates Total Units 4.8E+02 1.2E-00 2.7E-01 4.8E+02 Ci/sec Start of Release: A while ago - Today End of Release: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after start - Today Time of Trip: A while ago - Today Start of Projection: A while ago - Today Dose Projection Parameters: Projection Period: 24.00 Hours Estimated Remaining Release Duration: 4.00 Hours Dose Results Dispersion: (s/m3) Projected Dose: (mrem) Current Dose Rate: (mrem/hr)

Child Child TEDE/EDE at Distance X/Q TEDE Thyroid CDE TEDE Thyroid CDE EDE TEDE Rate EAB / Site Boundary 2.8E-05 6.2E-03 4.1E-02 0.0E+00 0.0E+00 0.0E+00 0.00 2 miles 5.2E-06 1.6E-03 6.9E-03 0.0E+00 0.0E+00 0.0E+00 5 miles 1.6E-06 5.8E-04 1.9E-03 0.0E+00 0.0E+00 0.0E+00 10 miles 5.5E-07 1.4E-04 5.6E-04 0.0E+00 0.0E+00 0.0E+00 Prepared by: _ Joel C. Teck_ ____________ /_ Today___ Reviewed by: _________________________ /____________

(Date) (Date)

OT-3701-ADM_308_SRO Rev 0 JPM CUE SHEET INITIAL

Plant was operating at 100% power with a B.5.b shuffle in progress in CONDITIONS: the Fuel Handling Building.

At 2100, an Alert was declared based on a fuel handling accident in the FHB (RA2.2).

At 2109, a minor increase from the UNIT 1 Plant vent was identified.

At 2110, the Initial Notification and Pager Message forms were completed.

At 2112, Off-Site Notifications were completed.

At 2120, NRC Form NOP-OP-1015 was completed and given to the Communicator for the ALERT Classification for the FHB Accident.

At 2130 an increase in DW pressure and temperature was identified

At 2135, a manual Reactor SCRAM was successfully inserted.

The current time is 2140 and DW pressure is 2.2 psig and stable, DW temperature is 170°F and stable.

INITIATING CUE: With the information provided complete the required E-Plan paperwork.

Task is Time Critical EVENT CLASSIFICATION CHECKLIST PNPP No. 7983A Rev. 6/20/13 Page 1 of 2 EPI-A2 Event classified as a/an: General Emergency Site Area Emergency X Alert Unusual Event At 2100 on Today .

Time Date Checklist completed by: Shift T. Manager (Shift Manager/TSC Operations Manager/ EOF Emergency Coordinator)

A. IMMEDIATE ACTIONS INITIAL TIME

1. Announce event classification and reason for declaring emergency over the Plant PA System.

STM 2101

2. Sound Plant Emergency Alarm. (Request Control Room to initiate Alarm) STM 2102

3. [CONTROL ROOM ONLY] Call the shift I&C technician to the Control Room as a communicator.

STM 2102 NOTE: For a GENERAL EMERGENCY, ensure that PAR is included using EPI-B8, Attachment 1, PAR Decision Flowchart.

4. Complete an Initial Notification form (PNPP No. 7794), approve, and forward to STM communicators within 10 minutes of decision to classify event or revise Protective Action Recommendations (PARs).

5a. Determine facilities to be activated using table below: (R-required; O-optional) STM 2102 Classification/Facility OSC TSC PIRT EOF JIC Unusual Event O O O O O Alert R R R O O Site Area Emergency R R R R R General Emergency R R R R R 5b. Announce facilities to be activated over the Plant PA System. STM 2102 NOTE: (1) Completion of form is delegated to the Security Coordinator once TSC is operational. (2) Activation of ERO pagers is NOT needed if required facilities have already been/are being mobilized. STM

6. Complete the Pager Messages form (PNPP No. 9100), approve, and forward to the SAS within 10 minutes of decision to classify event or revise PARs.

7. Complete an Reactor Plant Event Notification Worksheet form (NOP-OP-1015-01),

approve, and forward to communicators immediately following notification of the State STM of Ohio and local counties, but within 50 minutes of event classification.

NOTE: It may be prudent to delay implementation of accountability in situations where personnel safety could be jeopardized, such as a security event or severe weather.

8. [SITE AREA EMERGENCY or above] Initiate personnel accountability per EPI-B5, if N/A not yet implemented: X Not Required 8a. Direct Shift Manager, or designee, to initiate appropriate Emergency message over the Exclusion Area Paging System, and use PA feature to provide further guidance on offsite assembly if required.

8b. The Shift Manager, or designee, is to manually repeat the accountability message approximately every 5 minutes until accountability is completed, if the automated message is not working

9. Verify that notifications and/or requests for offsite support were completed by the SAS:

a. Fire Department (911) X Not Required STM 2115

b. Ambulance (911) X Not Required

c. Hospital: Primary - TriPoint; Backup - Lake West X Not Required

10. If the NRC requests that the ENS circuit remains open then verify that an individual knowledgeable in system operations is assigned to the NRC ENS Circuit to answer STM 2115 questions and inquiries.

REFER TO page 2 of 2 FOR LISTING OF FOLLOW-UP ACTIONS

EVENT CLASSIFICATION CHECKLIST PNPP No. 7983A Rev. 6/20/13 Page 2 of 2 EPI-A2 B. FOLLOW-UP ACTIONS INITIAL TIME

1. Verify completion of initial notifications:

X State of Ohio X Geauga County STM X Ashtabula County X Lake County X Nuclear Regulatory Commission (NRC)

2. On-call ERO notifications completed at 2125 hours. STM

3. Personnel accountability completed at hours.

No. unaccounted for: .

N/A

4. Direct that a dose projection be performed to verify offsite doses and to determine the need to generate or change offsite PARs per EPI-B8.

5. Direct completion of a Follow-up Notification form (PNPP No. 7795), approve, and forward to Communicators within 50 minutes of event classification or PAR revision.

6. OSC IN THE PROCESS OF BEING ACTIVATED. Not applicable; proceed to Step 7.

6a. Determine status of OSC activation, and appoint an interim OSC Coordinator if needed.

6b. When OSC declared operational, relocate Field Supervisor and POs/PAs to the OSC.

7. TSC IN THE PROCESS OF BEING ACTIVATED. Not applicable; proceed to Step 8.

7a. [ALERT OR ABOVE ONLY] Non-delegable Emergency Coordinator duties transferred to the TSC:

(1) Event Classification at hours.

(2) Protective Action Recommendations (PARs) at hours.

(3) Offsite Notifications at hours.

7b. If requested, relocate the Control Room Communicator to TSC to expedite the transfer of offsite notification responsibilities.

8. Verify proper event classification using EPI-A1. If event classification is changed, start a new Event Classification Checklist.

9. [ALERT OR ABOVE ONLY] Direct the periodic notification of INPO and ANI using an Industry Event Notification form (PNPP No. 9596).

NOTE: Periodic follow-up notifications are to be performed on an approximately hourly basis, unless alternate schedule established with approval of offsite agency(ies).

10. Verify completion of first follow-up notification, and establish a schedule for subsequent periodic notifications.

State of Ohio Ashtabula County Lake County Geauga County

11. Perform a 10CFR50.72(c) follow-up notification to the NRC using the Reactor Plant Event Notification Worksheet Form (NOP-OP-1015-01) for further degradations in plant safety not requiring classification, results of evaluations/assessments of plant conditions, effectiveness of response or protective measures taken, information related to plant behavior that is not understood.

12. EOF IN THE PROCESS OF BEING ACTIVATED. Not applicable; proceed to Step 13.

12a. Non-delegable Emergency Coordinator duties transferred to the EOF: NA (1) Event Classification at hours.

(2) Protective Action Recommendations (PARs) at hours.

(3) Offsite Notifications at hours.

13. Determine the need for additional emergency facilities. Not Required, or as indicated: OSC TSC EOF PIRT JIC

14. [SITE AREA EMERGENCY, GENERAL EMERGENCY OR @ EMERGENCY COORDINATOR DISCRETION]

Contact the duty Fleet Emergency Preparedness individual via pager per the instructions in the ERO Telephone directory and request mobilization of the Corporate Planning Assistance Center in accordance with the NOBP-LP-5001.

15. Review FOLLOW-UP ACTIONS listed in Section 5.2 of EPI-A2, and perform as applicable.

REFER TO FORM 7983B FOR EVENT TERMINATION ACTIONS

FENOC NUCLEAR POWER USE FOR: STATE / COUNTY USE ONLY INITIAL CLASSIFICATIONS, PLANT INITIAL NOTIFICATION CHANGES IN CLASSIFICATIONS, DATE: TIME:

FORM CHANGES IN PROTECTIVE ACTION Perry RECOMMENDATIONS. MESSAGE NO:

PNPP No. 7794 Rev. 6/8/11 EPI-B1 EVENT TERMINATION

1. This is the: Perry Nuclear Power Plant

2. This is: An Actual Emergency A Drill

3. a. A(n) GENERAL SITE AREA EMERGENCY ALERT UNUSUAL EVENT was declared at: 2100 _ on Today__ based on EAL: RA2,2 (TIME) (DATE)

b. The Emergency situation has been terminated at: _ on ___

(TIME) (DATE)

c. The Protective Action Recommendation is being changed at: _____ on _

(TIME) (DATE)

4. Brief non-technical description of event: Significant lowering of water level above, or damage to, irradiated fuel

5. The radiological conditions are:

a. A non-routine release of radioactive material, as a result of this event, is in progress.

b. The release of radioactive material associated with this event has been terminated.

c. NO Radiological Release in progress as a result of this event.

6. Utility Protective Action Recommendations (PARs):

a. Evacuation:

(check applicable subareas) 1 2 3 4 5 6 7 LAKE AND that potassium iodide (KI) be administered to the general public in accordance with State procedures.

The general public in unaffected areas should be advised to go indoors and monitor EAS broadcasts.

b. Sheltering:

(check applicable subareas) 1 2 3 AND Evacuate the Lake AND that potassium iodide (KI) be administered to the general public in accordance with State procedures.

The general public in unaffected areas should be advised to go indoors and monitor EAS broadcasts.

c. None For Utility Use Only Approved: Shift T. Manager

PAGER MESSAGES PNPP No. 9100 Rev. 9/8/14 Page 1 of 1 EPI-B1 CONTROL ROOM/TSC SECURITY COORDINATOR INSTRUCTIONS:

1. Select appropriate Scenario ID number.

2. Initiate the notification by forwarding to SAS or by forwarding the information contained within the form to an ERS representative.

TSC SECURITY COORDINATOR/SAS OPERATOR INSTRUCTIONS:

Using the information on this form, initiate the notification of the Emergency Response Organization per SPI-0032.

Event

() Scenario ID No. Message Narrative Code 1 1111 Unusual Event - No facility activation.

2 2222 Alert - OSC, TSC, and PIRT to be activated.

3 3333 Site Area Emergency - OSC, TSC, EOF, PIRT, and JIC to be activated.

4 4444 General Emergency - OSC, TSC, EOF, and JIC to be activated.

51 5555 Event Termination 52 5555 OSC to be staffed.

53 5555 TSC to be staffed.

54 5555 PIRT to be staffed.

55 5555 OSC, TSC and PIRT to be staffed.

56 5555 EOF to be staffed.

57 5555 JIC to be staffed.

61 6666 ERO Teams respond to your offsite facility.

62 6666 Site Inaccessible, ERO Teams respond to your offsite facility.

81 8888 Drive-In Drill - OSC, TSC & EOF to be activated.

91 9999 Weekly Pager Test (Shift Manager approval not required) 92 9999 Unannounced Pager Test (Shift Manager approval not required) 93 9999 Security Pager Test (Shift Manager approval not required)

Approved by: Today / 2110 / Shift T. Manager Date Time Emergency Coordinator Delivered to: Today / 2111 / SAS Operator Date Time Name Activated by: Today / 2112 / Officer Herald

NOP-OP-1015-01 Rev. 01 EN #

REACTOR PLANT EVENT NOTIFICATION WORKSHEET Page 1 of NRC OPERATION TELEPHONE NUMBER: PRIMARY - 800-532-3469 or 301-816-5100 BACKUPS - [1st] 301-951-0550 or 800-449-3694 [2nd] 301-415-0550 [3rd] 301-415-0553 NOTIFICATION DATE & TIME FACILITY OR ORGANIZATION UNIT NAME OF CALLER CALL BACK NUMBER 2120 Perry Power Plant 1 Brock 440-280-5647 EVENT TIME AND ZONE EVENT DATE POWER/MODE BEFORE POWER/MODE AFTER 2100 Eastern Today 100% / 1 100% / 1 EVENT CLASSIFICATIONS 1-Hr. Non-Emergency 10 CFR 50.72(b) (1) (v)(A) Safe S/D Capability AINA GENERAL EMERGENCY GEN/AAEC TS Deviation (50.54x) ADEV (v)(B) RHR Capability AINB SITE AREA EMERGENCY SIT/AAEC 4-Hr. Non-Emergency 10 CFR 50.72(b) (2) (v)(C) Control of Rad Release AINC ALERT ALE/AAEC (i) TS Required S/D ASHU (v)(D) Accident Mitigation AIND UNUSUAL EVENT UNU/AAEC (iv)(A) ECCS Discharge to RCS ACCS (xii) Offsite Medical AMED 50.72 NON-EMERGENCY (see next columns) (iv)(B) RPS Actuation (scram) ARPS (xiii) Loss Comm/Asmt/Resp ACOM PHYSICAL SECURITY (73.71) DDDD (xi) Offsite Notification APRE 60-Day Optional 10 CFR 50.73 (a)(1)

MATERIAL/EXPOSURE B??? 8-Hr. Non-Emergency 10 CFR 50.72(b) (3) Invalid Specified System Actuation AINV FITNESS FOR DUTY HFIT (ii)(A) Degraded Condition ADEG Other Specified Requirement (Identify)

OTHER UNSPECIFIED REQMT.(see last column) (ii)(B) Unanalyzed Condition AUNA 10 CFR 50.72(c) NONR INFORMATION ONLY NINF (iv)(A) Specified System Actuation AESF NONR DESCRIPTION Include: Systems affected, actuations and their initiating signals, causes, effect of event on plant, actions taken or planned, etc. (Continue on next page)

While performing a B.5.b shuffle in the Fuel Handling Building, a spent fuel bundle was dropped resulting in Fuel Handling Building Area and Process radiation alarms and a slight increase in the release rate from the #1 Plant Vent.

NOTIFICATIONS YES NO WILL BE ANYTHING UNUSUAL OR NRC RESIDENT X NOT UNDERSTOOD? YES (Explain above) X NO STATE(s)

X DID ALL SYSTEMS X

X FUNCTION AS REQUIRED? YES NO (Explain above)

LOCAL OTHER GOV AGENCIES X MODE OF OPERATION ESTIMATED RESTART DATE: ADDITIONAL INFO ON MEDIA/PRESS RELEASE X

UNTIL CORRECTED:

1 (MM/DD/YYY)

N/A NEXT PAGE?

YES X NO

NOP-OP-1015-01 Rev 01 REACTOR PLANT EVENT NOTIFICATION WORKSHEET (CONTINUED) Page 2 of RADIOLOGICAL RELEASES: CHECK OR FILL IN APPLICABLE ITEMS (specific details/explanations should be covered in event description).

LIQUID GASEOUS UNPLANNED PLANNED ONGOING TERMINATED RELEASE RELEASE RELEASE RELEASE MONITORED UNMONITORED OFFSITE ODCM RM AREAS RELEASE EXCEEDED ALARMS EVACUATED PERSONNEL EXPOSED OR CONTAMINTED OFFSITE PROTECTIVE ACTIONS RECOMMENDED *State release path in description RELEASE RATE (Ci/sec) % ODCM/T.S. LIMIT HOO GUIDE TOTAL ACTIVITY (Ci) % ODCM/T.S. HOO GUIDE LIMIT Noble Gas 0.1 Ci/sec 1000 Ci Iodine 10 uCi/sec 0.01 Ci Particulate 1 uCi/sec 1 mCi Liquid (excluding tritium and 10 uCi/min 0.1 Ci dissolved noble gases)

Liquid (tritium) 0.2 Ci/min 5 Ci Total PLANT STACK CONDENSER/AIR EJECTOR MAIN STEAM LINE SG BLOWDOWN OTHER RAD MONITOR READINGS ALARM SETPOINTS

% ODCM/T.S. LIMIT (if applicable)

RCS OR SG TUBE LEAKS: CHECK OR FILL IN APPLICABLE ITEMS: (specific details/explanations should be covered in event description).

LOCATION OF THE LEAK (e.g., SG #, valve, pipe, etc.)

LEAK RATE UNITS: gpm/gpd T.S. LIMITS SUDDEN OR LONG-TERM DEVELOPMENT LEAK START DATE TIME COOLANT ACTIVITY PRIMARY SECONDARY AND UNITS:

LIST OF SAFETY RELATED EQUIPMENT NOT OPERATIONAL EVENT DESCRIPTION (continued from page 1)

OT-3701-C11_014SRO Rev 0 JOB PERFORMANCE MEASURE SETUP SHEET System: C11 CRDH Time Critical: No Alternate Path: No Applicability: SRO Safety Function: 1. Reactivity Control Validated Time: 30 Minutes

References:

Technical Specification 3.1.3 & 3.1.4, SVI-C11-T1006 Rev 21, & OAI-1701 Rev 15 Required Material: Technical Specifications 3.1.3 & 3.1.4, SVI-C11-T1006, Time Events Analyzer Scram Time Data, Full Core Display diagram, Form PNPP No.

7158 and calculator Task: 299-613-03-01 Comply with Requirements of Surveillance Test Control 299-550-03-06 Complete Surveillance Test Documentation 299-717-03-01 Notify US if Expected SVI Test Results are not Achieved 341-530-03-02 Complete Active and Potential LCO Tracking Sheets Task Standard: Evaluate scram time data using SVI-C11-T1006 to determine if control rods are Slow or INOP per Tech Specs and evaluate Tech Specs.

K/A: 201003 Control Rod and Drive Mechanism - A2.10 Ability to (a) predict the impacts of the following on the Control Rod And Drive Mechanism; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Excessive SCRAM time for a given drive mechanism. Importance: 3.4 and 2.2.23, Ability to track Technical Specification limiting conditions for operations.

Importance 4.6 OT-3701-C11_014SRO Rev 0

1. Instructions: Provide student with SVI-C11-T1006, Time Events Analyzer Scram Time Data, and Full Core Display diagram. Provide calculator if requested.

2. Location / Method: Simulator or Control Room / Administrative performance.

3. Initial Condition: Plant is in Mode 4. Vessel Leak Test was completed along with scram time testing per SVI-C11-T1006. During the testing, ICS became unavailable. Time events analyzer was primary method used to time the last 8 control rods. Currently no Control Rods have an Active or Potential LCO documented.

SVI-C11-T1006 provided

Time Events Analyzer Scram Time Data provided

Full Core Display provided

4. Initiating Cue: As the Unit Supervisor evaluate Time Events Analyzer data sheet for scram time testing per Section 5.2.3 using Attachment 1 of SVI-C11-T1006.

Complete SVI-C11-T1006 Scram Timing Data Sheet column 5.2.3.1.

Start Time End Time Operator OT-3701-C11_014SRO Rev 0 JPM BODY SHEET Standard: Performer obtains or simulates obtaining all materials, procedures, tools, keys, radios, etc before performing task.

Standard: Performer follows management expectations with regards to safety and communication standards.

Step 1 SVI-C11-T1006, Control Rod Maximum Scram Insertion Time 5.2.3 Test Completion (Attachment 3)

1. VERIFY scram time acceptability as follows:

a. EVALUATE speed per one of the following:

-- IF using ICS, THEN CONFIRM an Accept on ICS printout(s) or use RSTRZZ to view the database.

-- IF using the time events analyzer, THEN CONFIRM the scram time(s)

Max_Time(s) calculated per Attachment 1 or by using the following limiting Value(s):

Reactor Pressure Limiting (psig) Notch Max Time (sec) 0 to 600, inclusive 13 0.94 600 to 950 13 1.13 950 to 1050, inclusive 43 0.30 29 0.78 13 1.40 Critical Task: Calculate Scram time values per Attachment 1 for a reactor pressure of 1010 psig.

Notch 43 time is 0.306 Notch 29 time is 0.816 Notch 13 time is 1.478.

Instructor Cue: None Notes: Test completed at 1010 psig. Can use Limiting Values listed for some rods but will need to complete Attachment 1. For Rod 22-47:

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-C11_014SRO Rev 0 Step 2 5.2.3.1.a: Evaluate speed per one of the following,

-- IF using the time events analyzer, THEN CONFIRM the scram time(s)

Max_Time(s) calculated per Attachment 1 or by using the following limiting Value(s):

Attachment 1 Calculated values below:

Reactor Pressure Calculated (psig) Notch Max Time (sec) 1010 43 0.306 29 0.816 13 1.478 Critical Step: Evaluates Scram Time Data & determines scram times for rods 18-55, 50-23, & 50-43 are not acceptable.

Instructor Cue: If asked about typo on Attachment 1, tell the operator to use the appropriate Step (Step 4.0.3) and the typo will be documented on a Notification.

Notes: Attachment 1 has a typo. It refers to pressure recorded in Step 4.0.2. It should be Step 4.0.3 SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-C11_014SRO Rev 0 Step 3 5.2.3.1.a: Evaluate speed per one of the following,

-- IF using the time events analyzer, THEN CONFIRM the scram time(s)

Max_Time(s) calculated per Attachment 1 or by using the following limiting Value(s):

Attachment 1 Calculated values below:

Reactor Pressure Calculated (psig) Notch Max Time (sec) 1010 43 0.306 29 0.816 13 1.478 Standard: Evaluates Scram Time Data & determines scram times for rods 18-27, 22-47, 46-23, 14-51, & 18-51 are acceptable.

Instructor Cue:

If asked about typo on Attachment 1, tell the operator to use the appropriate Step (Step 4.0.3) and the typo will be documented on a Notification.

After evaluating Scram time data, direct Operator to evaluate Technical Specifications if necessary.

Notes: Attachment 1 has a typo. It refers to pressure recorded in Step 4.0.2. It should be Step 4.0.3 SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-C11_014SRO Rev 0 Step 4 5.2.3.1.b: If test performed to satisfy Technical Specification SR 3.1.4.2, review all slow rods to the representative sample plan and confirm no more than 7.5% of the control rods in the representative sample are slow, or EXPAND the sample size until either this 7.5% criterion is satisfied or the total number of slow control rods (throughout the core, from all surveillances) exceeds the LCO limit. Technical Specification BASES SR 3.1.4.2 Standard: Step 5.2.3.1b is N/A Instructor Cue: None Notes: If done for SR 3.1.4.2, 7.5% of 177 is 13, 2 rods are slow and one is inoperable.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 5 5.2.3.1.c: If the rod fails to scram on the first attempt or if the scram time to position 13 is >7 seconds, immediately declare it INOPERABLE, fully insert it, and comply with the applicable Technical Specification CONDITIONS and REQUIRED ACTION statements.

Critical Step: Identifies PLCO for rod 50-23 against Tech Spec 3.1.3 Condition C.

Instructor Cue: None Notes: Scram Time 7.5 seconds to position 13.

SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

OT-3701-C11_014SRO Rev 0 Step 6 5.2.3.1.d: If the rod is slow on the first attempt, then immediately declare it as such and comply with applicable Technical Specification CONDITIONS and REQUIRED ACTION statements.

Critical Step: Identifies PLCO for rods 18-55 and 50-43 against Tech Spec 3.1.4 Condition A.

Instructor Cue: None Notes: None SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Step 7 SVI-C11-T1006:

Completes Attachment 3 columns 5.2.3.1.

Standard: Per initiating cue complete columns 5.2.3.1 with SAT/UNSAT or Yes/No for all rods tested.

Instructor Cue: None Notes: None SAT ___ UNSAT ___

Comment(s): ________________________________________________________________

Terminating Cue: Evaluation of scram time data is completed and PLCOs identified for Tech Spec 3.1.3 and 3.1.4.

Evaluation Results: SAT_____ UNSAT_____

End Time OT-3701-C11_014SRO Rev 0 JPM CUE SHEET INITIAL

Plant in Mode 4.

CONDITIONS:

Vessel Leak Test was completed along with scram time testing per SVI-C11-T1006.

During the testing, ICS became unavailable.

Time events analyzer was primary method used to time the last 8 control rods.

Currently no Control Rods have an Active or Potential LCO documented.

o SVI-C11-T1006 provided o Time Events Analyzer Scram Time Data provided o Full Core Display provided INITIATING CUE: As the Unit Supervisor:

Evaluate Time Events Analyzer data sheet for scram time testing per section 5.2.3 using Attachment 1 of SVI-C11-T1006.

Complete SVI-C11-T1006 Scram Timing Data Sheet column 5.2.3.1.

OT-3701-C11_014SRO Rev 0 Time Events Analyzer Scram Time Data Control Time to position (in seconds)

Rod 43 29 13 18-55 .316 .818 4.530 18-27 .290 .775 1.382 50-43 .305 .815 2.498 22-47 .306 .816 1.478 46-23 .280 .765 1.302 14-51 .276 .733 1.294 50-23 .535 1.980 7.500 18-51 .286 .759 1.234 OT-3701-C11_014SRO Rev 0 Functional Location:

TEST COVER SHEET SVI-C11-T1006 (PY: DATA PACKAGE COVER SHEET)

Maintenance Plan #:

NOP-WM-2003-01 Rev. 02 Page 1 of 2 Order #:

Performance Section:

TEST COMPLETION FULL PARTIAL FAILED TECHNICAL SPECIFICATION DATA ACCEPTABLE UNACCEPTABLE NA ACCEPTABLE OUTSIDE LAIZ ACCEPTABLE OUTSIDE AV (Allowable Value)

Condition Report #

NON TECHNICAL SPECIFICATION/ ACCEPTABLE UNACCEPTABLE NA PERIODIC TEST DATA TEST CREDIT CREDIT NO CREDIT Additional Surveillance Test Crediting: Credit ST/PT: Plan: Order Number: Y N ST/PT: Plan: Order Number: Y N Work Start (Prerequisites) Approval: L.S Super / L.S Super Today 2 hrs ago Print Name Signature Date and Time Test Start Approval: Rex X. Opter / Rex X. Opter Today 30 min ago Print Name Signature Date and Time Test Completed (Test Leader): /

Print Name Signature Date and Time Lead Work Group Supervisor Review: /

Print Name Signature Date and Time Comments:

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21 1 of 32 CONTROL ROD MAXIMUM SCRAM INSERTION TIME Surveillance Instruction Effective Date: 10-25-16 Preparer: Thomas Naylor / 10-13-16 Date Approver: Paul Bordley / 10-13-16 Date Working Copy Verified - Today

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1.0 DESCRIPTION

1.1 Scope This instruction determines the control rod scram times to fully satisfy the Technical Specification SURVEILLANCE REQUIREMENTS of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3 and SR 3.1.4.4 and SR 3.1.3.3.

Satisfactory performance of this instruction for an individual control rod also satisfies the Exercise Open (EO), Exercise Closed (EC), Stroke Time (ST), and Fail Safe (FS) test requirements of Technical Specification 5.5.6 in accordance with PAP-1101, Inservice Testing of Pumps and Valves, for the following:

Scram Discharge Riser Check Valve, 1C11-EP-114 (EO),

(IST Valve 1C11-114, Scram Discharge Header Check Valve)

Accumulator Supply Check Valve, 1C11-EP-115 (EO),

(IST Valve 1C11-115, Charging Water Header Check Valve)

Inlet Scram Valve, 1C11-EP-126 (EO, ST, FS)

(IST Valve 1C11-126, Scram Inlet Valve)

Outlet Scram Valve, 1C11-EP-127 (EO, ST, FS)

(IST Valve 1C11-127, Scram Exhaust Valve)

Measurement uncertainty associated with ICS data and the Dranetz time events analyzer has been adequately addressed in determining acceptable performance of scram timing and analysis of data.

1.2 Frequency Special Requirement

1. SR 3.1.4.1, SR 3.1.3.3, and ISTP Valve Relief Request VR-1; for each control rod with reactor steam dome pressure t 950 psig, prior to exceeding 40% RTP after each reactor shutdown t 120 days, and

2. SR 3.1.4.2, SR 3.1.3.3, and ISTP Valve Relief Request VR-1; for a representative sample with reactor steam dome pressure t 950 psig every 200 days cumulative operation in MODE 1, and

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3. SR 3.1.4.3, SR 3.1.3.3, and ISTP Valve Relief Request VR-1; for each affected control rod with any reactor steam dome pressure:

a. Prior to declaring the control rod OPERABLE after work on the control rod or the CRD System that could affect scram time.

4. SR 3.1.4.4, SR 3.1.3.3, and ISTP Valve Relief Request VR-1; for each affected control rod with reactor steam dome pressure t 950 psig:

a. Prior to exceeding 40% RTP after fuel movement within the affected core cell and

b. Prior to exceeding 40% RTP after work on the control rod or the CRD System that could affect scram time.

1.3 Technical Specification Applicable MODES 1 and 2 2.0 PRECAUTIONS AND LIMITATIONS

1. Document all control rod movements per FTI-B0002, Control Rod Movements. The independent verification of control rod movement in FTI-B0002 satisfies Technical Specification SURVEILLANCE REQUIREMENT SR 3.10.7.1.

2. If performing this SVI in conjunction with ISI-B21-T1300-1, Reactor Coolant System Leakage Pressure Test, reactor pressure may increase slightly as a result of individual control rod scrams. A full core scram may increase pressure beyond the 1050 psig test condition. Review the termination criteria/response to over pressure condition in the ISI.

Discontinue testing if an SRV opens.

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3. During single rod scrams, return HCU Test Switches to NORMAL when at least 10 seconds have elapsed after the rod reaches the full-in position. The reset should be done as quickly as possible after the 10 second minimum is reached. This will minimize challenges to the scram discharge volume high water level scram, while still allowing adequate time for the ICS to measure the scram time. If the HCU Test Switches are returned to NORMAL while the control rod is still traveling, the CRD may be damaged.

4. A malfunction in the Single Rod Insertion (SRI) test switch can cause erroneous SCRAM time results. SCRAM times <200 msec to notch position 43 are indicative of an SRI switch malfunction. If erroneous data is received during post maintenance testing or surveillance testing, the associated control rod should be SCRAM timed a second time with the opposite SRI switch sequence (i. e., A then B vs. B then A) or, if surveillance testing, another rod can be selected for the surveillance sample. A notification should then be initiated for the affected test switch.

5. In MODES 1 and 2, data evaluation must be performed immediately after completion of single rod scrams to ensure timely disposition of potentially slow or INOPERABLE control rods.

6. If the scram time from fully withdrawn to position 13 is greater than 7 seconds on the first attempt, immediately declare the control rod INOPERABLE per Technical Specification SURVEILLANCE REQUIREMENT SR 3.1.3.3.

7. If a rod fails to scram on the first attempt, immediately declare it INOPERABLE, and comply with the applicable Technical Specification CONDITION and REQUIRED ACTION statements.

8. For scram time testing with reactor pressure below 950 psig, in order to demonstrate control rod OPERABILITY, only the scram time to notch 13 is measured. Any control rod declared OPERABLE as a result of satisfying the scram time to notch 13 is still required to be tested again when reactor pressure exceeds 950 psig.

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9. An OPERABLE control rod with scram times not within the limits of Technical Specification Table 3.1.4-1 is considered slow. If a control rod is slow on the first attempt, immediately declare it as such and comply with the applicable Technical Specification CONDITION and REQUIRED ACTION statements.

10. No more than 13 OPERABLE control rods shall be slow and no OPERABLE control rod that is slow shall occupy a location adjacent to another OPERABLE control rod that is slow or a withdrawn control rod that is stuck. Technical Specification LCO 3.1.4.

11. Unusually long delay times to position 43 are indicative of scram pilot valve deterioration, reference G.E. SIL 441.

12. A representative sample of control rods for Technical Specification SURVEILLANCE REQUIREMENT SR 3.1.4.2 contains at least 10% (18) of the control rods. The sample remains representative if no more than 7.5% of the control rods in the tested sample are determined to be slow. Per the criteria in Technical Specification Table 3.1.4-1, if more than 7.5% of the sample is declared to be slow additional control rods are tested until this 7.5% criterion (e.g., 7.5% of the entire sample size) is satisfied, or until the total number of slow control rods (throughout the core, from all surveillances) exceeds the LCO limit of 13 rods.

Technical Specification BASES SR 3.1.4.2.

13. The control rod drive (CRD) pumps shall be isolated from the associated scram accumulator for single rod scram tests. Technical Specification 3.1.4 SURVEILLANCE REQUIREMENT NOTE.

14. With any control rod scram accumulator INOPERABLE with reactor steam dome pressure t600 psig, the associated control rod scram time is declared slow, provided the associated control rod scram time was within the limits of Technical Specification Table 3.1.4-1 during the last scram time surveillance. Technical Specification REQUIRED ACTION 3.1.5.A.1 NOTE and 3.1.5.B.2.1 NOTE.

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21 6 of 32 3.0 MANPOWER AND EQUIPMENT 3.1 Manpower/Location/Communication 3.1.1 Using ICS Data Following a Reactor Scram

1. Technically qualified individual to obtain data from ICS computer and evaluate acceptance criteria.

3.1.2 Performing Single Rod Scrams in Mode 1 or 2

1. Control Room operator to drive control rods, direct single rod scrams, and observe nuclear instrumentation.

2. Licensed Operator positioned at the control rod drive HCUs on Containment El. 620 to perform switch manipulations and collect data as required for testing.

3. Plant Operator at the control rod drive HCUs on Containment El. 620 to position valves and collect data as required for testing.

4. Technically qualified individual to direct test, take data from ICS computer or time events analyzer, and confirm individual rod test acceptance.

5. Communication link between CRD HCU operator and the Control Room operator.

6. Technically qualified individual to verify EP139 solenoids are energized.

3.1.3 Performing Single Rod Scrams in Mode 3, 4 or 5

1. Test Director in Control Room to coordinate teams

2. Control Room operator to drive control rods, direct single rod scrams, and observe nuclear instrumentation. (Control Room Team 1)

3. Licensed operator and (2) Perry Plant Operators at the control rod drive HCUs on Containment El. 620 to maintain communication and operate HCU components (Containment Team 1)

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4. Licensed Operator positioned at the control rod drive HCUs on Containment El. 620 to maintain communication and perform switch manipulations and collect data as required for testing. (Containment Team 2)

5. Technically qualified individual to take data from ICS computer or time events analyzer, and confirm individual rod test acceptance. (Control Room Team 2)

6. (2) Perry Plant Operators at the control rod drive HCUs on Containment El. 620 to maintain communication and operate HCU components (Containment Team 3)

7. Communication link between Test Director and Containment Team 1

8. Communication link between Control Room Team 1 and Containment Team 2.

3.2 Required Measuring and Test Equipment (M&TE)

1. Dranetz 2000 Time Events Analyzer, or equivalent, if not using the ICS Computer scram time program.

2. IF Performing Single Rod Scrams in Mode 1 or 2:

a. Type 134 current probe amplifier with 6021 probe and Fluke 8050 or 8600 multimeter or equivalent for verifying scram pilot solenoids energized. (Preferred)

b. Fluke 8050 or 8600 multimeter or equivalent without current probe amplifier to measure voltage across scram pilot solenoids.

(Alternate to a above) 3.3 Additional Tools and Equipment

1. ICS computer for scram data, if applicable.

2. HCU venting tool, if needed.

3. Regular screwdriver for opening HCU electrical access cover, if needed.

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4. ISI-B21-T1300-1, Reactor Coolant System Leakage Pressure Test, if needed.

5. FTI-B0002, Control Rod Movements, if needed.

6. Stop watch (or similar time device), if desired to assist in confirming the elapse of 10 seconds after a rod reaches full-in position.

4.0 PREREQUISITES Initials

1. IF performing this SVI to evaluate data following a reactor scram, THEN PERFORM the following:

a. MARK Unit Supervisors Authorization to Start Prerequisites signature N/A. N/A

b. MARK Sections 5.1, Single Rod Scram Time Test (Mode 1 or 2), 5.2, Single Rod Scram Time Test (Mode 3, 4 or 5), and 5.4, Plant/System Restoration, N/A.

c. PROCEED TO Section 5.3, Full Core Scram Time Test

2. IF performing individual rod scrams, THEN OBTAIN Unit Supervisors Authorization to Start Prerequisites signature on the Cover Sheet. RXE NOTE Covers for the adjacent HCUs may be located in the GEM box in the SVI room in containment.

3. RECORD reactor vessel dome pressure.

(ICS pt. C34EA030 preferred) 1010 Pressure = ___________ psig RXE

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21 9 of 32 SECTION 4.0 Initials

4. VERIFY control rod drive system is in normal operation per the following instructions:

x SOI-C11(CRDH) RXE x SOI-C11(RCIS) RXE x SOI-C11(HCU) for control rods to be tested. RXE

5. IF using the time events analyzer, THEN VERIFY its calibration current. RECORD calibration due date and M&TE Functional Location.

Functional LocationL70N0120

__________Cal due date Next Month

___________ RXE

6. IF using the time events analyzer, THEN CONNECT it to 1H13-P610, CONTROL ROD TEST PANEL. RXE NOTE Additional requirements concerning choosing the representative sample plan may be desired by the RSE or Reactor Engineering.

7. IF performing per Technical Specification SR 3.1.4.2, THEN DETERMINE the representative sample of rods as follows:

a. DETERMINE those rods that have not been tested per SR 3.1.4.2 during the present cycle. N/A

b. DO NOT include rods requiring testing for specific reasons (re-test, etc.) in the sample.

c. VERIFY at least 1 rod selected in each core quadrant.

d. SELECT at least 18 rods.

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21 10 of 32 SECTION 4.0 Initials

8. IF performing individual rod scrams in MODE 3, 4 or 5, THEN PERFORM the following:

a. VERIFY that the one-rod-out interlock has been demonstrated OPERABLE per SVI-C71-T0427, Reactor MODE Switch Refuel Mode Channel Functional. RXE

b. VERIFY the Reactor MODE Switch in REFUEL position. RXE

c. VERIFY SVI-C71-T0051 has been completed since performance of maintenance on the scram air header or any HCU. RXE

9. IF performing individual rod scrams, THEN VERIFY approved Control Rod Movement Sheets cover all rods to be tested. RXE

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21 11 of 32 5.0 SURVEILLANCE INSTRUCTION Initials This SVI is a combination of 177 control rod scram timing tests that may be done simultaneously (reactor scram) or individually. As such, several control rods may be in various stages of testing as long as compliance with Technical Specifications is maintained for the specific plant MODE. For example, only one rod may be withdrawn in MODE 4.

Restoration of the HCU may be deferred until a later time, without holding the SVI open, as long as controls are in place to restore the HCUs to service. Generally, this would be LCO tracking, though other methods may be used.

5.1 SINGLE CONTROL ROD SCRAM TIME TEST (Mode 1 or 2)

1. OBTAIN Reactor Operators Authorization to Start Test signature on Cover Sheet. N/A

2. IF testing per Technical Specification SR 3.1.4.1, THEN determine which HCU test switch is to be placed first in the SRI position for as follows:

Enter Cycle number: _____

First HCU test switch in SRI: A for odd cycles B for even cycles

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21 12 of 32 SECTION 5.1.1 Initials 5.1.1 Setup N/A NOTE Documentation of performance of Steps 5.1.1.1 through 5.1.3.3, and 5.4.2 is accomplished on Attachment 3.

1. VERIFY the following Attachment 3:

- Reactor pressure is one of the following:

Testing for SR 3.1.4.3 only Any pressure Testing for SR 3.1.4.1, 3.1.4.2, or 3.1.4.4 950 psig - 1050 psig

2. VERIFY the following (Attachment 3):

- Accumulator N2 pressure > 1520 psig

- Charging Wtr Riser Valve, 1C11-EP-113 closed

- Rod to be tested at position 48 and selected 5.1.2 Test (Attachment 3)

1. On command Set from Control Room operator, place the first HCU test switch to SRI.

2. On command Mark from Control Room operator, place the second HCU test switch to SRI.

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21 13 of 32 SECTION 5.1.2 Initials N/A CAUTION If the HCU Test Switches are returned to NORMAL while the control rod is still traveling, the CRD may be damaged.

3. WHEN the Control Rod has been full-in for 10 seconds, on command Reset from Control Room operator, THEN PLACE both HCU test switches to NORMAL.

4. VERIFY HCU scram signal clear. (Flow noise stops)

NOTE In MODES 1 and 2, scram time acceptability shall be checked prior to proceeding to each successive control rod.

5.1.3 Test Completion (Attachment 3)

$ 1. VERIFY scram time acceptability as follows:

a. EVALUATE speed per one of the following:

-- IF using ICS, THEN CONFIRM an Accept on ICS printout(s) or use RSTRZZ to view the database.

-- IF using the time events analyzer, THEN CONFIRM the scram time(s) d Max_Time(s) calculated per Attachment 1 or by using the following limiting Value(s):

Reactor Pressure Limiting (psig) Notch Max Time (sec) 0 to 600, inclusive 13 0.94 600 to 950 13 1.13 950 to 1050, inclusive 43 0.30 29 0.78 13 1.40

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21 14 of 32 SECTION 5.1.3 Initials

b. IF test is performed to satisfy Technical Specification N/A SR 3.1.4.2, THEN REVIEW all slow rods to the representative sample plan and confirm no more than 7.5% of the control rods in the representative sample are slow, or EXPAND the sample size until either this 7.5% criterion is satisfied or the total number of slow control rods (throughout the core, from all surveillances) exceeds the LCO limit. Technical Specification BASES SR 3.1.4.2

c. IF the rod fails to scram on the first attempt or if the scram time to position 13 is >7 seconds, THEN immediately declare it INOPERABLE, FULLY INSERT it, and comply with the applicable Technical Specification CONDITIONS and REQUIRED ACTION statements.

d. IF the rod is slow on the first attempt, THEN IMMEDIATELY DECLARE it as such and comply with applicable Technical Specification CONDITIONS and REQUIRED ACTION statements.

2. RESTORE each HCU as follows:

a. VERIFY the pilot air solenoids re-energize, using a multimeter across terminal strip A points 1 and 2 and terminal strip B points 1 and 2.

$ b. At Containment El. 620, OPEN Charging Wtr Riser Valve 1C11-EP-113 and VERIFY accumulator pressure t1520 psig

[Exercise Open (EO) of 1C11-EP-115, Accumulator Supply Check Valve, T.S. 5.5.6].

c. CONFIRM core display accumulator fault LED has cleared.

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21 15 of 32 SECTION 5.1.3 Initials N/A

3. IF performing testing in Mode 1, APPLY a continuous insert signal for 40 seconds. After flow stabilizes, RECORD the stall flow rate.

(INTENTIONALLY BLANK)

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21 16 of 32 5.2 SINGLE CONTROL ROD SCRAM TIME TEST (Mode 3, 4 or 5) Initials

1. OBTAIN Reactor Operators Authorization to Start Test signature on Cover Sheet. RXE NOTE When performing during the RPV Leak Test during a refueling outage, use the next cycle number in Step 2. Otherwise use the current cycle number.

2. IF testing per Technical Specification SR 3.1.4.1, THEN determine which HCU test switch is to be placed first in the SRI position for as follows:

Enter Cycle number: 17 _____

First HCU test switch in SRI: A for odd cycles B for even cycles RXE NOTES x The drive water header pressure control valve is paralleled with an inline relief valve. Manual adjustment of drive water pressure must be below the relief valve setting (approx. 550 psig).

x Increasing drive water pressure will increase control rod speed; however, since all rod movements per this section are continuous rod withdrawals to position 48 under the one-rod-out interlock, control rod timing is NOT a concern.

3. RAISE drive water pressure to 400 psid. RXE

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21 17 of 32 5.2.1 Setup (Containment Team 1)

NOTES x Documentation of performance of Steps 5.2.1.1 through 5.2.3.2 and 5.4.2 is accomplished on Attachment 4.

x Subsections 5.2.1, 5.2.2 and 5.2.3 may be performed simultaneously.

1. CLOSE Charging Wtr Riser Valve, 1C11-EP-113.

2. CONFIRM Accumulator N2 pressure > 1520 psig and steady.

3. PLACE the first HCU test switch to SRI.

4. REPORT to the Control Room HCU Setup.

5.2.2 Test (Control Room Team 1 and Containment Team 2)

1. VERIFY Reactor pressure is one of the following: (Attachment 4):

Testing for SR 3.1.4.3 only Any pressure Testing for SR 3.1.4.1, 3.1.4.2, or 3.1.4.4 950 psig - 1050 psig

2. SELECT the control rod and WITHDRAW to position 48.

3. WHEN the control rod is withdrawn from the fully inserted position, OBSERVE the associated full-in indicator is NOT lit.

(SR 3.9.4.1), (LCO 3.10.3.b), (LCO 3.10.4.b.1)

4. On command Mark from Control Room operator, place the second HCU test switch to SRI.

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21 18 of 32 SECTION 5.2.2 Initials CAUTION If the HCU Test Switches are returned to NORMAL while the control rod is still traveling, the CRD may be damaged.

5. WHEN the Control Rod has been full-in for 10 seconds, on command Reset from Control Room operator, THEN PLACE both HCU test switches to NORMAL.

6. VERIFY HCU scram signal clear. (Flow noise stops)

(INTENTIONALLY BLANK)

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21 19 of 32 SECTION 5.2.3 Initials 5.2.3 Test Completion (Control Room Team 2 and Containment Team 3)

NOTES x Section 5.2.3, Test Completion, may be deferred or performed in parallel with testing of other HCUs.

x Step 5.2.3.1 and 5.2.3.2 may be performed concurrently or in any order.

$ 1. VERIFY scram time acceptability as follows:

a. EVALUATE speed per one of the following:

-- IF using ICS, THEN CONFIRM an Accept on ICS printout(s) or use RSTRZZ to view the database.

-- IF using the time events analyzer, THEN CONFIRM the scram time(s) d Max_Time(s) calculated per Attachment 1 or by using the following limiting Value(s):

Reactor Pressure Limiting (psig) Notch Max Time (sec) 0 to 600, inclusive 13 0.94 600 to 950 13 1.13 950 to 1050, inclusive 43 0.30 29 0.78 13 1.40

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21 20 of 32 SECTION 5.2.3 Initials

b. IF test is performed to satisfy Technical Specification SR 3.1.4.2, THEN REVIEW all slow rods to the representative sample plan and confirm no more than 7.5% of the control rods in the representative sample are slow, or EXPAND the sample size until either this 7.5% criterion is satisfied or the total number of slow control rods (throughout the core, from all surveillances) exceeds the LCO limit. Technical Specification BASES SR 3.1.4.2

c. IF the rod fails to scram on the first attempt or if the scram time to position 13 is >7 seconds, THEN immediately declare it INOPERABLE, FULLY INSERT it, and comply with the applicable Technical Specification CONDITIONS and REQUIRED ACTION statements.

d. IF the rod is slow on the first attempt, THEN IMMEDIATELY DECLARE it as such and comply with applicable Technical Specification CONDITIONS and REQUIRED ACTION statements.

2. RESTORE each HCU as follows:

$ a. At Containment El. 620, OPEN Charging Wtr Riser Valve 1C11-EP-113 and VERIFY accumulator pressure t1520 psig

[Exercise Open (EO) of 1C11-EP-115, Accumulator Supply Check Valve, T.S. 5.5.6].

b. CONFIRM core display accumulator fault LED has cleared.

Procedure Number:

PERRY NUCLEAR POWER PLANT SVI-C11-T1006

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21 21 of 32 5.3 FULL CORE SCRAM TIME TEST Initials

1. NOTIFY Reactor Operator prior to starting the surveillance and WRITE R.O. (print operators name) notified in the Reactor Operators Authorization to Start Test signature on the Cover Sheet. RECORD time, date and initial.

2. PERFORM scram timing analysis using ICS.

$ 3. OBTAIN ICS printout and VERIFY scram times are acceptable.

For all rods at position 48, CONFIRM Accept on ICS printout.)

5.4 Plant/System Restoration

1. ADJUST control rod drive pressure to the normal band per SOI-C11(HCU).

2. COMPLETE independent verification of the required components.

3. Test Director, VERIFY all control rods are at the correct positions as identified by the current approved Control Rod Movement or Special Maneuver Sheets, as applicable.

4. IF individual rod scrams were performed in MODE 3, 4 or 5, THEN PERFORM SVI-C71-T0051.

5. IF directed by the Unit Supervisor, THEN BYPASS the Reactor MODE Switch scram and PLACE the Reactor MODE Switch in SHUTDOWN, in accordance with SOI-C71, RPS Power Supply Distribution.

6. VERIFY the LCO Tracking Module is updated in accordance with the results of this surveillance. The search should include SLOW control rods and temperature affected CRDs.

/

Time Date US Signature

Procedure Number:

PERRY NUCLEAR POWER PLANT SVI-C11-T1006

Title:

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21 22 of 32 5.5 Acceptance Criteria Initials NOTE Satisfactory completion of this surveillance will be based on Technical Specification items (marked with $ signs).

$ 1. The scram times of each rod tested were less than or equal to the limits of Technical Specification 3.1.4.

YES NO, Unit Supervisor notified

-- IF the scram times for any control rod tested per Technical Specification SURVEILLANCE REQUIREMENT SR 3.1.4.3 fail to satisfy the scram time limits, THEN that control rod can not be declared OPERABLE.

$ 2. The Exercise Open (EO), Stroke Time (ST), and Fail Safe (FS) testing requirements for Technical Specification 5.5.6 for 1C11-EP-126 and 1C11-EP-127, and the open verification testing for Technical Specification 5.5.6 for 1C11-EP-114 for each rod tested was satisfied by the completion of Step 5.1.3.1 or 5.2.3.1.

-- IF full core scram time testing was performed per Section 5.3 of this SVI, THEN credit can be taken for Step 5.3.2.

YES NO, Unit Supervisor notified

$ 3. The Exercise Open (EO) testing requirement for Technical Specification 5.5.6 for 1C11-EP-115 for each rod tested was satisfied by recharging the accumulator after the scram per Step 5.1.3.2.b or Step 5.2.3.2.a.

-- IF full core scram time testing was performed per Section 5.3 of this SVI, THEN credit can be taken for Step 5.3.3.

YES NO, Unit Supervisor notified

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21 23 of 32 SECTION 5.5 Initials

4. All other items checked in this surveillance performed satisfactorily.

YES NO, Unit Supervisor notified

5. CHECK appropriate block on Cover Sheet to indicate acceptable or unacceptable test results.

Comments:

Performed By: / /

/ /

NOTE Additional SVI performers signature blocks are located on Attachment 2 of this SVI. Use as many copies as necessary.

6.0 RECORDS 6.1 Records Handling Records completed/generated by this document shall be handled in accordance with the established records management program.

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21 24 of 32 6.2 Records Capture The following documents are completed/generated by this instruction:

Quality Records Cover Sheet SVI-C11-T1006, pages 8 through 23, and:

Attachment 1, Calculation of Max_Time(s) for Single Rod Scrams, if used Attachment 2, Additional SVI Performers Signature Blocks, if used Attachment 3, Scram Timing Data Sheet (Section 5.1), if used Attachment 4, Scram Timing Data Sheet (Section 5.2), if used ICS Printouts, if used Time Events Analyzer Printouts, if used Non-Quality Records None

7.0 REFERENCES

7.1 Discretionary GE 762E268D D-302-871 D-302-872 B-208-040, sht. 10 GEK-75598B GEK-75612A, December 1983, Reactor Protection System, CEI VPF 14-G GE Design Specification 22A4622 FTI-B0002, Control Rod Movements

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21 25 of 32 ISI-B21-T1300-1, Reactor Coolant System Leakage Pressure Test SOI-C11 (CRDH), Control Rod Drive System SOI-C11 (HCU), Control Rod Hydraulic Control Units SOI-C11 (RCIS), Rod Control and Information System SOI-C71, RPS Power Supply Distribution SVI-C71-T0427, Reactor MODE Switch Refuel MODE Channel Functional PAP-1101, Inservice Testing of Pumps and Valves TAI-1101-2, Inservice Testing of ASME Section XI Valves 7.2 Obligations Technical Specifications LCO 3.1.4 SURVEILLANCE REQUIREMENT SR 3.1.4.1 SURVEILLANCE REQUIREMENT SR 3.1.4.2 and BASES SR 3.1.4.2 SURVEILLANCE REQUIREMENT SR 3.1.4.3 SURVEILLANCE REQUIREMENT SR 3.1.4.4 SURVEILLANCE REQUIREMENT SR 3.1.3.3 REQUIRED ACTION 3.1.5.A.1 NOTE REQUIRED ACTION 3.1.5.B.1 NOTE Technical Specification 5.5.6, Inservice Testing Program Inservice Testing Program (ISTP) Valve Relief VR-1 NRC Letter to FENOC, May 19, 2011,

Subject:

Perry Nuclear Power Plant, Unit 1 - Issuance of Amendment re: License Amendment to Modify Technical Specification 3.1.4, Control Rod Scram Times, To Incorporate Technical Specification Task Force (TSTF) Change Traveler TSTF-460, Revision 0 (TAC No. ME5194) (License Amendment 156)

NRC Letter to FENOC, May 19, 2011,

Subject:

Perry Nuclear Power Plant, Unit 1 - Issuance of Amendment re: License Amendment to Modify Technical Specification 3.1.4, Control Rod Scram Times, To Incorporate Technical Specification Task Force (TSTF) Change Traveler TSTF-222 (TAC No. ME5192) (License Amendment 157)

Procedure Number:

PERRY NUCLEAR POWER PLANT SVI-C11-T1006

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21 26 of 32 NRC Letter to FENOC, February 22, 2012,

Subject:

Perry Nuclear Power Plant, Unit No. 1, RE: Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Request VR-1, Revision 1, for the Third 10-Year Interval Inservice Testing Program (TAC No. ME7380)

USAR 3.9.4, Control Rod Drive System 4.3.2.5.5, Control Rod Patterns and Reactivity Worths 4.6.1.1, Information for CRDS 4.6.2.3.2.1, Safety Evaluation 4.6.3.1.1.3, Testing and Verification of the CRDS 4.6.3.1.1.5, Testing and Verification of the CRDS 5.2.2, Overpressurization Protection 5.2.4, Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 15.4.1, Rod Withdrawal Error - Low Power 15.4.9, Control Rod Drop Accident (CRDA)

CR 94-072 CR 93-066 CR 91-211 CR 91-161 CR 90-292 CR 89-404 CR 89-301 CR 02-00121 CR 02-01722 LER 89-030 INPO SER 9-90 SIL 575

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21 27 of 32 SIL 509 SIL 441 CRCA 03-03530-002 CRCA 04-00042-001 CR 04-02840 CR 04-03002 CR 2011-97563 LER 05000461/89-016, Clinton Power Station, Inadequate Test Procedure Results in Group 1 Containment Isolation During Restoration from Control Rod Scram Time Testing with the Reactor Pressure Vessel Solid Letter David T. Shen, General Electric, to C.S. Orogvany/P.W. Bordley, CEI, Zero Reactor Pressure Control Rod Scram Time for Perry, dated January 12, 1994 Calculation B13-1, Revision 0, Hydrostatic/Scram Time Testing Procedure Pressure Increase and Design Change Control DCC-001 written for B13-01 R0 Commitments addressed in this document:

None 8.0 SCOPE OF REVISION Rev. 21 1. Add Step 5.4.6 to verify operations have reviewed the results and created/updated PLCOs as necessary. (601053207, 601053227)

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21 28 of 32 9.0 ATTACHMENTS Attachment 1 - Calculation of Max_Time(s) for Single Rod Scrams Attachment 2 - Additional SVI Performers Signature Blocks Attachment 3 - Scram Timing Data Sheet (Section 5.1)

Attachment 4 - Scram Timing Data Sheet (Section 5.2)

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21 29 of 32 ATTACHMENT 1: CALCULATION OF MAX_TIME(S) FOR SINGLE ROD SCRAMS Page 1 of 1 Initials

1. Rod ____

2. Determine the maximum time to notch positions, for the pressure (psig) recorded in Step 4.0.2:

-- If pressure (psig) is between 0 and 600, inclusive, calculate the maximum time to notch 13, as follows, otherwise mark the following equation N/A:

§ psig 0*

Max_Time 13 0.94 ¨ ¸

0.19 ________

-- If pressure (psig) is between 600 and 950, calculate the maximum time to notch 13, as follows, otherwise mark the following equation N/A:

§ psig 600

Max_Time 13 1.13 ¨ ¸

0.27 ________

-- If pressure (psig) is between 950 and 1050, inclusive, calculate the maximum time to notches 43, 29, and 13, as follows, otherwise mark the following equations N/A:

§ psig 950

Max_Time 43 0.30 ¨ ¸

0.01 ________

§ psig 950

Max_Time 29 0.78 ¨ ¸

0.06 ________

§ psig 950

Max_Time 13 1.40 ¨ ¸

0.13 ________

Procedure Number:

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21 30 of 32 ATTACHMENT 2: ADDITIONAL SVI PERFORMERS SIGNATURE BLOCKS Page 1 of 1 Print Name Signature Initials Date

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21 31 of 32 ATTACHMENT 3: SCRAM TIMING DATA SHEET (SECTION 5.1)

Page 1 of 1 First SRI Switch A B NA Date (5.1.2)

Setup Test Test Completion Independent Verif Rod Reactor N2 Pres. EP-113 Rod Rod Both Test $ $ Solenoids $ Accum Stall Both Test EP-113 Press t 1520 CLOSED Selected Scrammed Switches In Scram Rod Energized EP-113 Fault Flow Switches OPEN 5.1.1.1 psig 5.1.1.2 and at 5.1.2.2 NORMAL Times Slow 5.1.3.2 OPEN & Clear 5.1.3.3 In 5.4.2 5.1.1.2 Pos. 48 5.1.2.3 Acceptable (Y / N) t 1520 psig 5.1.3.2 NORMAL (See 5.1.1.2 5.1.3.1 5.1.3.1 5.1.3.2 5.4.2 Note)

NOTE: 950 - 1050 psig for TS 3.1.4.1, 3.1.4.2, 3.1.4.4 scram time testing

Procedure Number:

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Title:

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21 32 of 32 ATTACHMENT 4: SCRAM TIMING DATA SHEET (SECTION 5.2)

Page 1 of 1 First SRI Switch A B NA Date Today (5.2.2)

Setup Test Test Completion Independent Verification Rod EP-113 N2 First Reactor Rod Full In Rod Both Test $ $ $ Accum. Both Test EP-113 CLOSED Pres. Test Pressure Selected Ind. Scrammed Switches Scram Rod EP-113 Fault Switches In OPEN 5.2.1.1 t 1520 Switch 5.2.2.1 and at NOT lit 5.2.2.4 In Times Slow OPEN & Clear NORMAL psig in SRI Pos. 48 5.2.2.3 NORMAL Acceptable (Y / N) t 1520 5.2.3.2 5.2.1.2 5.2.1.3 (See 5.2.2.2 5.2.2.5 5.2.3.1 5.2.3.1 psig Note) 5.2.3.2 18-55 1010 x

x 18-27 1010 x

x 50-43 1010 x

x 22-47 1010 x

x 46-23 1010 x

x 14-51 1010 x

50-23 x 1010 x

x 18-51 1010 NOTE: 950 - 1050 psig for TS 3.1.4.1, 3.1.4.2, 3.1.4.4 scram time testing

ML22146A090

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