Text

Final - Lsro Section a Operating
	ML021700671
Person / Time
Site:	Peach Bottom, Limerick
Issue date:	05/24/2002
From:	Wasong A; Exelon Nuclear
To:	Conte R; NRC/RGN-I/DRS/OSB
References
	-nr, 50-277/02301, 50-278/02301, 50-352/02301, 50-353/02301
	Download: ML021700671 (139)
	v • d • e

EXELON NUCLEAR TITLE: Working Hour Limitations TASK PERFORMED BY:

EVALUATOR SIGNATURE:

DIRECTIONS TO EVALUATOR:

EVALUATOR:

DATE:

EVALUATION METHOD:

PERFORM EVALUATION LOCATION:

ANY APPROXIMATE COMPLETION TIME:

Completion Time: 10 minutes IMPORTANCE RATING(S):

3 3.4 SYSTEM NUMBER(S):

2.

1.5 REFERENCES

1. A-C-40 Rev 1, WORKING HOUR LIMITATIONS (Common)

TASK STANDARD(S):

RPO determined to not be able to stand watch.

FHD determined to need working hour limit deviation form to permit helping with surveillance A1-7.doc Page 3 of 8 A l-l.doc Page 3 of 8

S.

C-."

1'

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EXELON NUCLEAR TASK CONDITIONS:

1. The following Limerick Unit I outage working hour history is given for an on coming Fuel Handling Director (FHD) and a Refuel Platform Operator (RPO). The hours worked are on the Unit 2 Refueling Platform performing core alterations L _______________

I _________________

DS=Dayshift NS=Nightshift The RPO and FHD are scheduled to work dayshift today, 6/6 INITIATING CUES:

You are directed to evaluate the working hour history for the FHD and the RPO and determine whether one or both individuals can work a full dayshift of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A1-7AYOc Page 4 of 8 Date FHD RPO 5/28 (Day 1) 12 hrs (DS) 12 hrs (NS) 5/29 12 hrs (DS)

OFF 5/30 12 hrs (DS)

OFF 5/31 OFF 12 hrs (DS) 6/1 12 hrs (DS) 12 hrs (DS) 6/2 12 hrs (DS) 12 hrs (DS) 6/3 12 hrs (DS) 12 hrs (DS) 6/4 12 hrs (DS) 12 hrs (DS) 6/5 12 hrs (DS) 12 hrs (DS)

Page 4 of 8 I

A l-f.dOC

EXELON NUCLEAR Critical Element(s) indicated by

in Performance Checklist.

PERFORMANCE CHECKLIST:

STEP STANDARD F SAT/UNSAT Assess working hour limitations in A-C-40 Note: The JPM steps below are excerpts from A-C-40 that represent the most likely steps the candidate will cite when performing the JPM. The entire procedure is not duplicated.

Fuel Handling comes under the control of Nuclear Safety Related work under the definitions section 4 2 2

1. Obtain A-C-40 A-C-40 obtained

2. The number of hours worked shall be Assess hours worked and conclude the controlled in accordance with the following:

limitations set in Working Hour Limitation The FHD may work the entire 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Deviation Form (Exhibit A-C-40-1) dayshift

3. *The number of hours worked shall be Assess hours worked and conclude the controlled in accordance with the following:

limitations set in Working Hour Limitation The RPO may not take the shift because Deviation Form (Exhibit A-C-40-1) the next hour worked will result in the RPO exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 168 hour0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months period

4. CUE: Assume you have completed a 12 N/A hour shift as FHD and are in the NMD office. You have just received a call one hour after turnover from the refuel bridge that your help is needed in performing a partial ST on the bridge for about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Assess whether you can assist with the ST AI-1.doc Page 5 of 8 A l-l. doc Page 5 of 8

AI-1.doc Page 6 of 8 i

i I ________________________

EXELON NUCLEAR STEP STANDARD SAT/UNSAT

5. *The number of hours worked shall be Conclude that the ST work may NOT be controlled in accordance with the performed due to exceeding working limitations set in Working Hour Limitation hour limits of Deviation Form (Exhibit A-C-40-1) 24 in 48 hrs, and 0

72 in 168 hrs OR Yes, the work may be performed provided a Working Hour Limits Deviation is processed and approved.

6. CUE: You have reached the termination N/A point for this JPM. You may stop here.

"A l-l.doc Page 6 of 8

EXELON NUCLEAR Comments:

Note: Any grade of UNSAT requires a comment.

JPM Overall Rating:

SAT/UNSAT A I -1.doc Page 7 of 8 A l-l.doc Page 7 of 8

EXELON NUCLEAR TASK CONDITIONS:

1. The following Limerick Unit I outage working hour history is given for an on-coming Fuel Handling Director (FHD) and a Refuel Platform Operator (RPO). The hours worked are on the Unit 2 Refueling Platform performing core alterations I

I DS=Dayshift NS=Nightshift The RPO and FHD are scheduled to work dayshift today, 6/6 INITIATING CUES:

You are directed to evaluate the working hour history for the FHD and the RPO and determine whether one or both individuals can work a full dayshift of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Date FHD RPO 5/28 (Day 1) 12 hrs (DS) 12 hrs (NS) 5/29 12 hrs (DS)

OFF 5/30 12 hrs (DS)

OFF 5/31 OFF 12 hrs (DS) 6/1 12 hrs (DS) 12 hrs (DS) 6/2 12 hrs (DS) 12 hrs (DS) 6/3 12 hrs (DS) 12 hrs (DS) 6/4 12 hrs (DS) 12 hrs (DS) 6/5 l2hrs(DS) 12 hrs (DS)

CANDIDATE I

Effective Date:

1/18/96 Transition Concurrence A.

S.

MacAinsh

/

Date A-C-40, Rev.

1 Page 1 of 6 RMR/rmr PECO Energy Company Nuclear Generation Group WORKING HOUR LIMITATIONS (This Revision is a Complete Rewrite)

PURPOSE To establish the administrative controls to limit, approve and review the working hours of personnel performing Nuclear Safety Related (NSR) functions at Peach Bottom Atomic Power Station (PBAPS)/Limerick Generating Station (LGS) in excess of working hour limits.

(3.1.1 and 3.1.2)

To reduce the probability of personnel error potentially caused by personnel performing work in a fatigued condition.

SCOPE Working hour limitations apply to personnel performing NSR functions at PBAPS/LGS.

Requirement to notify supervision if working in a fatigued condition applies to Nuclear Generation Group (NGG) personnel.

This procedure supersedes NQA-40.

SOURCES AND REFERENCES SOURCE DOCUMENTS

1.

Limerick Generating Station Technical Specifications Section 6.2.2.f

2.

Peach Bottom Atomic Power Station Technical Specifications Section 5.2.2.e

3.

NUREG-0737, Clarification of TMI Action Plan Requirements, Item I.A.I.3, Shift Manning and LGS UFSAR Section 1.13.2 PORC YES SQR YES QR YES 50.59 YES RESP.

MGR.

YES 1.0 1.1 1.2 2.0 2.1 2.2 2.2 3.0 3.1

A-C-40, Rev.

1 Page 2 of 6

4.

NRC Safety Evaluation Report dated August 30,

1995, "Issuance of Improved Technical Specifications, Peach Bottom Atomic Power Station, Units 2 and 3"

5.

NUREG-0991, Supplement 1, Section 13.1.2, NRC Safety Evaluation Report Related to the Operation of Limerick Generating Station

6.

NRC Generic Letter 82-12, Nuclear Power Plant Staff Working Hours

7.

NRC Generic Letter 83-14, Definition of 'Key Maintenance Personnel' (Clarification of Generic Letter 82-12)

8.

NRC NUREG/CR-5569, HPPOS-173, Applicability of Generic Letter 82-12 to Radiation Protection Staff

9.

QA Finding AL87-128-01

10.

QA Recommendation Q0000095

11.

PEP Recommendation 10001451 3.2 CROSS REFERENCE None 4.0 DEFINITIONS 4.1 PERSONNEL PERFORMING NSR FUNCTIONS Personnel performing or immediately supervising the performance of NSR functions at PBAPS/LGS.

This includes, but is not limited to, licensed senior operators, licensed operators, non-licensed operators, radiation protection personnel, chemistry personnel, key maintenance personnel, contractors (provided they are not covered by an equivalent procedure) and licensed limited senior reactor operators who perform NSR functions.

(3.1.7, 3.1.8 and 3.1.9) 4.2 NUCLEAR SAFETY RELATED FUNCTIONS 4.2.1 Functions physically performed on equipment identified as NSR on the Component Record List (CRL) by the presence of a "Q",

"P" or "A" under QA Class.

A-C-40, Rev.

1 Page 3 of 6 4.2.2 These functions include:

physical manipulation of NSR equipment; troubleshooting; performance of surveillance tests; quality verifications; testing; calibration; repair/replacement of NSR systems, structures and components (SSC's); installation of a NSR modification where actual physical (hands on) interaction with NSR SSC's occurs.

This also includes radiation protection and chemistry personnel who are assigned certain emergency response duties including assignment to in-plant rescue teams, environmental monitoring and dose calculations, or who handle, process or provide data and input to emergency response decision makers.

In addition, these functions include reactor engineers supporting reactivity manipulations during critical reactor operations.

(3.1.7, 3.1.8 and 3.1.9) 4.2.3 These functions exclude but are not limited to the following examples:

document, procedure or instruction preparation; corporate and local area network (LAN) computer input and performance of non-emergency radiation surveys.

4.3 WORKING HOURS - All hours that are worked, exclusive of shift turnover time.

5.0 RESPONSIBILITY AND AUTHORITY 5.1 ALL PERSONNEL - Notify supervision if working in a fatigued condition regardless of the type of work being performed.

5.2 PERSONNEL PERFORMING WORK AT LGS/PBAPS

1.

Monitor their hours.

2.

Notify supervision prior to exceeding any working hour limitations.

5.3 SUPERVISION

1.

Assigns work in a manner which minimizes the need to deviate from working hour limitations and avoids allowing workers to perform work in a fatigued condition.

2.

Processes documentation for obtaining approval to deviate from working hour limitations.

3.

For workers performing NSR functions, obtains verbal or written management approval prior to the worker exceeding working hour limitations.

4.

Notifies worker of approval/disapproval to exceed working hour limitations.

A-C-40, Rev.

1 Page 4 of 6 5.4 EMPLOYING OFFICER

(

1.

Ensures adequate staffing levels to preclude the routine use of overtime.

2.

Approves departure from applicable working hour limitations.

5.5 PLANT MANAGER OR APPOINTED INDIVIDUALS

1.

Reviews individual overtime monthly.

(3.1.1 and 3.1.2)

2.

Retains documentation for one year.

5.6

MANAGER, SENIOR MANAGER, OR PLANT MANAGER

1.

Approves departure from applicable working hour limitations.

6.0 PREREQUISITES None 7.0 PROCEDURE 7.1 GENERAL 7.1.1 NGG personnel shall notify supervision if fatigued while working regardless of the type of work being performed.

7.1.2 Routine deviation from working hour limitations shall not be utilized.

7.1.3 Workers shall monitor their hours using their work center timekeeping records or by personal tracking methods.

7.1.4 Workers shall notify supervision prior to exceeding working hour limitations.

7.1.5 Supervisors should schedule and assign work in a manner which minimizes the need to deviate from the appropriate working hour limitations and avoid allowing workers to perform work in a fatigued condition.

7.1.6 Employing Officer ensures adequate staffing levels to preclude routine heavy use of overtime.

7.2 LIMITATIONS ON HOURS WORKED 7.2.1 The number of hours worked shall be controlled in accordance with the limitations set in Working Hours Limitation Deviation Form (Exhibit A-C-40-1).

A-C-40, Rev 1 Page 5 of 6 7.2.2 Use of the form or other means of documentation is required for documenting working hour deviations for NSR functions.

For groups who use a time system to monitor working hours, other means of documentation such as a time slip may be used in lieu of the form provided the other means of documentation contains the same information as the form.

7.2.3 When recommending personnel for exceedance of working hour limitations, the Supervisor shall consider the impact of previously worked NSR and non-NSR hours in the analysis as to whether personnel can perform additional NSR functions safely.

7.2.4 Use of the form is not required for documenting working hours exceeding the limitations of this procedure when those hours exceeding the limitations are limited to non-NSR functions.

7.3 DEVIATIONS FROM WORKING HOURS LIMITATIONS 7.3.1 Except during extended shutdown periods, working hour deviations shall be considered on an individual basis and not for the entire staff on a shift.

(3.1.1 and 3.1.2) 7.3.2 During extended shutdown periods (e.g., refueling outages),

working hour deviations may be approved for more than one individual at a time (e.g., an entire work team).

Each deviation approval shall list the affected individuals, the description and duration of the approved deviation.

7.3.3 Supervision examines alternatives prior to requesting management approval to exceed the working hour limitations and processes the form.

7.3.4 When a worker will be performing NSR functions, supervision shall obtain verbal or written management approval prior to the worker exceeding the working hour limitations.

1.

When written approval is obtained by supervision, the form shall be completed instantaneously with the approval.

2.

When verbal approval is obtained by supervision, the form should be completed by supervision by the end of the next working day.

Supervision indicates on the form or other means of documentation the time and date approval was obtained and the name of the approver.

(3.1.11) 7.3.5 Supervision shall obtain verbal or written management approval as follows:

Employing Officer, Plant Manager, Senior Manager or the Manager of the individual(s)

(this may not be delegated lower in an organization).

(3.1.1 and 3.1.2) 7.3.6 Supervision notifies worker of approval/disapproval prior to exceeding the working hour limitations.

A-C-40, Rev.

1 Page 6 of 6 7.3.7 Individual overtime shall be reviewed monthly by the Plant Manager or appointed individuals to assure that excessive working hours have not been assigned.

(3.1.1 and 3.1.2) 8.0 DOCUMENTATION 8.1 The deviation forms or other means of documentation and the monthly review documentation shall be retained for one year by the Plant Manager or appointed individuals.

9.0 EXHIBITS 9.1 A-C-40-1, Working Hours Limitation Deviation Form

PORC YES SQR YES QR YES 50.59 YES RESP.

MGR.

YES Exhibit A-C-40-1, Rev.

2 Page 1 of 1 JMP:jmp WORKING HOURS LIMITATION DEVIATION FORM

1.

Are the functions described Nuclear Safety Related?

YES NO*

2.

Name(s)

Payroll No.

Job Classification Place additional names and other information on back.

(3.1.10)

3.

Limit to be exceeded excluding shift turnover time (3.1.1 3.1.2):

Work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight.

Work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in a 24-hour period.

Work more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48-hour period.

Work more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 168 hour0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months period.

Break of less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months between work periods including shift turnover time.

4.

Time Period to Work:

From:

Date/Time To:

Date/Time

5.

Describe work to be performed while in deviation:

6.

Justification for exceeding the limit (why is it necessary?):

(3.1.1 and 3.1.2)

7.

Recommended by:

/

Supervisor 8.*

Approved by:

See Paragraph 7.3.5 for approvers D

Date

(*)

Completion not required when documenting non-NSR activities.

(SEND ORIGINAL TO PLANT MANAGER OR APPOINTED INDIVIDUALS)

Date

EXELON NUCLEAR' TITLE: Fuel Handlingi Director Turnover Conditions Plant Differences TASK PERFORMED BY:

EVALUATOR SIGNATURE:

DIRECTIONS TO EVALUATOR:

EVALUATION METHOD:

PERFORM EVALUATION LOCATION:

ANY APPROXIMATE COMPLETION TIME:

Completion Time: 10 minutes IMPORTANCE RATING(S):

3 3.5 SYSTEM NUMBER(S):

2.2.28 Generic

REFERENCES:

1. Exhibit NOM-L-4.1:13, Rev 0, FUEL HANDLING DIRECTOR SHIFT TURNOVER CHECKLIST TASK STANDARD(S):

Turnover conditions assessed with consideration of the different administrative limits. Determine that core alterations may proceed at PBAPS and may NOT proceed at LGS.

A 7-it. UOC Page 3 of 7 EVALUATOR:

DATE:

I A- -. doc Page 3 of 7

I EXELON NUCLEAR TASK CONDITIONS:

The following plant conditions for Limerick Unit 2 and Peach Bottom Unit 2 are provided INITIATING CUES:

Evaluate turnover conditions and determine whether core alterations are permitted and the basis for any discrepancy on Both Units A 1-id. cloc Page 4 of 7 Plant Condition Limerick Unit 2 Peach Bottom Unit 2 Plant Operating Condition OPCON 5 Mode 5 Activities Shuffle Part 2 to start Shuffle Part 2 to Start Core Status Per the Attached Map Per the Attached Map Recirc Pumps A and B Secured A and B Secured RHR "A" SDC 6200 gpm "A" SDC 6200 gpm Coolant Temp 110 degrees F 110 degrees F RPV Water Level 483" LR42-1 R608 460 " PR/FR-2-06-097 Fuel Pool Gates Removed Removed Control Rods All inserted All inserted 9

9*4

9 9 A7I-2.cdOC Page 4 of 7

EXELON NUCLEAR Critical Element(s) indicated by in Performance Checklist.

PERFORMANCE CHECKLIST STEP STANDARD SAT/UNSAT

1. Obtain Exhibit NOM-L-4.1:13 Exhibit NOM-L-4.1:13 obtained

2. Assess plant conditions against requirements

a. *RHR SDC flow <6000 gpm at Determine RHR flow is too high at LGS or < 6500 gpm at PBAPS Limerick. Determine from the map that or all in-core instruments numerous instruments are not supported supported, and the flow limits apply Conclude that co re alterations may not commence at Limerick

b. RHR SDC flow <6000 gpm at All conditions are satisfactory at Peach LGS or < 6500 gpm at PBAPS Bottom or all in-core instruments Conclude that core alte.rations my supported commence at Pach Bottom CUE: You have reached the termination point N/A N/A for this JPM. You may stop here.

A I -Z. (70C Page 5 of 7

EXELON NUCLEAR

~*~~

Comments:

Note: Any grade of UNSAT requires a comment.

JPM Overall Rating:

SAT/UNSAT AI-2.doc Page 6 of 7 A1-2.doc Page 6 of 7.

TASK CONDITIONS:

The following plant conditions for Limerick Unit 2 and Peach Bottom Unit 2 are provided INITIATING CUES:

Evaluate turnover conditions and determine whether core alterations are permitted and the basis for any discrepancy on Both Units Plant Condition Limerick Unit 2 Peach Bottom Unit 2 Plant Operating Condition OPCON 5 Mode 5 Activities Shuffle Part 2 to start Shuffle Part 2 to Start Core Status Per the Attached Map Per the Attached Map Recirc Pumps A and B Secured A and B Secured RHR "A" SDC 6200 gpm

'A" SDC 6200 gpm Coolant Temp 110 degrees F 110 degrees F RPV Water Level 483 " LR42-1 R608 460 " PR/FR-2-06-097 Fuel Pool Gates Removed Removed Control Rods All inserted All inserted CANDIDATE EXELON NUCLEAR..

QJ DCN

~0 oE S.

r-3/4

-c E

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C 0

ExeOP-AA-112-101 Rxl mm Revision 0 Page 1 of 12 Nuclear Level 3 - Information Use SHIFT TURNOVER AND RELIEF

1.

PURPOSE 1.1.

This procedure provides Operations shift turnover and relief guidance to all Exelon Nuclear generating stations.

2.

TERMS AND DEFINITIONS 2.1.

Shift relief: occurs when the off-going and on-coming crewmembers are satisfied a proper turnover has been performed. The off-going crew member is not relieved until the on-coming crew member has accepted the turnover sheet or checklist signifying their acceptance of the duties.

3.

RESPONSIBILITIES 3.1.

All shift personnel are responsible for reviewing and understanding the logs and checklists applicable to their shift position before assuming the shift.

3.2.

The off-going operator shall not leave his / her work area until he / she is satisfied that his / her relief is fully aware of existing conditions.

3.3.

Each individual is personally responsible for ensuring their "Active Status" is in good standing prior to assuming duty. This status includes proper number of shifts / hours worked per quarter, training qualifications, and up-to-date medical status.

3.4.

No person shall assume a shift position unless they are physically and mentally fit to competently assume the responsibilities.

3.5.

No person shall permit their relief to assume the shift if any doubt exists concerning the on-coming persons alertness, coherence, and capability of performing their assigned duties.

3.6.

The off-going operator shall not be relieved until the equipment he / she is responsible for is stable or the condition has been discussed and understood by the on-coming operator.

3.7.

A shift briefing shall be conducted by the Shift Manager or his designee for the operating shift to discuss items such as planned evolutions, pertinent daily orders, unusual conditions and general shift routine. The shift briefing may be made up of more than one meeting as deemed necessary by the Shift Manager.

OP-AA-1 12-101 Revision 0 Page 2 of 12 3.8.

The STA-qualified Supervisor / Independent Assessor (IA) shall remain within 10 minutes of the Control Room at all times.

3.9.

The STA-qualified Supervisor I IA shall remain cognizant of plant status during normal operations.

4.

MAIN BODY 4.1.

General Guidelines For Shift Turnover And Relief:

NOTE:

These turnover requirements apply only to positions that are staffed.

4.1.1.

CONDUCT shift relief and turnover in a professional, disciplined, and formal manner.

All turnovers should follow a similar protocol. The turnovers transfer specific legal responsibilities to qualified individuals.

4.1.2.

PERFORM control room shift turnovers without interfering with or distracting the operator from important evolutions requiring continued operator attention, such as the approach to criticality, rapid down-power, significant plant evolution, transient response, etc.

4.1.3.

VERIFY important operating parameters, especially those relating to safety systems, as identified on the turnover sheet prior to assuming the shift. Each licensed person must be cognizant at all times of the position of the other shift licensed Reactor Operators, Supervisors, Shift Technical Advisor, and Fire Brigade Chief.

4.1.4.

VERIFY the on-coming person is properly licensed and/or qualified to assume the shift position.

NOTE:

Any turnover in progress may be temporarily suspended to allow participation in the shift turnover meeting. The turnover should be completed immediately after the shift turnover meeting.

4.1.5.

PERFORM turnovers at the normal workstation unless plant conditions dictate a change (e.g. during a EP condition, or field work in progress).

OP-AA-1 12-101 Revision 0 Page 3 of 12

1.

For non-position specific turnover of a job in progress in the field, utilize a general turnover that addresses the following attributes:

A.

Oncoming operator's name B.

Date and Shift of oncoming shift C.

Major activities in progress related to the job D.

Activities to be accomplished by the oncoming shift E.

Special notes or conditions F.

Signatures of off-going and oncoming operators acknowledging turnover

2.

If a turnover occurs in the field during performance of a task, the Shift Manager will ensure the individual(s) receive a follow-up shift turnover briefing.

4.1.6.

DOCUMENT shift turnovers on a written hard copy or computer generated facsimile or on the operator logs. The turnover documents must contain all pertinent status, items in progress or that have occurred, and other issues or concerns.

4.1.7.

CONDUCT mid-shift turnover and relief between individuals. The turnover and relief should be comprehensive enough to ensure the relief person is fully cognitive of plant status and aware of ongoing activities.

Control room watch station turnover and relief for less than one (1) hour with the individual remaining in the control room area only requires a review of the current shift turnover checklist and update of any deviations of plant status or activities from this checklist.

4.2.

Shift Turnover Sheets:

4.2.1.

FILL OUT a shift turnover sheet I checklist for each duty station. The basic turnover sheet should reflect the following areas of review:

Plant or Unit Status such as, MODE, reactor power, main generator megawatts, etc., as appropriate for the position.

Major evolutions, clearances, and surveillances in progress at tumover.

Pending activities such as surveillances and major evolutions planned for the on-coming shift.

Administrative changes such as Temporary Procedures, Temporary Modifications, Daily Orders.

4.2.2.

INCLUDE additional items on various watch station turnover sheets as required, including the following:

LCOAR / LCO status.

Temporary Procedures.

OP-AA-1 12-101 Revision 0 Page 4 of 12 Temporary Modifications.

Standing Orders.

Surveillance Schedule.

Work Planning Schedule.

Temporary Lifts.

Electric Operation/Transmission System.

Condition Reports.

Daily Orders.

Daily Manning Schedules.

Reporting Requirements (GSEP/ENS/LER-CR).

Material Condition.

Radiation Protection (RWP/Releases/Precautions/ALARA).

Infrequent Plant Activities (IPA).

EO/Division LD.

Safeguards Systems.

MCB/Indication/Controller/Alarm.

Fire Protection.

Chemistry.

SSPS / RPS Channels / Bistables.

4.3.

Shift Manager Turnover Responsibilities:

4.3.1.

PERFORM turnover in either the Shift Manager's office, the Work Execution Center, or the Main Control Room, as conditions dictate. The Shift Manager's position must be filled by a qualified licensed senior reactor operator.

4.3.2.

INITIATE a turnover sheet and ENTER the current status for each Unit (Off going

-Shift Manager). Examples of items that may be included on the turnover sheet are:

Unit status, LCOAR / LCO status, Major out-of-services, Surveillances, Major evolutions in progress or planned, Abnormal line-ups or conditions, Changes in radiological conditions, Status of radwaste items.

Releases Capacity(ies)

OP-AA-1 12-101 Revision 0 Page 5 of 12 4.3.3.

Prior to relief, the on-coming Shift Manager should PERFORM the following:

READ the Shift Manager logs through the last previous date on shift, or the preceding four days logs, whichever is less.

DISCUSS with the off-going Shift Manager all items listed on the turnover sheet and any other information pertinent to proper continuity of supervision.

REVIEW the Daily Order Book.

REVIEW Standing Orders for new entries.

4.3.4.

After relief, the on-coming Shift Manager should PERFORM the following:

CONDUCT a shift turnover meeting of Shift Personnel covering the scope of the shift activities. Shift turnover meetings shall be held at a location to minimize distraction to Main Control Room personnel.

TOUR the Main Control Room.

Visually INSPECT the control boards.

4.4.

Unit Supervisor's Turnover Responsibilities 4.4.1.

PERFORM turnover in the Main Control Room. The Unit Supervisor position must be filled by a qualified licensed senior reactor operator.

OP-AA-112-101 Revision 0 Page 6 of 12 4.4.2.

INITIATE a turnover sheet and ENTER the current status for their respective Unit.

(Off-going Unit Supervisor) Examples of items that may be included are:

Unit status, LCOAR I LCO timeclock status, Major clearances, Planned reactivity changes, Degraded equipment, Abnormal positioned components, Daily orders or new standing orders, Surveillances, Major evolutions in progress or planned.

4.4.3.

Prior to relief, the on-coming Unit Supervisor should PERFORM the following:

READ the applicable Unit logs through the last previous date on shift, or the preceding four days logs, whichever is less.

DISCUSS with the off-going Unit Supervisor all items listed on the turnover sheet and any other information pertinent to proper continuity of supervision.

DISCUSS/REVIEW contingency actions / abnormal procedures when redundant pieces of equipment are unavailable (e.g., stator cooling pump)

TOUR the Main Control Room with the off-going Unit Supervisor, Visually INSPECT I WALKDOWN the control boards, REVIEW annunciator status, REVIEW the Unit logs and Unit Reactor Operator turnover sheets.

REVIEW Standing Orders for new entries.

4.4.4.

After relief, the Unit Supervisor should PERFORM the following:

ANNOUNCE completion of shift turnover and relief to the Unit ROs.

REVIEW the Daily Orders.

4.5.

WEC Supervisor Turnover Responsibilities 4.5.1.

Normally PERFORM turnover in the Work Execution Center. The WEC Supervisor position must be filled by a qualified licensed senior reactor operator.

4.5.2.

INITIATE a turnover sheet and ENTER the current status for their respective Unit (Off-going WEC Supervisor). Examples of items that may be included are:

Unit status, LCOAR / LCO timeclock status, Major OOS, Planned reactivity changes, Major evolutions in progress or planned.

OP-AA-1 12-101 Revision 0 Page 7 of 12 4.5.3.

Prior to relief, the oncoming WEC Supervisor should PERFORM the following:

REVIEW WEC Supervisor turnover sheet DISCUSS with the off-going Supervisor all items listed on the turnover sheet and any other information pertinent to proper continuity of supervision.

4.5.4.

Following relief, the oncoming WEC Supervisor should PERFORM the following:

Review Daily Orders and Standing Orders Review Unit Supervisor turnover sheet Review RO turnover sheet 4.6.

STA Turnover Responsibilities (when STA function is required) 4.6.1.

INITIATE a turnover sheet (may be included as part of Unit Supervisor turnover sheet) and ENTER the current status for their respective Unit (Off-going STA).

4.6.2.

PERFORM turnover in the Main Control Room. The STA position must be filled by a qualified individual.

4.6.3.

Prior to relief, the oncoming STA should PERFORM the following:

DISCUSS with the off-going STA all items listed on the turnover sheet.

REVIEW current plant status, trends, and anticipated actions.

4.6.4.

After relief, the STA should PERFORM the following:

ANNOUNCE completion of shift turnover to the Unit Supervisors and Shift Manager, including how the STA can be contacted.

REVIEW the Daily Orders and Standing Orders for new entries.

REVIEW Unit Supervisor turnover sheet.

REVIEW the RO turnover sheets.

4.7.

Field Supervisor Shift Turnover Responsibilities:

4.7.1.

INITIATE a turnover sheet and ENTER the current status for their respective responsibilities (Off-going Field Supervisor). Examples of items that may be included are:

Systems status in plant.

Alarms in plant.

Ground & Test Devices installed.

Abnormal controllers in plant.

Abnormal instrumentation in plant.

Special chemistry or radwaste evolutions in progress or pending.

Releases Switchyard evolutions in progress or pending switching orders.

Radiation Precautions.

OP-AA-1 12-101 Revision 0 Page 8 of 12 Secondary/Balance of Plant (BOP) steam leaks.

Miscellaneous Secondary/BOP items.

Fire Chief Duties Fire Protection System status and impairments.

4.7.2.

CONFER with the Shift Manager concerning the scope of the shift and applicable responsibility for that shift.

4.8.

Reactor Operator Shift Turnover Responsibilities:

4.8.1.

PERFORM turnover in the Main Control Room. The Reactor Operator position must be filled by a qualified licensed reactor operator.

4.8.2.

INITIATE a turnover sheet and ENTER the current status for their respective Unit (Off-going Reactor Operator). Examples of items that may be included are:

Unit status, LCOAR / LCO status, Major Out-Of-Services, Abnormal positioned components, Daily orders, Surveillances, Major evolutions in progress or planned.

4.8.3.

Prior to relief, the on-coming Reactor Operators should PERFORM the following:

READ the Control Room logs through the last previous date on shift, or the preceding four days logs, whichever is less.

DISCUSS with the off-going Reactor Operator all items listed on the turnover sheet, Shiftly and Daily Surveillance, and any other information pertinent to proper continuity of operations.

TOUR the Main Control Boards with the off-going Reactor Operators and DISCUSS the following as a minimum:

Status of safety-related systems.

Running equipment and safety train alignments.

Inoperable equipment, including instrumentation, and Limiting Conditions of Operations, including surveillance requirements.

Contingency actions / abnormal procedures when redundant pieces of equipment are unavailable (e.g., stator cooling pump)

Reasons for new annunciator alarms.

Out-of-service equipment including any surveillance or equipment work in progress at time of the shift relief.

Abnormal events occurring during the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

TOUR Main Control Room back panel areas

OP-AA-1 12-101 Revision 0 Page 9 of 12 4.8.4.

After relief, the on-coming RO's should PERFORM the following:

ANNOUNCE shift turnover and relief to the Unit Supervisor.

CONFER with the Unit Supervisor to determine the scope of planned shift activities and their responsibilities for that shift.

REVIEW Daily Orders.

REVIEW Standing Orders for new entries.

4.9.

Radwaste Panel Operator Turnover Responsibilities 4.9.1.

PERFORM turnover at the radwaste panel as appropriate.

4.9.2.

INITIATE a turnover sheet and ENTER the current status for their respective responsibilities (Off-going Radwaste panel operator). Examples of items that may be included are:

Liquid Systems.

Gaseous Systems.

Waste Water Treatment status.

Sewage Treatment status.

Makeup Demins status.

CP Demins status.

Shipments status.

LCO / TRM / RETS I DEL status.

Need to Order status.

Gas Decay Tanks status.

Abnormal Radwaste Controllers.

Abnormal Radwaste Instrumentation.

Water Inventory status in Radwaste.

Tank Capacity status.

Radiation releases in progress or pending.

Radiation Precautions.

Volume reduction system status.

OP-AA-112-101 Revision 0 Page 10 of 12 4.9.3.

Prior to relief, the oncoming radwaste panel operator should PERFORM the following:

Read the radwaste panel log through the last date on shift or the preceding four days, whichever is less.

Discuss with the off-going radwaste panel operator all items listed on the turnover sheet and any other information pertinent to proper continuity of operations.

Tour the radwaste control panels.

4.10.

Non-Licensed Operator (NLO) Shift Turnover and Relief:

4.10.1.

PERFORM turnover at the respective "Ready Rooms", local stations, or other area approved by the SOS or Shift Manager as necessary, and DISCUSS the following subject areas and /or others as appropriate:

Abnormal systems status in plant for that position.

Alarms in plant.

Ground & Test Devices installed.

Abnormal controllers in plant.

Abnormal instrumentation in plant.

Switchyard evolutions in progress.

Unusual Radiation Precautions.

New Secondary/Balance of Plant (BOP) steam leaks.

Miscellaneous Secondary/BOP items.

Miscellaneous Safety Systems items.

Required Fire Watches or increased monitoring by Operations Out-Of-Services in progress or pending.

Major evolutions in progress of pending.

Surveillances in progress.

LCOAR / LCO status.

4.10.2.

The on-coming operator should PERFORM the following:

REVIEW the logs, noting and discussing any readings deviating from accepted ranges or trending.

DISCUSS with off-going operator any major evolutions pending or in progress.

4.10.3.

MEET with their supervisors to discuss the scope of the shift, activities, etc.

4.11.

Fuel Handling Supervisor Shift Turnover Responsibilities 4.11.1.

During fuel moves and other activities involving handling of fuel or sources:

OP-AA-1 12-101 Revision 0 Page 11 of 12

1.

INITIATE a turnover sheet and ENTER the current status for their respective responsibilities (off-going supervisor). Examples of items that may be included are:

Fuel Handling-related systems' status in plant.

Fuel Handling-related Out-Of-Services (OOS) or Return-to-Services (RTS).

Fuel Handling-related jumpers and temporary circuit changes.

New procedures, memos, tests, and special conditions.

Status of fuel handling work in progress or completed during previous shift.

4.11.2.

CONFER with the Shift manager concerning the scope of the shift and applicable responsibility for that shift.

4.12.

Shift Briefings:

4.12.1.

ATTEND the briefing:

All personnel assigned to the on-coming shift unless exempted by Shift Management.

Personnel involved in work to be accomplished during the next shift to describe what is to be accomplished, if appropriate, or ENSURE the topic is adequately covered in the pre-job brief.

Personnel as designated by the Shift Manager.

4.12.2.

CONDUCT a shift turnover meeting in a formal and professional manner in a location that minimizes distraction to Main Control Room personnel.

4.12.3.

STRESS personnel safety issues by using questions, reminders and examples during the briefing.

4.12.4.

REVIEW each Unit's status, emphasizing nuclear safety. Briefings should include, planned reactivity changes, shift priorities, evolutions in progress, and any problems with plant equipment.

4.12.5.

BRIEF the crew on plans to accomplish scheduled activities. Crew assignments and responsibilities should be reviewed.

4.12.6.

ANNOUNCE any HLA briefings associated with operations at the shift briefing.

4.12.7.

DISCUSS the following items as appropriate:

Daily Orders.

Standing Orders.

LCO timeclocks.

Contingencies.

Lessons learned - Exelon or industry OPEX.

OP-AA-1 12-101 Revision 0 Page 12 of 12 Focus area of the week.

4.12.8.

SOLICIT input from personnel assigned to specific shift positions and PROVIDE an opportunity for other participants to provide input or ask questions.

Each NLO report status of watch station.

Each RO report status of Unit.

Field Supervisor report status of plant systems.

Each Unit Supervisor report status of Unit.

WEC Supervisor report on upcoming activities.

Shift Manager report status of Station.

4.12.9.

Shift Manager will ENSURE clear priorities are set and personnel roles and responsibilities are clear.

4.13.

Turnover of Control Room Command Durinq Transient and Casualty Situations (if required) 4.13.1.

CONDUCT the transfer of emergency procedure command and control during stable periods of low activity.

4.13.2.

PERFORM a walk-down / familiarization / pass-through of the status trees / EOPs prior to taking control room command.

4.13.3.

REVIEW procedure status, relevant primary and secondary plant parameters, abnormal conditions, known radiation releases, and initiating events as appropriate prior to taking control room command.

4.13.4.

RECEIVE permission from the Shift Manager prior to completing transfer of control room command.

5.

DOCUMENTATION - None

6.

REFERENCES 6.1.

Regulatory Guide 1.33, Appendix A.

6.2.

INPO 96-008, "Guidelines for the Conduct of Operations at Nuclear Power Stations."

6.3.

INPO Good Practice OP-201, "Shift Relief and Turnover."

7.

ATTACHMENTS - None

r-I'-1"eG1IV U Udi::l*;

REVIEW PORC NO SQR YES QR NO

".59 YES I\\

llUI

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Page 1 of 4 JAD/JHG:eer FUEL HANDLING DIRECTOR SHIFT TURNOVER CHECKLIST NOTE: Complete this form at turnover Date:

Oncoming Shift: X, Y, Z (circle one)

Part A - Offgoing Fuel Handling Director (FHD)

Evolutions in Progress / Planned for next shift:

Comments and Problem Areas:

Offgoing FHD Signature:

I L

JI Al I

I

-IVI-.-

I.

I S.

Ii..V.

Page 2 of 4 Part B - Oncoming FHD Checklist C

FHD Log and information on page 1 reviewed.

1.

2.

Fuel Floor local panel annunciators (both units at LGS) and Refuel Platform

2.

Interlock/Controls reviewed.

3.

At PBAPS ONLY: Verify Refueling Normal up limit is selected on Refuel Platform.

3.

4.

Status of Fuel Floor work reviewed with Maint. Supervisor and HP Technician.

4.

5.

HP Drywell Control Point contacted and has verified access control to upper elevations is in place as required.

HP Contacted:

5.

6.

Refueling Platform checkout is complete as required.

6.

7.

Visual check for foreign material in the Reactor cavity/Vessel completed per

7.

A-C-1 30, "Fuel Floor Foreign Material Exclusion".

8.

FHDs, Refuel Platform Operators (RPOs), and Spotters are qualified and have

8.

completed all required reading.

9.

Continuous communications established between Control Room and Refueling

9.

Platform during CORE ALTERATIONS.

10. FHD conducted, as necessary, a pre-evolution brief with refuel platform personnel.

10.

C-i.

Work environment assessed for error-free challenges; actions taken as necessary.

11.

(water clarity; water surface; ambient noise; ambient lighting; distractions on platform/floor; refuel platform operation - workarounds; time pressure)

12.

An HP qualified person is present on the Refuel Floor for any work in fuel pool,

12.

Reactor cavity or equipment pit.

13.

CCTAS reviewed, and verified same title/rev, as Reactor Operator and Spotter copies.13.

14. Auxiliary hoist jam blocks set properly during in-vessel maintenance.

14.

15.

At LGS Refuel Floor inflatable seal backup bottle cart staged (not blocked by other

15.

equipment) and diamond plate unobstructed at seal station for Seals No. 3 & 4.

16.

Fuel Floor permanent ARMs, temporary ARM(s) if required, and Refuel Platform

16.

ARM(s) indicate on scale.

"17. Spent Fuel Pool water level at least 22' over the top of irradiated fuel assemblies seated in the racks as required by Tech Specs. Verify as follows:

LGS:

Water level at wave suppression scupper OR over flowing weirs OR 17a.

water level not more than 1' 11" below curb level.

(*

PBAPS: Water level at least 232' 3".

17b._

These steps may apply to handling items over irradiated fuel. Refer to FH-105 at LGS and FH-6C at PBAPS for applicability.

Page 3 of 4 Part B - Oncoming FHD Checklist Continued

(

Refueling Interlocks and Refuel Platform equipment operable as required by

18.

Tech Specs. At LGS ST-6-097-630-1 (2) OR ST-6-107-633-1 (2), or at PBAPS ST-0-018-120-2(3) OR ST-O-018-125-2(3).

19.

Reactor Mode switch locked in REFUEL position.

19.

20.

All control rods are fully inserted and indicate full in as required by Tech Specs.

20.

"*21. RPV Water Level verified as required by Tech Specs as follows:

At LGS level as stated on OP Aid next to Shutdown and Upset Range 21a.

instruments LR-42-*R605, LR-42-*R608 AND is not to be less than 481.5" per Tech Specs.

At PBAPS level shall be > 458" above RPV instrument zero.

21b.

(Normally read on PR/FR-*-06-097, Black Pen, with Switch 6A-$3 in Reactor Refuel Level)

22.

Moderator temperature > 680 F OR a lower cycle-specific temperature

22.

specified by Reactor Engineering.

I

23. Both Recirc pumps secured OR all In-core Instruments supported.

23.

24.

RHR SDC flow < 6000 gpm at LGS or < 6500 gpm at PBAPS OR C r all In-core instruments supported.

24.

25.

At LGS SRMs are operable as required by Tech Specs.

25.

(Circle operable SRMs A B C D)

26. At PBAPS WRNMs are operable as required by Tech Specs.

26.

(Circle operable WRNMs. A B C D E F G H)

27. At PBAPS WRNM Signal-to-Noise Ratio checked per S12(3)N-60C-WRNM-A(B)4FM

27.

OR S12(3)N-60C-WRNM-A(-H)I C2 AND S12(3)N-60C-WRNM-A(-H)I MX.

28. Main Control Room Emergency Fresh Air Supply System at LGS or Main Control

28.

Room Emergency Ventilation System at PBAPS AND instrumentation operable as required by Tech Specs.

29. SBGTS operable as required by Tech Specs.

29.

30. Secondary Containment, SCIVs AND instrumentation operable as required by

30.

Tech Specs.

31.

ECCS operable as required by Tech Specs.

31.

32. Electrical systems AND instrumentation operable as required by Tech Specs.

32.

(

hese steps may apply to handling items over irradiated fuel. Refer to FH-105 at LGS and FH-6C at

'rB-APS for applicability.

LAIMIDM I INUIIVI-L'-t.

I.

1)

, nUV. V Page 4 of 4 Part B - Oncoming FHD Checklist Continued r

At PBAPS verify fuel pool cooling system(s) specified by Reactor Engineering in FH-6C, "Core Component Movement - Core Transfers" are in service, as required by FH-6C.

34.

RHR method(s) set per Tech Spec requirements.

35.

Current plant status reviewed with Unit Reactor Operator.

Comments 33.

34.__

35.

Oncoming FHD Signature:

Time:

(Revisions to this procedure shafl be verified to be consistent with FH-105, "Core Component Movement - Core Transfers" at LGS and FH-6C, "Core Component Movement - Core Transfers" at PBAPS.)

EXELON NUCLEAR,

..TITLE: Core Counting Locations at Limerick and Peach Bottom TASK PERFORMED BY:

EVALUATOR SIGNATURE:

DIRECTIONS TO EVALUATOR:

EVALUATOR:

DATE:

EVALUATION METHOD:

PERFORM EVALUATION LOCATION:

ANY APPROXIMATE COMPLETION TIME:

,Completion Time: 10 minutes IMPORTANCE RATING(S):

3.3 SYSTEM NUMBER(S):

2.2.3 Generic

REFERENCES:

1. RE-C-40 Rev 90, CORE COMPONENT TRANSFER AUTHORIZATION SHEET GENERATION AND ADMINISTRATION TASK STANDARD(S):

Component locations identified per the attached grading keys AZ. doc Page 3 of 7 A2.doc Page 3 of 7

EXELON NUCLEAR TASK CONDITIONS:

1. Overhead views of Peach Bottom and Limerick vessels are provided INITIATING CUES:

Circle the location of the following instruments and label with the respective core location:

Limerick - "D" Source Range Monitor (SRM)

Peach'Bottom -

E" Wide Range Nuclear Monitor (WRNM)

A2.dOC Page 4 of 7

/.,,......................

y.*

A2.dOC Page 4 of 7

EXELON NUCLEAR Critical Element(s) indicated by' in Performance Checklist.

PERFORMANCE CHECKLIST:

STEP STANDARD SP'iUNSAT

1. CUE: Provide the candidate with two blank core diagrams.

Note that the answer keys are attached to the JPM and are labeled. LGS and PBAPS cores are rotated 90 degrees from each other. LGS does not have WRNMs

2. *Mark the Limerick map at location 16-21 16-21 marked

3. *Mark the Peach Bottom map at location 32-29 marked 32-29 CUE: You have reached the termination point N/A N/A for this JPM. You may stop here.

I t*

A;* dOC Page 5 of 7

EXELOh NUCLEAR'.,..-

Comments:

Note: Any grade of UNSAT requires a comment.

JPM Overall Rating:

SSAT/UNSAT A2.doc Page 6 of 7 A2.do c Page 6 of 7

.t..

S..

V *'

M{. I

EXELON NUCLEAR

_XLO NUCEA TASK CONDITIONS:

1. Overhead views of Peach Bottom and Limerick vessels are provided INITIATING CUES:

Circle the location of the following instruments and label with the respective core location:

Limerick - "D" Source Range Monitor (SRM)

Peach Bottom - "E" Wide Range Nuclear Monitor (WRNM)

CANDIDATEj I

ICANDIDATE Limerick Unit 1 JPM A2

-pp "Cm

-o.

0J j

"-C

"D" SRM/

16-21 ANSWER KEY-LOS "A2

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JPM A2 ANSWER KEY 32-29

-PB wt

RE-C-40, Rev.

9 Page 1 of 47 JWH:cms SQR YES PORC NO NQA NO 50.59 YES RESP MGR YES PECO Energy Company NUCLEAR GENERATION GROUP CORE COMPONENT TRANSFER AUTHORIZATION SHEET GENERATION AND ADMINISTRATION 1.0 PURPOSE This procedure provides instructions for generation and administrative control of Core Component Transfer Authorization Sheets (CCTAS).

(Ref. 3.1.4) 2.0 SCOPE This procedure applies to CCTAS generated for refueling, in-vessel maintenance, and non-refueling movement of core components.

This procedure replaced FH-605 (LGS) and RE-43 (PBAPS).

3.0 SOURCES & REFERENCES 3.1 SOURCE DOCUMENTS 3.1.1 LGS Event Investigation Reports, 92-04-27, 92-04-35 "Double Blade Guide Misposition" (A/R A0400493 eval 02) 3.1.2 EPRI NSAC 164L, "Guidelines for BWR Reactivity Control During Refueling", April 1992 3.1.3 PEP Issue 10000293 3.1.4 LGS UFSAR Section 13.5.1.20, (A/R A0411306 Eval 77) 3.1.5 LGS PORC Commitment #91-021 3.1.6 NCRS LG 92-00329, LG 92-000330 3.1.7 Nuclear Regulatory Commission, IE Circular 80-21 3.1.8 PEP Issue 10003697 3.1.9 PEP Issue 10005166 3.1.10 LGS EWR A0004257, Attachment A 3.1.11 LGS Unit 2 Licensing Report For Spent Fuel Storage Capacity Expansion (HOLTEC Calculation HI-931012 Section 9.4)

RE-C-40, Rev.

9 Page 2 of 47 3.1.12 CM-I, Source Document NRC URI 76-46/36; re: annotation of fuel transfer sheets for authenticity, (T01850) 3.1.13 CM-2, PBAPS Event Investigation Report 3-91-201, (T02075) 3.1.14 CM-3, PEP Issue 10000326, (T03250) 3.1.15 CM-4, Letter to NRC 8/19/96; re: Unit 1 Fuel Pool Capacity Limitation to 2500, (T03881) 3.1.16 CM-5, Letter to NRC 8/20/96; re: LGS Procedural Controls for Projected Spent Fuel Pool Heat Loads, (T03941) 3.1.17 PEP Issue 10009435 3.1.18 PEP Issue 10009589 3.1.19 ECR PB 98-02407, MOD P00758 Refueling Platform Controls Upgrade 3.1.20 PEP Issue 10010012 3.1.21 PEP Issue 10010407 3.1.22 PEP Issue 10011047 3.1.23 PEP Issue 10011130 3.2 CROSS REFERENCES 3.2.1 Procedure FM-C-4, "Generation of Core Component Transfer Authorization Sheets using SHUFFLEWORKS".

3.2.2 Procedure RE-C-44, "Accountability and Control of SNM".

3.2.3 ShuffleWorks User's Manual, Combustion Engineering Inc.

3.2.4 Procedure ST-R-002-930-2(3),

"Refueling SDM Verification" (PBAPS Only).

3.2.5 Procedure FM-UG-423, "Determination of SRM/WRNM Count Rates" 3.2.6 Procedure SF-300, "TN-68 Cask Spent Fuel Assemblies Storage Selection and Document Requirements" (PBAPS only) 4.0 DEFINITIONS 4.1

CCTAS, CORE COMPONENT TRANSFER AUTHORIZATION SHEET:

An approved sequence specifying the movement of core components which include a paper copy and at PBAPS may include an electronic copy on disk for loading into the Refueling Platform computer.

4.2

ICA, ITEM CONTROL AREA:

Area designated in ShuffleWorks

RE-C-40, Rev.

9 Page 3 of 47

(

4.3 4.4 4.5 4.6 4.7 5.0 5.1 5.1.1 5.1.2 5.1.3 5.2 5.2.1 5.2.2 5.3 5.3.1 allowed to contain SNM, i.e.

fuel prep machine, fuel pool, reactor cores and casks.

CORE COMPONENT:

A component which is or was an integral part of the reactor core.

This includes fuel pins or assemblies (excluding "Dummy" bundles and individual fuel channels),

control blades, local power range monitors, wide range neutron monitors, blade guides, fuel support pieces and startup sources.

IRM/SRM/WRNM replacements above vessel are controlled by this procedure. Under vessel replacement of Nuclear Instruments (NIs) are controlled by reference 3.2.2.

SPECIAL NUCLEAR MATERIAL (SNM):

Any material containing plutonium, U-233, or uranium enriched in U-233 or U-235.

SNM is controlled by reference 3.2.2.

REFERENCE LOADING PATTERN:

The specific core loading arrangement of the nuclear fuel for the next operating cycle.

This arrangement of fuel is used for reload licensing analysis and is obtained from the associated Core Design Report.

SHUTDOWN MARGIN (SDM):

Amount of reactivity by which the reactor is subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in shutdown condition; cold (68 'F) and xenon free. CM-3, (Ref. 3.1.2).

SHUFFLEWORKS: Computer software which is used to generate fuel/core component sequences.

RESPONSIBILITY AND AUTHORITY REACTOR ENGINEERING BRANCH MANAGER:

CM-3 Assign a member(s) of staff to perform this procedure and generate CCTAS.

Authorize CCTAS generation and performance.

Designate individuals qualified to evaluate and authorize changes to CCTAS in accordance with section 8.4.

STAFF MEMBER(s) assigned to perform this procedure shall:

Maintain current data files for SHUFFLEWORKS.

Generate and review the CCTAS in accordance with section 8.0.

CCTAS USERS shall:

Ensure CCTAS is appropriate for the task to be performed.

RE-C-.40, Rev.

9 Page 4 of 47 5.3.2 Ensure the CCTAS step completion is documented appropriately.

5.3.3 Return completed CCTAS to Reactor Engineering Branch.

5.3.4 Initiate CCTAS changes as required in accordance with section 8.4.

6.0 PREREQUISITES I 6.1 Appropriate SDM model AND Once Burnt Bundle constraint have been identified by Nuclear Fuel Management (NFM)

(applicable to refueling activities only).

6.2 Reference loading pattern is available (applicable to refueling activities only).

6.3 K-infinities, SDM criteria and radial weighting factors have been loaded into ShuffleWorks.

(Fuel moves in reactor vessel only) CM-3.

6.4 IF CCTAS involves steps in the core, THEN determine neutron monitoring instrumentation (SRM or WRNM) applicable to CCTAS being generated.

See Exhibit RE-C-40-11.

6.5 Information required for other CCTAS applications such as fuel pool reconfiguration, new fuel receipt, CRB exchanges and cask loading.

6.6 ICA files with current and accurate information.

6.7 Any fuel pool locations that are inaccessible or deemed to be unusable have been designated as "UNUSABLE" in the ShuffleWorks computer program.

See Exhibits RE-C-40-9 & 10.

6.8 Any core components that are deemed unusable/not moveable (e.g., damaged blade guide, leaking fuel assembly, etc.)

have been designated as "UNUSABLE" in the shuffleWorks program.

See Exhibits RE-C-40-9 & 10.

6.9 For a reload sequence, FSD has provided the predicted SRM/WRNM (as appropriate) count rates for the first four bundles surrounding each SRM or WRNM in accordance with Ref. 3.2.5.

6.10 For cask loading, an acceptable cask loading pattern has been developed in accordance with SF-300.

(PBAPS only)

RE-C-40, Rev.

9 Page 5 of 47 7.0 PRECAUTIONS AND LIMITATIONS 7.1 When moving fuel in the reactor, sequences must meet the SRM or WRNM connectivity constraint as described in Exhibit RE-C-40-13.

7.2 Do not generate a CCTAS step which places fuel into a cell which has its associated control rod withdrawn.

7.3 Do not generate a CCTAS step which withdraws a fuel assembly or blade guide unless the control rod in the cell is supported diagonally by blade guides or fuel.

7.4 Do not place irradiated fuel with less than 5 year decay time within 3 feet of fuel pool gates (actual or storage locations).

Refer to Exhibit RE-C-40-15 for impacted locations.

(Ref. 3.1.10) (LGS ONLY) 7.5 Fuel assemblies discharged prior to 1R04 located in the Unit 1 spent fuel pool must be inspected or evaluated for debris prior to use in core.

(Ref. 3.1.5)

(LGS ONLY) 7.6 The spent fuel pool capacity is limited to 2500 fuel assemblies, and a nominal 6.625 inches center-to-center spacing between fuel assemblies shall apply, until the completion of the high density re-racking of the spent fuel pool.

(LGS Unit 1 ONLY)

CM-4 7.7 Single blade guides must be oriented such that their channel spacers are adjacent to the control rod.

Place a note stating such requirement prior to CCTAS step which uses a single blade guide in the reactor core.

7.8 During complete offload/reload or core shuffle with SRM operable, bundles in the cells on the A/D, A/B, B/C, and C/D quadrant boundaries should be considered to be monitored by the D, B,

C and C SRM respectively. Connectivity to these SRMs should be verified.

See Exhibit RE-C-40-11.

7.9 CCTAS columns which are not applicable to the specific application should have N/A placed in the column which does not apply.

7.10 SHUFFLEWORKS shall be used to generate CCTAS when it is available and recognizes all the components to be moved.

CM-2

RE-C-40, Rev.

9 Page 6 of 47 NOTE All open cells of a given fuel rack should be inspected when any of the cells of that rack are planned to be used (Ref. 3.1.22).

7.11 Prior to CCTAS use, NMD is to verify empty cells that will be used do not contain foreign material.

Foreign material is not permitted in the component location.CM-2 7.12 CCTAS changes shall be made in accordance with Section 8.4 of this procedure.

(Ref. 3.1.2) 7.13 For reload sequence, load four bundles around each SRM/WRNM prior to any other steps.

(Ref. 3.1.2)

Place a note on the CCTAS stating the predicted count rate for the first four bundles surrounding each SRM or WRNM.

7.14 IF peripheral core locations listed in Exhibit RE-C-40-10 are being used in the sequence, THEN verify a note is added prior to the step which states:

"Caution: Mast camera may interfere with core shroud"(PBAPS Only).

NOTE (LGS ONLY)

The objective of performing RT-1-053-310-* is to confirm that adequate heat removal capacity is present in the target fuel pool.

Performance of a new RT is not required if the results of a previous performance of the RT bounds the anticipated fuel pool configuration.

7.15 Before distributing a CCTAS which moves spent fuel into a Fuel Pool, RT-1-053-310-* must be or have been performed to ensure that adequate heat removal capacity is available in the Fuel Pool (LGS ONLY).

CM-5 7.16 Movement of bundles into Fuel Pool rows "A" or "B" is difficult, because the spotter can NOT view the specified rows from the bridge.

I__F practical, THEN avoid rows "A" and "B"

during outage work (LGS Unit 1 ONLY).

7.17 During a refueling outage, the CCTAS should not discharge fuel to locations within two feet of the spent fuel pool walls as this causes the refuel platform to move at slow speeds and causes unnecessary delays.

RE-C-40, Rev.

9 Page 7 of 47 S7.18 Ensure that fuel is stored next to the fuel pool coupons.

For Unit 1, the coupons are located next to the cells noted in Exhibit RE-C-40-9.

For Unit 2, the coupon tree may be moved such that it is surrounded by freshly discharged fuel assemblies.

A refuel outage discharge plan should leave a space available for this requirement.

If a bundle surrounding the Unit 2 coupon tree is moved, another freshly discharged bundle shall be moved to replace it.

(LGS ONLY) 7.19 In order to avoid obstruction of the RHR and FPC discharge piping flow path to support natural circulation operation, Do Not load rack cells directly under and within two cells of this piping in the LGS 2 spent fuel pool.

(LGS ONLY) 7.20 WHEN preparing CCTASs for new CRB receipt or CRB exchange where the combined CRB/FSP grapple may be used, THEN verify that locations for the new CRBs as well as locations planned for the discharge or temporarily stored CRBs have CRB spacers installed prior to using the locations either by documentation or direct observation.

(LGS Only) 7.21 Sipping canisters may only be used in CRB rack cells that do not contain CRB spacers.

(LGS only) 7.22 Fresh discharged fuel shall not be placed in the eastern most rows to prevent high exposure to personnel.

(Ref. 3.1.11) (LGS Unit 2 ONLY) 7.23 When generating a CCTAS to allow for the removal/reinstallation of a CRB/fuel support piece, it is desirable that at least one of the bundles face adjacent to the north-west quadrant (at LGS) or south-west quadrant (at PBAPS) of the fuel support piece is off-loaded, provided that SRM or WRNM connectivity constraint is met as described in exhibit RE-C-40-13.

This is in order to allow camera access for verification of plunger engagement.

7.24 When preparing a CCTAS that includes re-channeling of fuel bundle(s),

include a note prior to each step (as appropriate) which provides the expected channel serial number and requires the fuel handlers to check that the correct channel is being used. (Ref. 3.1.17)

7.25 7.25.1 7.25.2 7.25.3 7.25.4 7.26 "7.27 7.28 7.29 7.29.1 RE-C-40, Rev.

9 Page 8 of 47 For items stored in locations designated in Exhibits RE-C-40-9 & 10, the location may be made usable and the item may be moved provided:

If the item is listed under Bad Fuel, the move is noted in the CCTAS to facilitate resolution of ARM alarms, AND The item is not used in the core without, after being apprised of the item's status in Shuffleworks, specific FSD and Reactor Engineering Branch Manger approval is obtained, AND Other constraints on the location (such as bundles next to Coupons) are met or the move is specifically approved by the System Manager or Reactor Engineering Branch Manager, AND The new storage location is made unusable (unless the location is in-core) and a PPIS issued against the appropriate exhibit following item movement.

If it is determined that a bundle should not be used in-core (due to debris, failure, future inspections, etc.), then SHUFFLEWORKS shall be updated such that the bundle's storage location is made unusable and a note added to the bundle which describes the issue.

If the bundle cannot be used in-core permanently, a PPIS to update Exhibits RE-C-40-9/10 should be created.

(Ref. 3.1.18)

Spent Fuel Assemblies that are being transferred to a cask may be loaded into the cask in any orientation unless a specific orientation is desired, such as cask hold-down ring interference with the channel fasteners.

IF very old fuel is being moved, THEN ensure the CCTAS includes a note identifying this and that the following precautions should be taken:

(Ref. 3.1.20)

1.

Carefully monitor hoist load during lifting to promptly identify bundle tie rod failure.

2.

Channel fasteners and bail handles have been found bent on old bundles which can make seating or releasing difficult.

3.

Older style channel fasteners protrude further making it easier for fastener to catch when seating.

FOR CCTAS preparation in support of unit outage (Reference 3.1.21):

Ensure no current outage discharge fuel is placed in rack rows 25-48 between rack columns A-H, inclusive (at LGS Unit 1), in rack rows 32-54 between rack columns TT-AB, inclusive (at LGS Unit 2)

OR in rack rows A-G between rack columns 25 48, inclusive (at PBAPS).

(Ref. 3.1.21).

I

RE-C-40, Rev.

9 Page 9 of 47 7.29.2 Uninspected cells should be made unusable once pre-outage inspections are complete (Ref. 3.1.22).

7.30 For non-outage CCTAS preparation, add requirements to perform the following CCTAS steps, as applicable:

7.30.1 Following placement of spent fuel within 8 rows of spent fuel pool/reactor cavity gates or fuel pool/cask pit gates (LGS Only),

contact HP to perform dose measurement at fuel pool gates at reactor cavity or fuel pool side.

7.30.2 Ensure general area dose rate reading at fuel pool gates at reactor cavity/ cask pit (LGS Only) or fuel pool side is acceptable per HP.

I 7.30.3 IF dose rate is unacceptable, return fuel bundle to original location or other location as designated by Reactor Engineering.

7.31 Ensure all four fuel assemblies off-loaded around each LPRM that Will be replaced during the Maintenance window.

IF this is not possible due to other constraints, THEN plan a mini-shuffle between the CRB/CRD AND LPRM maintenance windows.

7.32 Do not insert single blade guides to take the place of double blade guides in cells with the control rod withdrawn.

The single blade guide may damage the control rod during insertion or "kick out" to adjacent cells.

7.33 When moving core components between Spent Fuel Pools, the CCTAS cover sheet should contain a note reminding the FHD to contact HP to have the Reactor Buildings' 313' elevation properly posted for core component movement (LGS Only).

7.34 When moving LPRMS, ensure LPRM hanger number/rigging tag number is recorded on CCTAS via note associated with respective CCTAS movement step.

8.0 PROCEDURE SECTIONS 8.1 CCTAS GENERATION WITH SHUFFLEWORKS 8.1.1 Complete Exhibit RE-C-40-1, "CCTAS Information/Verification Sheet" with information that is applicable for the particular evolution.

8.1.2 Execute the SHUFFLEWORKS computer program in accordance with Exhibit RE-C-40-8.

8.1.3 IF the CCTAS are being generated automatically, THEN perform the following steps, Otherwise the following steps are N/A.

RE-C-40, Rev.

9 Page 10 of 47

1.

Ensure the following constraints are turned on in SHUFFLEWORKS:

a.

Shutdown Margin model specified by NFM in step 6.1.

b.

SRM or WRNM Connection Path (consistent with neutron instrumentation determined in step 6.4)

c.

SRM or WRNM Count Rate (consistent with neutron instrumentation determined in step 6.4)

d.

Fuel Movement Device Mast Orientation

e.

Control Rod Blade Support

f.

Fuel Movement Device Unusable locations

g.

Once Burnt Bundle constraint specified by NFN in step 6.1.

2. Enter the SHUFFLE menu, select automatic shuffle planning.

a.

Supply information as requested by the SHUFFLEWORKS program (maintenance locations, discharge and temporary storage locations, bundles to be sipped/inspected).

b.

Generate and save the sequence.

c.

PBAPS ONLY:

IF sequence is required to be loaded into Refueling Platform Computer "Fuel Machine Sequence"

(.fms)

format, THEN perform the following:

Select Reports/Create fuel machine sequence Select appropriate sequence source and planning sequence Save the sequence to file (converts from.seq to.fms output)

Select "WRITE SEQUENCE TO A DISK FILE" Enter "SHUFFLE" for output file name Verify "SHUFFLE.FMS" file was created

RE-C-40, Rev.

9 Page 11 of 47 Copy this file to a disk and label disk with CCTAS title CCTAS preparer shall initial and date the disk label.

3.

Ensure SDM log values are above the cycle specific SDM specified by NFM and all applicable constraints are met.

CM-3.

a.

Confirm SDM limit is correct in appropriate SDM Model.

b.

Verify the sequence with SHUFFLEWORKS.

c.

Perform ST-R-002-930-2(3) for CCTASs containing steps which insert-positive reactivity into the core.

(PBAPS Only).

d.

IF the SDM log values are below the cycle specific SDM value specified by NFM, THEN send the sequence to NFM for SDM evaluation. See FM-C-4.

4.

Print the shuffle sequence and applicable ICA maps (if desired) via the Reports window and proceed to section 8.3 of this procedure.

8.1.4 IF the CCTAS are being generated interactively, THEN perform the following steps, Otherwise the following steps are N/A.

1.

Ensure the following constraints are turned on in SHUFFLEWORKS:

a.

Shutdown Margin model specified by NFM in step 6.1.

b.

SRM or WRNM Connection Path (consistent with neutron instrumentation determined in step 6.4)

c.

SRM or WRNM Count Rate (consistent with neutron instrumentation determined in step 6.4)

d.

Fuel Movement Device Mast Orientation

e.

Control Rod Blade Support

f.

Fuel Movement Device Unusable locations

RE-C-40, Rev.

9 Page 12 of 47

g.

Once Burnt Bundle constraint specified by NFM in step 6.1.

2.

Enter the SHUFFLE menu, select interactive shuffle planning.

a. Move the desired component(s) from the initial ICA location to the desired ICA location.

3.

Ensure all applicable constraints are met.

a. Verify and save the sequence with SHUFFLEWORKS.

b. PBAPS ONLY:

IF sequence is required to be loaded into Refueling Platform Computer "Fuel Machine Sequence"

(.fms) format, THEN perform the following:

Select Reports/Create fuel machine sequence Select appropriate sequence source and planning sequence Save the sequence to file (converts from

.seq to.fms output)

Select "WRITE SEQUENCE TO A DISK FILE" Enter "SHUFFLE" for output file name Verify "SHUFFLE FMS" file was created Copy this file to a disk and label disk with CCTAS title CCTAS preparer shall initial and date the disk label.

c.

IF sequence involves moves in the core, THEN ensure SDM log values are above the cycle specific SDM specified by NFM.

CM-3

1.

Confirm SDM Limit is correct in appropriate SDM Model.

2.

Perform ST-R-002-930-2(3) for CCTASs containing step which insert positive reactivity into the core.

(PBAPS Only)

3.

IF the SDM log values are below the cycle specific SDM value specified by NFM, THEN send the sequence to NFM for SDM evaluation.

See FM-C-4.

4.

Print the shuffle sequence and applicable ICA maps (if

RE-C-40, Rev.

9 Page 13 of 47 desired) via the Reports window AND proceed to section 8.3 of this procedure.

8.1.5 IF the CCTAS are being generated for core OR bundle offload/reload, THEN perform the following steps, Otherwise, the following steps are N/A.

1.

Ensure the following constraints are turned on in SHUFFLEWORKS:

a.

SRM or WRNM Connection Path (consistent with neutron instrumentation determined in step 6.4)

b.

SRM or WRNM Count Rate (consistent with neutron instrumentation determined in step 6.4)

c.

Fuel Movement Device Mast Orientation

d.

Control Rod Blade Support

e.

Fuel Movement Device unusable locations

f.

Once Burnt Bundle constraint specified by FSD in step 6.1.

2.

Enter the SHUFFLE menu, select interactive shuffle planning.

a.

Determine the required positioning of blade guides to support planned in-vessel maintenance activities.

b.

For each quadrant, move bundles from core to spent fuel pool using the following guidelines:

1.

IF SRMs were selected in step 6.4, THEN from core periphery, spiral toward the SRM leaving 4 fuel bundles around the SRM.

IF WRNMs were selected in step 6.4, THEN from core periphery, spiral toward the WRNM detector leaving 4 fuel bundles around each WRNM.

2.

Remove fuel bundles from diagonally adjacent fuel cell locations in a cell where a double blade guide is to be located.

3.

Install a double blade guide where fuel was removed.

RE-C-40, Rev.

9 Page 14 of 47

4.

Remove remaining two bundles from the cell.

5.

Perform steps 1 through 4 for all other core cells.

6.

WHEN all other fuel has been offloaded, THEN remove last 16 bundles around the SRMs OR remove last 32 bundles around the WRNMs.

c.

For each quadrant, move bundles from the spent fuel pool to the final core locations using the following guidelines:

(Ref. 3.1.2)

1.

Load 4 bundles around each SRM (IF SRMs were selected in step 6.4) OR each WRNM (IF WRNMs were selected in step 6.4).

Place a note on the CCTAS prior to each step or group of steps stating the predicted count rates for the first four bundles surrounding each SRM or WRNM.

2.

From the SRM OR WRNM, spiral towards core periphery

3.

Load 2 fuel bundles in vacant locations

4.

WHEN two fuel bundles are installed in a cell THEN remove blade guide.

5.

Place fuel in vacant locations created by blade guide removal.

3.

Verify AND save the sequences with SHUFFLEWORKS.

4.

PBAPS ONLY:

IF sequence is required to be loaded into Refueling Platform Computer "Fuel Machine Sequence"

(.fms)

format, THEN PERFORM the following:

Select Reports/Create fuel machine sequence Select appropriate sequence source and planning sequence Save the sequence to file (converts from

.seq to.fms output)

Select "WRITE SEQUENCE TO A DISK FILE" Enter "RELOAD" OR "OFFLOAD" for output file name, as appropriate.

Verify the appropriate FMS file was created Copy this file to a disk and label disk

RE-C-40, Rev.

9 Page 15 of 47 with CCTAS title CCTAS preparer shall initial and date the disk label.

5.

Print the offload/reload sequence AND applicable ICA maps (if desired) via the Reports window AND proceed to section 8.3 of this procedure.

8.1.6 During CCTAS execution, update CCTAS status log sheet (see Exhibit RE-C-40-2).

8.2 CCTAS GENERATION WITHOUT SHUFFLEWORKS 8.2.1 Generate CCTAS using a word processor.

1.

The CCTAS shall contain the following information:

Title Page #/total

of pages Written, reviewed, and authorized spaces Date Step #

Component Serial Orientation (when applicable)

From location To location FHD signoff RPO signoff CRO signoff (when applicable)

SRM/WRNM count rate columns consistent with neutron instrumentation determined in step 6.4 (when applicable).

Date/Time for each step

RE-C-40, Rev.

9 Page 16 of 47 8.2.2 IF generating CCTAS for LPRM replacement THEN perform the following steps, Otherwise, the following steps are N/A.

1.

Use Exhibit RE-C-40-3A or RE-C-40-3B as a format/info guide (Separate CCTASs may be generated for batch removal/bending and for installation of LPRMs).

2.

Identify LPRM to be replaced

3.

Ensure blade guide does not interfere with LPRM

4.

Specify LPRM removal and storage locations.

Storage locations to avoid due to area radiation monitors are:

a.

LGS North Walls

b.

PBAPS Unit 2 North Wall

c.

PBAPS Unit 3 South Wall

5.

Use LPRM serial number as core component ID 8.2.3 IF generating CCTAS for control rod withdrawal, THEN perform the following steps.

Otherwise, the following steps are N/A.

1.

Use Exhibit RE-C-40-4A or RE-C-40-4B as a format/info guide.

2.

Identify control rods to be withdrawn

3.

Verify cell is offloaded

4.

Select control rod in offloaded cell

5.

Withdraw control rod in offloaded cell

6.

Uncouple control rod in offloaded cell NOTE Fuel Support Pieces and Guide Tubes must be returned to the same core cells they were removed from unless specifically noted in the CCTAS and agreed upon by NMD.

8.2.4 IF generating the CCTAS for control rod replacement, THEN perform the following steps.

Otherwise, the following steps are N/A.

1.

Use Exhibit RE-C-40-5A or RE-C-40-5B as a format/info guide

2.

Identify control rod(s) to be replaced

RE-C-40, Rev.

9 Page 17 of 47

3.

Verify control rod is uncoupled

4.

Remove blade guide from target cell

5.

Remove fuel support piece from target cell

6.

Remove spent control rod from cell to storage location

7.

IF Guide Tube removal is required, THEN provide steps to:

a.

Place Blade Guide and Fuel Support Piece in storage locations.

b.

Remove Guide Tube and place in storage location (if required).

c.

Return Guide Tube to core cell.

d.

Pick up Fuel Support Piece and Blade Guide from storage locations.

8.

Move new control rod from storage location to core cell

9.

Replace fuel support piece in target cell

10.

Replace blade guide in target cell

11.

IF the NMD combined Blade/Fuel Support Piece grapple is to be used, THEN add appropriate notes stating that steps 5 and 6, and steps 8 and 9 above may be performed concurrently.

8.2.5 IF generating CCTAS for non-core transfer OR spent fuel assemblies transfer to/from cask, THEN perform the following steps.

Otherwise, the following steps are N/A.

1.

Use Exhibit RE-C-40-6A or RE-C-40-6B as a format/info guide

2.

Identify core component to be transferred

3.

Specify the starting and resting location of the core component.

8.2.6 IF SHUFFLEWORKS is unavailable AND CCTAS for offload/reload fuel moves as described in step 8.1.5 are required, THEN perform the following step.

Otherwise, the following steps are N/A.

1.

Use Exhibit RE-C-40-7A or RE-C-40-7B as a format/info guide

2.

Perform Steps 8.1.5.2.a through 8.1.5.2.c.

RE-C-40, Rev.

9 Page 18 of 47 8.2.7 S8.2.8 8.3 8.3.1 Proceed to the section 8.3 of this procedure.

During CCTAS execution, update CCTAS status log sheet (see Exhibit RE-C-40-2).

CCTAS REVIEW/APPROVAL/DISTRIBUTION The CCTAS preparer shall perform the following steps:

1.

Ensure no unexplained/unexpected constraint violations exist.

(Ref. 3.1.2)

2.

Ensure component identification numbers, orientations and locations are correct.

(Ref. 3.1.2)

3.

Ensure moves containing known/suspected fuel failures are noted in CCTAS to facilitate HP monitoring/resolution of ARM alarms.

4.

Annotate steps which insert or remove a bundle from next to an SRM/WRNM (if time allows).

5.

Ensure accuracy and legibility.

(Ref. 3.1.3)

6.

Ensure the prepared by, reviewed by, AND authorized by names are printed on each sheet.

CM-I

7.

PBAPS ONLY:

IF the shuffleworks file is to be loaded into the Refueling Platform computer, THEN ensure the prepared by, reviewed by AND authorized by names are printed on the electronic disk copy label.

8.

For control blade AND LPRM replacements, ensure specific item types are placed in required locations.

(Ref. 3.1.8)

9.

Fill out appropriate information on CCTAS status log sheet (see Exhibit RE-C-40-2).

10. Generate AND sign the CCTAS Instruction/Approval sheet.

(Exhibit RE-C-40-12)

CM-i.

11.

PBAPS ONLY:

IF the shuffleworks file is to be loaded into the Refueling Platform Computer, THEN initial and date the electronic disk copy label.

12.

Ensure Fuel Pool decay heat removal capacity has been demonstrated in accordance with RT-I-053-310-* for any increased heat load on Fuel Pool (LGS Only).

CM-5

RE-C-40, Rev.

9 Page 19 of 47

13.

PBAPS ONLY:

IF CCTAS moves the fuel to the final configuration, THEN perform the following:

a.

Step to the end of the sequence using SHUFFLEWORKS.

b.

Print "Current" core map AND spent fuel pool map.

c.

Verify "Current" core map matches Core Design Report (reference loading pattern).

d.

Complete Exhibit RE-C-40-14 AND attach to spent fuel pool map.

14.

Complete Exhibit RE-C-40-1, "CCTAS Information /

Verification Sheet" 8.3.2 The CCTAS reviewer shall perform the following:

1.

Ensure no unexpected/unexplained constraint violations exist.

(Ref. 3.1.2)

2.

Ensure component identification numbers, orientations and locations are correct.

(Ref. 3.1.2)

3.

Ensure moves containing known/suspected fuel failures are noted in CCTAS to facilitate HP monitoring/resolution of ARM alarms.

4.

Ensure steps which insert or remove a bundle from next to an SRM/WRNM are correctly annotated (if performed by preparer).

5.

Ensure accuracy and legibility.

(Ref. 3.1.3)

6.

Ensure the preparer and authorized names are printed on each sheet.

CM-i

7.

PBAPS ONLY:

IF the shuffleworks file is to be loaded into the Refueling Platform computer, THEN ensure the preparer and authorized names are printed on the electronic disk copy label.

8.

For control blade AND LPRM replacement CCTAS, ensure specific item types are placed in required locations.

(Ref. 3.1.8)

9.

Ensure maximum heat load/temperature of SFP is within limits in accordance with RT-I-053-310-* for any increased heat load on Fuel Pool (LGS Only).

CM-5

RE-C-40, Rev.

9 Page 20 of 47

10.

Complete Exhibit RE-C-40-1, "CCTAS Information /

Verification Sheet"

11.

Sign the CCTAS Instruction/Approval sheet.

(Exhibit RE-C-40-12)

CM-i

12.

PBAPS ONLY: IF the shuffleworks file is to be loaded into the Refueling Platform Computer, THEN initial and date the electronic disk copy label.

13.

PBAPS ONLY:

IF CCTAS moves the fuel to the final configuration, THEN perform the following:

a.

Step to the end of the sequence using SHUFFLEWORKS.

b.

Print "Current" core map AND spent fuel map.

c.

Verify "Current" core map matches Core Design Report (reference loading pattern).

d.

Complete Exhibit RE-C-40-14.

8.3.3 The CCTAS authorizer shall:

1.

Sign the CCTAS info Instruction/Approval sheet (RE-C-40-12) sheet indicating authorization to execute the steps.

CM-i.

2.

PBAPS ONLY: IF the shuffleworks file is to be loaded into the Refueling Platform Computer, THEN initial and date the electronic disk copy label.

8.3.4 Distribute copies of approved CCTAS to the following locations.

1.

Control Room (when performing in-core moves or as requested for information only).

2.

Refuel platform (2 copies plus electronic disk copy if required)

3.

Other locations as deemed necessary by the evolution. (e.g., Fuel Floor Coordinator info only copy)

4.

Cognizant personnel from Nuclear Fuel Management (Control Blade replacement CCTASs only) (Ref. 3.1.8)

5.

Dry Cask Storage Campaign Personnel (For cask loading/unloading only)

RE-C-40, Rev.

9 Page 21 of 47 CCTAS CHANGE PROCESS AFTER REVIEW/APPROVAL IF it is deemed necessary to change the CCTAS to support unanticipated conditions on the fuel floor, THEN perform following steps.-

(Ref. 3.1.2) the

1.

Suspend CCTAS execution prior to execution of CCTAS step requiring change.

2.

Contact a member of Reactor Engineering who is authorized to make changes as listed on the CCTAS instruction sheet, Exhibit RE-C-40-12.

3.

Describe the reason for the change and a proposed revision The Reactor Engineering group member contacted shall perform the following steps based on the information provided.

(Ref. 3.1.3)

1.

Evaluate the effect of the change on the remainder of the CCTAS sequence AND all other previously approved CCTAS's (including completed and partially completed sequences) AND make appropriate changes to maintain congruency of sequence.

NOTE PBAPS ONLY: A pen and ink change made to the paper copy CCTAS does NOT require a change to the electronic copy CCTAS loaded in the Refueling Platform computer.

Refueling Platform may be operated in the Manual Mode to execute the affected steps using the paper CCTAS.

Full Automatic Mode operation using the electronic copy CCTAS loaded in the Refueling Platform computer may resume after completion of the pen and ink changed steps.

8.4 8.4.1 8.4.2

RE-C-40, Rev.

9 Page 22 of 47

2.

IF all the changes involve one of the situations listed

below, THEN the changes do not affect SDM AND can be processed as a pen & ink change.

a.

Blade guides

b.

LPRMs

c.

WRNMs

d.

Fuel in or being moved to the spent fuel pool

e.

Control rod blade changes in defueled cells or in the spent fuel pool

f.

Fuel in or being moved to or from a storage cask

g. All of the bundles currently in the core are in their final configuration AND the fuel bundle(s) of interest are being moved to their final core configuration (final core reload pattern remains the same)

3.

IF any of the changes involve a fuel assembly with a change in core configuration, THEN use SHUFFLEWORKS to ensure SDM is met in accordance with Section 8.1 and follow approval process in Section 8.3.

CM-3

4.

IF the changes do not adversely affect SDM, THEN perform the following:

a.

The Fuel Handling Director (or Designated Alternate) or Reactor Engineer shall perform pen & ink change on Refuel Floor copies.

b.

The Reactor Engineering group member shall perform pen & ink change on the original CCTAS AND troubleshooting guides when applicable. Updates pen & ink change log if used.

c.

The Reactor Engineering group member, Fuel Handling Director (or Designated Alternate)

OR CRO shall perform the Pen & Ink change on the control room copy of the CCTAS (if applicable).

d.

The Reactor Engineering group member OR CCTAS Lead shall make the change to the sequence Saved in SHUFFLEWORKS.

RE-C-40, Rev.

9 Page 23 of 47

e.

The pen & ink change should follow the following format.

Single line through item to be changed Date AND initials of both Fuel Handling Director (or Designated Alternate) AND Reactor Engineering group member

5.

IF the change is believed to have a significant effect on decay heat calculations, THEN consider reperforming RT-1-053-310-* AND follow approval process in Section 8.3 (LGS Only).

CM-5 8.5 CCTAS TROUBLESHOOTING GUIDE DEVELOPMENT 8.5.1 IF desired, THEN develop CCTAS troubleshooting guide so any changes can be evaluated appropriately.

1.

The guide should contain the following information:

a.

Appropriate CCTAS sequences

b.

Maps at key milestones/configurations such as:

(Ref. 3.1.2)

LPRM's/WRNM's to be replaced CRD's requiring maintenance Control blades requiring replacement Core maintenance configuration

c.

List of bundle K-infinity for fuel moved during refueling

d.

Copy of RE-C-40

e.

CCTAS Info sheet

f.

Pen & Ink Change/CCTAS Revision Status Log.

g.

Lists/maps of maintenance control blades and drives and LPRMS.

h.

Lists/maps of previously inspected fuel pool locations suitable for item storage.

i.

Copies of/support information for RE sign offs in any applicable FH procedures.

2.

Provide troubleshooting guide to Reactor Engineering group members authorized to approve changes.

RE-C-40, Rev.

9 Page 24 of 47 9.0 DOCUMENTATION 9.1 Completed CCTAS sheets AND electronic disk copy OR copies of contaminated sheets shall be returned to Reactor Engineering by the CCTAS users.

9.1.1 IF copies of contaminated CCTASs are received, THEN cover sheet should be annotated "Copy - Original Contaminated",

initialed, and dated.

9.2 IF the CCTAS become contaminated, AND copies cannot be made, THEN the CCTAS originals can be used to document the completion of the CCTAS steps.

9.2.1 Note on the original CCTAS that field copies were contaminated and were executed between days X and Y.

9.3 Reactor Engineering shall periodically send completed CCTAS OR copies of appropriate CCTAS if contaminated, to NRMS and NFM (if required).

9.4 Update SHUFFLEWORKS files upon completion of the CCTAS.

Every implemented CCTAS shall be incorporated in SHUFFLEWORKS regardless of whether or not the CCTAS appears to change the location/orientation of any fuel bundle.

(Ref. 3.1.9) 9.5 Update SNM records in accordance with RE-C-044, Accountability and Control of SNM.

If re-channeling has been performed, ensure channel serial number's are appropriately updated on the Fuel History Sheets.

9.6 Backup SHUFFLEWORKS files appropriately.

9.7 Provide copies of completed blade replacement CCTASs to FSD.

9.8 Provide copies of SHUFFLEWORKS sequence files (or copy of controlled CCTAS if not created with SHUFFLEWORKS) for all completed fuel moves to NFM for updating of SNM records/isotopic reports.

10.0 EXHIBITS 10.1 RE-C-40-1 CCTAS Info Sheet 10.2 RE-C-40-2 CCTAS Status Log 10.3 RE-C-40-3A Example CCTAS for LPRM Replacement (SRM only) 10.4 RE-C-40-3B Example CCTAS for LPRM Replacement (WRNM only) 10.5 RE-C-40-4A Example CCTAS for Blade Withdrawal (SRM Only) 10.6 RE-C-40-4B Example CCTAS for Blade Withdrawal (WRNM Only) 10.7 RE-C-40-5A Example CCTAS for Blade Replacement (SRM Only)

RE-C-40, Rev.

9 Page 25 of 47 10.8 RE-C-40-5B Example CCTAS for Blade Replacement (WRNM Only) 10.9 RE-C-40-6A Example CCTAS for Non Core Transfers 10.10 RE-C-40-6B Example CCTAS for Spent Fuel Assemblies Transfers 10.11 RE-C-40-7A Example CCTAS for Core Transfers (SRM Only) 10.12 RE-C-40-7B Example CCTAS for Core Transfers (WRNM Only) 10.13 RE-C-40-8 Executing SHUFFLEWORKS 10.14 RE-C-40-9 LGS Unusable/Inaccessible Spent Fuel Pool Locations 10.15 RE-C-40-10 PBAPS Unusable/Inaccessible/Restricted Locations 10.16 RE-C-40-11 Core Map 10.17 RE-C-40-12 CCTAS Instruction/Approval Sheet 10.18 RE-C-40-13 SRM/WRNM Connectivity Constraint 10.19 RE-C-40-14 Final Configuration Map Checklist 10.20 RE-C-40-15 LGS Fuel Pool Cells Within Three (3)

Feet Of Gate Seals

Exhibit RE-C-40-1 Page 1 of 1 RE-C-40, Rev.

9 Page 26 of 47 CCTAS INFORMATION/VERIFICATION SHEET CCTAS

Title:

File Name:

Directory:

Description of CCTAS:

Location/file name of Discharge Locations:

Location/file name of Temporary Storage:

NI's to be replaced:

Control Blades to be replaced/swapped:

Control Rod Drives to be worked:

Bundles to be Inspected/Sipped/Rechanneled/Reseated:

(Circle/provide value or N/A if not applicable)

SDM Limit:

SDM Model:

Neighborhood Size:

1 5

9 Exposed Assemblies around each SRM Once Burned Assemblies around a Control Blade 1

Once Burned Face Adjacent Allowed 1

Minimum k-infinity Value for Once Burned Assemblies:

Connectivity Path:

SRM Control Rod Blade Support U

Core Unusable Locations:

U SFP U

SFP U

Core Item Type Locations:

U SFP U

SFP U

Core Fuel Movement Device Mast Orientation:

U SFP U

SFP U

Core Date/WO of RT-I-053-310-*

(LGS Only):_

Spent Fuel Cask (PBAPS ONLY):

S/N Other Precautions and Limitations of RE-C-40 Section 7.0 met:

Yes STROD2 STROD3 25 1

2 2

Full 2

3 3

WRNM U

Core "U Core "U Core Core 3

4 4

None None None U

Core None N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A ro (Exceptions Listed)

Signature specified have been confirms that constraints/data provided above are correct and applied as AND that all verifications specified in Section 8.3.1/8.3.2(as appropriate) done:

Prepared By:

Reviewed By:

Date:

Rev.

Date:

Constraints:

Exr t RE-C-40-2 Page 1 of 1 47 RE-C-40 V. 9 Page 2. -)f 47 CCTAS STATUS LOG:

TAB CCTAS Steps DATE/TIME PERFORMED ORIGINALS RETRIEVED**

Updated In CCTAS TITLE REV #

STATUS*

Performed START END FHD RPO CR0 S/Works SNM

Ended or Canceled "X"

it original or "Copy" it original was contaminated

Exhibit RE-C-40-3A Page 1 of 1 Unit Title EXAMPLE CCTAS FOR LPRM REPLACEMENT SRM ONLY r,,

RE-C-40, Rev.

9 Page 28 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO SRM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

I LPRM XX-YY NONE BENDER NONE NA NA NA NA X,YYY,ZZZ II 2

LPRM BENDER NONE UNIT 1 SPENT FUEL POOL NONE NA NA NA NA X,YYY,ZZZ WALL Above LPRM hung on Tag #

3 LPRM FUEL FLOOR NONE XX-YY NONE NA NA NA NA X,YY,ZZZ

r-r Exhibit RE-C-40-3B WRNM ONLY RE-C-40, Rev.

9 Page 1 of 1 Page 29 of 47 Unit Title EXAMPLE CCTAS FOR LPRMIWRNM REPLACEMENT Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO WRNM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

E F

G H

1 LPRM/WRNM XX-YY NONE BENDER NONE NA NA NA NA NA NA NA NA XYYY,ZZZ 2

LPRM/WRNM BENDER NONE UNIT 2 SPENT FUEL POOL NONE NA NA NA NA NA NA NA NA X,YYY,ZZZ WALL Above LPRMIWRNM hung on Tag #

3 LPRM/WRNM FUEL FLOOR NONE XX-YY NONE NA NA NA NA NA NA NA NA X,YYY,ZZZ

Exhibit RE-C-40-4A SRM ONLY Page 1 of 1 RE-C-40, Rev.

9 Page 30 of 47 Written By:

Unit Title EXAMPLE CCTAS FOR BLADE WITHDRAWAL Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO SRM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

Step 1 for each rod may be performed anytime after fuel movement has ceased for maintenance I

1 VERIFY CONTROL CELL 02-43 IS OFFLOADED NA NA NA NA NA The CRD/CRB Coordinator may sign as FHD for steps 2, 3, and 4 2

SELECT CONTROL ROD 02-43 NA WARNING: IF ANY SRM COUNTRATE DOUBLES INSERT CONTROL ROD TO POSITION 00 3

WITHDRAW CONTROL ROD 02-43 TO POSITION 48 NA 4

UNCOUPLE CONTROL ROD 02-43 PER M-C-747-011 OR M-C-741-301 NA NA NA NA NA

WRNM ONLY Exhibit RE-C-40-4B Page 1 of 1 Unit Title EXAMPLE CCTAS FOR BLADE WITHDRAWAL RE-C-40, Rev.

9 Page 31 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO WRNM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

E F

G H

Step 1 for each rod may be performed anytme after fuel movement has ceased for maintenance 1

VERIFY CONTROL CELL 02-43 IS OFFLOADED NA NA NA NA NA NA NA NA NA The CRD/CRB Coordinator may sign as FHD for steps 2, 3, and 4 2

SELECT CONTROL ROD 02-43 NA WARNING: IF ANY WRNM COUNTRATE DOUBLES INSERT CONTROL ROD TO POSITION 00 3

WITHDRAW CONTROL ROD 02-43 TO POSITION 48 NA 4

UNCOUPLE CONTROL ROD 02-43 PER M-C-747-011 OR M-C-741-301 NA NA NA NA NA NA NA NA NA

Exhibit RE-C-40-5A SRM ONLY Page 1 of 2 Unit Title EXAMPLE CCTAS FOR BLADE REPLACEMENT RE-C-40, Rev.

9 Page 32 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO SRM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

1 VERIFY CONTROL ROD 30-47 IS UNCOUPLED NA NA NA NA NA 2

DBL B/G 31-48/29-46 NONE HANG MN HOIST NONE NA NA NA NA 3

SUP PC 30-47 NW HANG HOIST NONE NA NA NA NA 4

IF CRD EXCHANGE IN PROGRESS, NA NA NA NA NA NA NA THEN PERFORM STEPS 4A, 4B, & 4C, OTHERWISE, N/A STEPS 4A,4B, & 4C 4A INFORM UNDERVESSEL TEAM OF INTENT TO REMOVE CRB NA NA NA NA NA NA 4B RAISE CRB 30-47 APPROXIMATELY 12 INCHES NA NA NA NA NA 4C RECEIVE CONFIRMATION FROM UNDERVESSEL TO CONTINUE NA NA NA NA NA NA 5

CRB 30-47 NONE CBRCK 5N-1 NONE NA NA NA NA 6

CRB H605 CBRCK 6N-4 NONE 30-47 NONE NA NA NA NA 7

SUP PB HANG HOIST NONE 30-47 NW NA NA NA NA

SRM ONLY Exhibit RE-C-40-5A Page 2 of 2 Unit Title EXAMPLE CCTAS FOR BLADE REPLACEMENT RE-C-40, Rev.

9 Page 33 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO SRM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

8 NOTIFY HCU/CRD COORDINATOR CRB 30-47 FULLY SEATED NA NA NA NA NA 9

DBL B/G HANG MN HOIST NONE 31-48M29-49 NONE NA NA NA NA 10 CHECK SEATING OF DBL B/G WITH CAMERA NA NA NA NA NA

Exhibit RE-C-40-5B Page 1 of 2 Unit Title EXAMPLE CCTAS FOR BLADE REPLACEMENT WRNM ONLY RE-C-40, Rev.

9 Page 34 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO WRNM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

E F

G H

1 VERIFY CONTROL ROD 30-47 IS UNCOUPLED NA NA NA NA NA NA NA NA NA 2

DBL BIG 31-48/29-46 NONE HANG MN HOIST NONE NA NA NA NA NA NA NA NA 3

SUP PC 30-47 NW HANG HOIST NONE NA NA NA NA NA NA NA NA 4

IF CRD EXCHANGE IN PROGRESS, NA NA NA NA NA NA NA NA NA NA NA THEN PERFORM STEPS 4A, 4B, & 4C, OTHERWISE, N/A STEPS 4A,4B, & 40C 4A INFORM UNDERVESSEL TEAM OF INTENT TO REMOVE CRB NA NA NA NA NA NA NA NA NA NA 48 RAISE CR1 30-47 APPROXIMATELY 12 INCHES NA NA NA NA NA NA NA NA NA 4C RECEIVE CONFIRMATION FROM UNDERVESSEL TO CONTINUE NA NA NA NA NA NA NA NA NA NA 5

CRB 30-47 NONE SFP NONE NA NA NA NA NA NA NA NA 6

CRB H605 CBRCK 6N-4 NONE 30-47 NONE NA NA NA NA NA NA NA NA 7

SUP PB HANG HOIST NONE 30-47 NW NA NA NA NA NA NA NA NA

Exhibit RE-C-40-5B WRNM ONLY RE-C-40, Rev.

9 Page 2 of 2 Page 35 of 47 Unit Title EXAMPLE CCTAS FOR BLADE REPLACEMENT Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO WRNM COUNTRATE DATE TIME NO, SERIAL NO.

A B

C D

E F

G H

8 NOTIFY HCUICRD COORDINATOR CRB 30-47 FULLY SEATED NA NA NA NA NA NA NA NA NA 9

DBL 1IG HANG MN HOIST NONE 31-48/29-46 NONE NA NA NA NA NA NA NA NA 10 CHECK SEATING OF DBL B/G WITH CAMERA NA NA NA NA NA NA NA NA NA EXhiblt RE-C-40-6A RE-

, Rev.

C Page 36 of 47 Page 1 of 1 Unit Title EXAMPLE CCTAS FOR NON CORE TRANSFERS Written By: _

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO DATE TIME NO.

SERIAL NO.

1 LY8487 L2SPENT KK-158 SW LlSPENT T-58 SW 2

LY8488 L2SPENT KK156 SW L 1SPENT U-57 SW 3

LY8811 L2SPENT KK-154 SW L1SPENT U-56 SW 4

LYNO91 L2SPENT KK-152 SW LISPENT V-58 SW 5

LY8812 L2SPENT KK-150 SW LISPENT V-56 SW 6

LY8813 L2SPENT KK148 SW LlSPENT W-58 SW 7

LYN104 L2SPENT KK146 SW L1SPENT W-57 SW

C C

Exhibit RE-C-40-6B Page 1 of 1 Unit Title EXAMPLE CCTAS FOR SPENT FUEL ASSEMBLIES TRANSFERS C

RE-C-40, Rev.

9 Page 37 of 47 Written By:

Reviewed By:

Authorized By:

r 1

1 f

COMPONENT SERIAL NO.

MOVE FROM ORIENT MOVETO ORIENT I

+

1-T 1

P7RPFNT C.-2 N/A CASK S/N A-1 N/A FHD RPO DATE

'1

lr,')

N/A It t

4 1

+

I I

1 1

1 PYG651 P2SPENT N-46 N/A CASK S/N G-1 N/A 2

PYG6151 I

I1 P2SPENT C-21 N/A CASK SIN C-6 N/A

'A_

PY46 t2PN C 21 4

PYG764 P2SPENT P-46 N/A CASK SIN

, F-5 N/A 5

PJ1407 P2SPENT C-20 N/A CASK S/N_

_ A-7 N/A 1

1

.1.

1 1-I I

1 1

1 TIME NOTE:

See RE-C-44 Figure 10 for cask alphanumeric layout STEP NO.

DVNIK91 2

PVIt 'JA...

.1.

1 I

I 1

1 1

A SRM ONLY Exhibit RE-C-40-7A Page 1 of 1 Unit Title EXAMPLE CCTAS FOR CORE TRANSFERS RE-C-40, Rev.

9 Page 38 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO SRM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

I LYN521 LICORE 01-44 SE L1SPENT C-22 NW 2

LYG651 LICORE 03-42 NW LiSPENT N-46 SW 3

DBL B/G L1SPENT B-31/C-32 NONE LICORE 01-44/03-42 NONE 4

LYN463 LiCORE 01-42 NE L1SPENT C-21 SW 5

LYG764 L1CORE 03-44 SW L1SPENT P-46 SW 6

YJ1407 L1CORE 19-52 SW L1SPENT C-20 SW 7

YJ1447 LICORE 17-50 NE L1SPENT C-18 SW

WRNM ONLY Exhibit RE-C-40-7B Page 1 of 1 Unit Title EXAMPLE CCTAS FOR CORE TRANSFERS RE-C-40, Rev.

9 Page 39 of 47 Written By:

Reviewed By:

Authorized By:

STEP COMPONENT MOVE FROM ORIENT MOVE TO ORIENT FHD RPO CRO WRNM COUNTRATE DATE TIME NO.

SERIAL NO.

A B

C D

E F

G H

PYN521 P2CORE 01-44 SE P2SPENT C-22 NW 2

PYG651 P2CORE 03-42 NW P2SPENT N-46 SW 3

DBL B/G P2SPENT B-31/C-32 NONE P2CORE 01-44/03-42 NONE 4

PYN463 P2CORE 01-42 NE P2SPENT C-21 SW 5

PYG764 P2CORE 03-44 SW P2SPENT P-46 SW 6

PJ1407 P2CORE 19-52 SW P2SPENT C-20 SW PJ1447 P2CORE 17-50 NE P2SPENT C-18 SW

Exhibit RE-C-40-8 RE-C-40, Rev.

9 Page 1 of 1 Page 40 of 47 EXECUTING SHUFFLEWORKS

1.

Log on to LAN on SHUFFLEWORKS-loaded Computer.

If SHUFFLEWORKS is running directly off of the PC desktop, then verify with NFM that the correct version of shuffleworks is being run.

2.

Double-click on SHUFFLEWORKS icon.

3.

Change data directory using the "set-up" menu option.

4. Load appropriate plant information using the "set-up" window option.

5.

Follow program prompts/pull-down menu options in order to execute the desired SHUFFLEWORKS functions.

Exhibit RE-C-40-9 Page 1 of 1 RE-C-40, Rev.

9 Page 41 of 47 LGS UNUSABLE/INACCESSIBLE SPENT FUEL POOL LOCATIONS UNIT 1 UNIT 2 BINDING:

COVERED BY LABELS A-il R-14 Z-18 JJ-23 B-2 thru B-18, A-19 thru A-66,B-73 thru B-84 A-35 T-18 AA-69 KK-16 R-2 thru R-84 G-21 U-19 EE-47 SS-23 LL-72 thru LL-85, U-79 thru KK-79 P-20 X-22 FF-70 SS-35 AB-19 thru AB-70 Q-19 X-34 GG-23 VV-22 A-I thru T-1, U-5 thru ZZ-5 Q-21 Y-15 JJ-ii X-42 A-85 thru T-85 CC-43 HH-38 S-52 T-51 U-56 thru ZZ-56 T-52 T-53 U-52 U-21 thru ZZ-21 OBSTRUCTED E-44 thru 49 A-77 E-67 thru 71 KK-77 through TT-77 OBSTRUCTED BY GATE BRACKETS LL-88 through SS-88 AB-53 thru AB-59, AB-49 thru AB-51 BAD FUEL AB-26 thru AB-33, AB-36 B-14 JJ-18 through JJ-22 D-12 KK-18 J-31 KK-20 OBSTRUCTED OBSTRUCTED BY CLIPS K-32 KK-22 A-I thru A-3, A-50 S-3 V-20 L-23 KK-23 A-82 thru A-85 S-17 KK-20 M-24 LL-15 B-i thru B-3 S-20 NN-20 CC-21 LL-17

B43, B50, and B73 S-35 YY-20 DD-16 through DD-22 LL-19 B-82 thru B-85 S-38 EE-16 LL-43 C-I thru C-4 S-53 V-55 P.E-18 through EE-22 RR-20 C50 and C73 S-56 KK-55

"-16 through FF-23 RR-22 C-82 thru C-85 S-70 NN-55 b-16 RR-51 D-I thru D-4 S-73 YY-55 GG-18 through GG-22 UU-18 D-82 thru D-85 S-83 HH-16 through HH-22 UU-22 E-i thru E-5 A-3 JJ-16 E-82 thru E-85 A-17 F-I thru F-6 A-74 TEST COUPONS ADJACENT TO THE FOLLOWING CELLS F-82 thru F-85 A-83 P-58 BB-70 Z-70 Q-58 G-1 thru G-6 V-4 T-58 DD-70 EE-70 R-58 H-i thru H-3 KK-4 CC-70 Y-70 J-I thru J-3 NN-4 K-i thru K-3 YY-4 OTHER ZZ-i thru ZZ-3 V-80 FF-63 AB-I thru AB-4 JJ-80 NN-36 BAD FUEL K-23 K-15 X-1 OTHER V-1 L-83 IT-9 I T-i IU-52 FF-69 FF-70 ZZ-22 HH-47 QQ-36 INACCESSIBLE DUE TO LEVEL SWITCH SS-36 N-84 EP-84 N:Q-84 SS-39 N-85 P-85 Q-85 INACCESSIBLE DUE TO REFUEL BRIDGE ACCESS All of Rows 1,

2, and 85 TO PREVENT UNPLANNED MOVEMENT 8 Bundles around coupon tree

Exhibit RE-C-40-10 Page 1 of 1 RE-C-40, Rev.

9 Page 42 of 47 Exhibit RE-C-40-10 PBAPS UNUSABLE/INACCESSIBLE SPENT FUEL POOL LOCATIONS POTENTIAL MAST CAMERA INTERFERENCE PERIPHERAL CORE LOCATIONS (Both Units) 01-18 11-06 43-02 55-12 01-44 11-56 43-60 55-50 05-12 17-02 49-06 59-18 0--5-50 17-60 49-b6 59-44 UNIT 2 UNIT 3 Covered by rack markers:

Covered by rack markers:

A-25 thru WW-25, A-17 THRU LL-17, YY-18 thru HHH-18, XX-21 thru XX-55 LL-1 thru LL-20, MM-20 thru XX-20, Fuel Pool Gate Heavy Load Path:

NN-21 thru NN-55 A-21 thru A-55 Fuel Pool Gate Heavy Load Path:

Boraflex Liner Damage:

A-21 thru A-55 JJ-30, MM-30 FPM Mounting Concerns:

Boraflex Coupon:

HHH-2 thru HHH-5, HHH-19 thru HHH-22 Q-l, S-i Damaged DBG:

Boraflex Coupon:

T-71/S-72 CCC-71 Failed Fuel Pin:

Dechanneled Bundle:

CCC-71 Z-51 Debris:

Bowed Bundles:

LLL-30, LLL-33 X-54, EE-57 Other:

Other:

'G-2, GGG-3 S-72 thru XX-72 EEE-66 thru KKK-66 11-45, JJ-46, GGG-2, HHH-2 EEE-67 thru EEE-71 GGG-66 ("skeleton" bundle)

S-72 thru XX-72 JJ-22 Unusable Fuel Locations:

Unusable Fuel Locations:

BB-18, BB-19, BBB-25, BBB-3, AA-39, AAA-14, AAA-7, BB-36, BBB-4, CC-13, CC-15, CC-16, BBB-13, BBB-15, BBB-19, BBB-20, CCC-6, DD-13, DD-16, DD-17, DD-18, BBB-21, BBB-3, BBB-61, BBB-63, DD-19, DD-20, DD-21, DD-22, DD-23, BBB-64, CCC-2, D-22, DD-26, DD-58, E-1, DD-24, EEE-5, F-18, F-21, FFF-3 EE-37, EE-5, EE-6, F-17, GG-21, HHH-17, JJ-12, KK-12, LL-12, HH-12, HH-18, HH-4, HH-56, HH-62, NN-7, P-19, PP-51, PP-55, PP-7, HH-64, HH-7, HHH-1I, HHH-14, PP-8, QQ-8, RR-10, T-55, U-18, U-19, HHH-17, 11-10, 11-19, 11-61, JJ-8, V-18, WW-67, YY-65, YY-7, ZZ-2, K-41, KK-55, L-31, L-41, LL-15, LL-20, ZZ-3, ZZ-4, ZZ-5 LL-54, LLL-32, M-6, MM-9, N-40, NN-16, 0-40, 00-54, PP-17, PP-5, PP-53, PP-7,- QQ-53, QQ-54, QQ-62, QQ-68, QQ-8, SS-53, SS-54, SS-55, SS-6, T-13, TT-13, TT-54, U-38, U-9, VV-16, VV-17, W-13, W-38, W-39, W-7, X-37, X-54, Y-36, YY-67, YY-8, Z-39, Z-51, Z-68, ZZ-67, ZZ-68

Exhibit RE-C-40-11 Page 1 of 1 RE-C-40, Rev.

9 Page 43 of 47 CORE MAP 0L 030507 09 l. I3 t5 1

7

[

9 21 EL 3252 60 56A 54 DM 50 52 ASS 48 I

3] 3335 373941 434547495153555759

- ~

60 I ID 59 ED]

u ODOM56 q rfP 54 52 50

-46

~SH 48 EDgO Bum 36 V/_. _ :1~~_* _ m m~~~_

'34

+-++'+

m~~~

r7/--++-+P+

BEIM

I$,+**s n-n r-nn, rrc, rrrtrr.rn rnrlrfl Irsi ml

.r

n.ru an.rsrun.rr¶ an. [vu-an.rrv It o

LPRM H

IRM OR WRNM a

SRM OR Spare WRNM Tube IcŽN V.../I VA'N./I*, V./hl*'.

VAJLŽN, [/ZJIL'&k 4,Jl VA.I* VAJl U.

29 Q o

~n 26 I~~tlM I~

EM-l B-t 122

-- maim 14 12 to fr~z

-f02 J 3.335 37 39 4j 43 454749 5J 53 555759 FJIF7 24 BO 0

m 08 mjv**

22r '1j N,*N 10 104g 06 Di*

M 0I]ID**

02 1

9L95 0 L 03109 0709 i131921 17 L9 2120325 2; im-i

Exhibit RE-C-40-12 RE-C-40, Rev.

9 Page 1 of 1 Page 44 of 47 CCTAS Instruction/Approval Page (Sample Sheet Only)

Core Component Transfer Authorization Sheet Page of Unit Date Title (Rev. #)

1.

This CCTAS provides the moves necessary to...

2.

This CCTAS assumes that SRM A, B,

C, D (or WRNM A through H) are in service.

3.

(Guidance on placement of a fuel bundle which cannot be placed in original target location, e.g. SRMs/WRNM inoperable or in an emergency)

4.

This CCTAS must be used in conjunction with appropriate maintenance procedures.

5.

This CCTAS must be used in conjunction with appropriate fuel handling procedures.

6.

This CCTAS must be used in conjunction with appropriate spent fuel

(

procedures.

(41r.

All changes to this CCTAS must be made in accordance with procedure RE-C-40.

8.

This CCTAS involves the transfer of Special Nuclear Material (SNM).

SNM transfers shall be controlled in accordance with procedure RE-C-44.

9.

Return completed CCTAS (or a copy if original is contaminated) to Reactor Engineering.

10.

(RE contacts and phone numbers)

(etc.)

Prepared By:

Reviewed By:

Authorized By:

Exhibit RE-C-40-13 RE-C-40, Rev.

9 Page 1 of 1 Page 45 of 47 SRM/WRNM CONNECTIVITY CONSTRAINT The SRM or WRNM connectivity constraint consistent with the neutron instrumentation determined in step 6.4 is to be used for in-core shuffles and will also be applied to offload/reload.

Face Adjacent Rules:

1.

A continuous face adjacent path must exist between each bundle and an operable SRM OR between each bundle and at least one operable WRNM.

2.

The only exceptions to rule #1 are bundles that fulfill the diagonal coupling rules.

Diagonal Coupling Rules:

1.

It is permissible for single fuel bundles ONLY to be diagonally connected to an array of face adjacent fue-"bu-ndles.

No other fuel bundles can be connected to the single diagonally adjacent fuel bundle except per rule #2.

2.

In addition to the diagonally adjacent connection path allowed by rule

1, two fuel bundles diagonally adjacent within the same control cell are considered connected.

Exhibit RE-C-40-14 RE-C-40, Rev.

9 Page 1 of 1 Page 46 of 47 FINAL CONFIGURATION MAP CHECKLIST Preparer Reviewer

1.

"Current" Final Configuration core map matches Core Design Report.

2.

Discharge fuel bundles identified on spent fuel pool map.

3.

New fuel/reinsert bundle empty locations identified on spent fuel pool map.

4.

Reload bundle (temporary) empty locations identified on spent fuel pool map.

5.

Discharge bundles and empty locations on preparer's map matched reviewer's map.

6.

Previously completed Exhibit RE-C-40-14 AND spent fuel pool map discarded.

Exhibit RE-C-40-15 Page 1 of 1 RE-C-40, Rev.

9 Page 47 of 47 LGS FUEL POOL CELLS WITHIN THREE (3)

FEET OF GATE SEALS Per AT EWR A0004257, irradiated fuel which has decayed for less than 5 years cannot be stored within three (3) feet of the fuel pool gate seals.

This is to support justification for extending the seal lifetime from S to 10 years.

This constraint is conservatively assumed to include gate storage locations as well as gate locations in place.

UNIT 1 SPENT FUEL POOL Cattle Chute Gate Storage:

(North Side)

E-77 through BB-77 E-88 through BB-88 E-1 through BB-1 E-2 through BB-2 E-3 through BB-3 F-4 through AA-4 (South Side)

V-64 through V-70 W-64 through W-70 X-64 through X-70 Y-64 through Y-70 Z-64 through Z-70 AA-65 through AA-70 BB-66 through BB-70 Cattle Chute Gate In Place:

A-25 through A-46 B-25 through B-46 C-25 through C-46 D-25 through D-46 E-25 through E-46 F-26 through F-45 UNIT 2 SPENT FUEL POOL Cattle Chute Gate Storage:

(North Side)

VV-44 through VV-64 WW-43 through WW-65 XX-43 through XX-65 YY-43 through YY-65 ZZ-43 through ZZ-65 AB-43 through AB-65 (South Side)

VV-22 through VV-42 WW-21 through WW-43 XX-21 through. XX-43 YY-21 through YY-43 ZZ-21 through ZZ-43 AB-21 through AB-43 Cattle Chute Gate In Place:

VV-33 through VV-53 WW-32 through WW-54 XX-32 through XX-54 YY-32 through YY-54 ZZ-32 through ZZ-54 AB-32 through AB-54 Cask Pit Gate RR-77, RR-88, SS-77, SS-88, TT-77, TT-88, UU-77, UU-88, VV-77, VV-88, In Place:

RR-I through SS-l through TT-I through UU-I through VV-l through RR-16 SS-16 TT-16 UU-16 VV-16 Cask Pit Gate In Place:

A-1 through A-20 B-I through B-20 C-i through C-20 D-1 through D-20 E-1 through E-20 F-i through F-19

.,N.,.

.NUCLEAR EXELON NUCLEAR I

Administrative Questions for PBAPS Cat A.3 Radiation Control A3.doc

Question A3-1 Candidate Copy Reference material is not available for this question Given the following conditions:

An Maintenance Technician has just exited the refueling platform and walked to the Personal Contamination Monitor (PCM)..

The individual was handling poles and cables that were submerged in the spent fuel pool[

The operator alarmed the PCM one time

1. What are the operators required actions?

2. If the monitor alarms a second time, what are the required actions?

CANDIDATE

Question A3-2 Candidate Copy Question:

Peach Bottom Unit 2 conditions are as follows:

LPRM exchange is scheduled to begin within the next 15 minutes Work is in progress in the drywell on Elevation 135' Assuming the drywvell work must continue, what protective measures are taken by Health Physics to prevent those individuals from receiving excessive dose as the result of LPRM removal?

CANDIDATE

Question A3-1 GRADING KEY Reference material is not available for this question Given the following conditions:

An Maintenance Technician has just exited the refueling platform and walked to the Personal Contamination Monitor (PCM)

The individual was handling poles and cables that were submerged in the spent fuel pool Theoperator alarmed the PCM one time

1. What are the operator's required actions?

2. If the monitor alarms a second time, what are the required actions?

Answer:

1. Re-monitor one time

2. If the monitor alarms for a second time, then

a. Remain in the area

b. Minimize contact with personnel, tools, or equipment

c. Contact Health Physics*

Minimum required answer K/A 2.3.2 (2.9) Knowledge of Facility ALARA program Source: New ANSWER KEY

I Question A3-2 GRADING KEY Peach Bottom Unit 2 conditions are as follows:

LPRM exchange is scheduled to begin within the next 15 minutes Work is in progress in the drywell on Elevation 135' Assuming the drywell work must continue, what protective measures are taken by Health Physics to prevent those individuals from receiving excessive dose as the result of LPRM removal?

1. *Access must be restricted to prevent access to the upper drywell elevations (194' and above and 184' inside the bioshield)

2. A radiological monitoring system must be installed and monitoring the upper drywell and bioshield access points with local alarms

3. Drywell workers must be briefed prior to entering the drywell How will the FHD be informed that the actions have been taken?

The drywell control point will contact the FHD

Minimum required answer K/A 2.3.10 '(3.3)

Source: New

Reference:

HP-321 ANSWER KEY

Effective Date:10/01/01 HP-321, Rev.

6 Page 1 of 5 HBM:gpm I SQR YES PORC NO NQA NO 50.59 YES RESP MGR YES PECO Energy Company Nuclear Generation Group HEALTH PHYSICS REQUIREMENTS FOR IRRADIATED COMPONENT MOVEMENT 1.0 PURPOSE The purpose of this procedure is to define the Health Physics requirements and radiological controls for the Fuel Floor and Drywell during irradiated core component movement in the vessel and transfer to and from the Spent Fuel Pool.

Drywell requirements are for when the Drywell is accessible.

2.0 REFERENCES

2.1 GP -

15, Local Evacuation 2.2 ON -

124, Fuel Floor and Fuel Handling Problems -

Procedure 2.3 "Drywell Access During Irradiated Core Component Movement at

PBAPS, An Evaluation to Support Increased Drywell Access During 2R10,"

M. A. Horvatinovic and D.

C. DiCello, June 22, 1994.

2.4 "Addendum to 'Drywell Access During Irradiated Core Component Movement at PBAPS'",

D.

A. Droddy and R.

P.

Farrell, September 7, 1995 3.0 PREREQUISITES 3.1 The Refueling Channel Shield (Cattle Chute) should be installed during irradiated core component transfers.

NO ACCESS shall be allowed to the Drywell if it is not installed.

3.2 Radiological monitoring shall be utilized during irradiated core component transfers and will remain in effect until transfer is complete.

3.3 Communication capabilities shall be established and maintained between the Fuel Floor and Drywell control points during irradiated core component movements.

4.0 PRECAUTIONS 4.1 No access shall be allowed to elevations above the 194' elevation of the Drywell or inside the bioshield above 1841 during core component movements.

5.0 5.1 5.1.1 5.2 5.2.1 5.2.2 5.2.3 5.2.4 5.3 5.3.1 Sw6.0 6.1 6.2 6.3 6.4 7.0 7.1 7.1.1 7.1.2 7.1.3 HP-321 Rev.

6 Page 2 of 5 RESPONSIBILITY HEALTH PHYSICS SUPERVISION Directing HP Technician response to an Irradiated Core Component movement incident.

HEALTH PHYSICS TECHNICIANS Briefing of workers accessing drywell and the fuel floor.

Providing positive Health Physics Coverage during irradiated core component movement.

Notifying HP Supervision and Shift Management of any adverse conditions during movement that affect Radiological conditions.

Initiating emergency corrective actions.

i.e.

Drywell evacuation, Fuel floor stop work authority.

FUEL FLOOR FUEL HANDLING DIRECTOR Responsible for notifying the Drywell control point prior to the starting and completion of irradiated component movements.

APPARATUS AND SPECIAL EQUIPMENT Portable Area Radiation Monitor located on the fuel bridge.

Continuous Air Monitor (AMS3 or equivalent)

Portable Area Radiation Monitors located in the Drywell.

Communication Equipment PROCEDURE REFUEL FLOOR IRRADIATED CORE COMPONENTS OPERATIONS Fuel Floor Fuel Handling Director shall notify the Drywell Control Point HP prior to Irradiated Core Component movement.

There shall be a Area Radiation Monitor (ARM) on the fuel bridge at all times during irradiated core component movement.

The Drywell Control Point technician shall notify the Fuel Floor Fuel Handling Director that the Drywell is ready and no one is above the 194' elevation or inside the bioshield above 184' prior to allowing irradiated core component movements to begin.

HP-321 Rev.

6 Page 3 of 5 7.1.4 Upon completion of the movement activities, the Fuel Floor Fuel Handling Director shall notify the Drywell Control Point HP.

7.2 DRYWELL IRRADIATED CORE COMPONENT OPERATIONS 7.2.1 The Drywell Control Point HP shall be notified by the Fuel Floor Fuel Handling Director of irradiated core component movements.

1. The Drywell HP technician shall document all notifications in the Drywell Control Point logbook.

This documentation should include the name of the person making the notification and their position title.

7.2.2 Access control will be established to control access to 194' and above.

This can be accomplished by controlling access to the Drywell Airlock (prohibiting entry) or by installing ladder locks on 135',

157' or 186' to prohibit access to 194'.

A sign shall be posted on the airlock or ladder locks which states "Fuel Movement in Progress No Entry Above this Point".

7.2.3 Open Bioshield penetrations above 184' shall be posted that access is prohibited.

7.2.4 The Drywell HP technician shall ensure no one is on or above the 194' elevation by visual inspection.

7.2.5 A radiological monitoring system shall be installed in the drywell.

1. Two area radiation monitors with local alarms should be located on elevation 1941 approximately 180 degrees apart from each other.

One of the detectors should be placed underneath the refueling channel (U2 @ AZ 90 deg.,

U3 @

AZ 270 deg.).

2.

Area radiation monitors with local alarms should be placed in the vicinity of each open bioshield penetration being worked between the 160' and 184' el.

7.2.6 The Drywell Control Point HP technician shall inform the Fuel Floor Fuel Handling Director that no personnel are on or above the 194' elevation, the area is controlled and access is restricted.

7.2.7 All Drywell workers shall be briefed prior to entering the Drywell when irradiated core component movement is pending or occurring.

This briefing shall include the potential hazards, precautionary measures taken and personnel actions to take in the event of an alarm of the local ARM or notification of a hazard or evacuation.

7.2.8 The Fuel Floor Fuel Handling Director shall notify the Drywell Control Point HP when irradiated core component movement is complete.

HP-321 Rev.

6 Page 4 of 5 7.2.9 The Drywell Control Point HP shall survey the 194' elevation and above upon notification irradiated core component movement is complete.

7.2.10 If no adverse conditions are found, the Drywell Control Point HP technician can depost the area and allow access.

7.2.11 If significantly elevated dose rates are found, continue restricting access and notify Fuel Floor Fuel Handling Director, Fuel Floor Control Point HP technician, and HP Supervision.

7.3 DRYWELL EMERGENCY RESPONSE 7.3.1 The Drywell Control Point HP technician shall initiate a

Drywell evacuation if notified by the Fuel Floor of an irradiated core component incident or on alarm of a Drywell local ARM.

1. Contact the Shift Operations Supervisor in the Main Control Room and request a local evacuation of the Drywell to the SOP area in accordance with GP-15.

2.

Contact HP Supervision.

3.

If the evacuation is in response to a local ARM alarm, contact the Fuel Floor HP technician.

7.3.2 All personnel in the Drywell shall evacuate to the Drywell SOP and await instructions from the Drywell Control Point HP technician.

7.3.3 The Drywell Control Point HP technician should closely monitor the remote readouts for dose rate readings.

7.3.4 The HP technician shall obtain approval from HP Supervision to enter the Drywell for dose rate evaluation.

7.4

ý FUEL FLOOR EMERGENCY RESPONSE 7.4.1 Area Radiation Monitor Alarm - Fuel Bridge

1. The Fuel Floor technician shall survey and evaluate dose rates on the fuel floor and fuel bridge to determine sources of any ARM alarm.

2.

Notify HP supervision, Shift Manager, and the Fuel Handling Director of any adverse conditions.

3.

If the potential exists to exceed the exposure limits specified on any applicable RWPs, HP Supervision shall approve and initiate corrective actions to eliminate the source of the ARM alarm.

HP-321 Rev.

6 Page 5 of 5 7.4.2 Accidental Release or Damage of Irradiated Core Component

1. The Fuel Floor HP will initiate Stop Work Authority.

a. All personnel shall evacuate to the nearest location off the Fuel Floor as directed by Health Physics.

b. Contact the Drywell Control Point HP technician.

c. Contact HP Supervision and the Shift Manager.

2.

The Fuel Floor HP shall evaluate conditions and report them to HP Supervision.

7.4.3 Fuel Floor evacuations will be performed in accordance with ON-124 Fuel Floor and Fuel Handling Problems - Procedure.

8.0 DEFINITIONS None 9.0 EXHIBITS 9.1 None

. EXELON NUCLEAR TITLE: Determine EAL for refueling accident conditions TASK PERFORMED BY:

EVALUATOR SIGNATURE:

DIRECTIONS TO EVALUATOR:

EVALUATOR:

DATE:

EVALUATION METHOD:

PERFORM EVALUATION LOCATION:

ANY APPROXIMATE COMPLETION TIME:

Completion Time: 10 minutes IMPORTANCE RATING(S):

4.1 SYSTEM NUMBER(S):

2.4.41

REFERENCES:

1. ERP-101 Rev.23, CLASSIFICATION OF EMERGENCIES TASK STANDARD(S):

Event is Classified as an Alert per ERP-101 R4. 000 Page 3 of 8 A

a.COC Page 3 of 8

I EXELON NUCLEAR TASK CONDITIONS:

1. Peach Bottom Unit 3 is in Mode 5

2. A breach of the bottom of the spent fuel pool liner has resulted in a loss of level in the reactor cavity and spent fuel pool

3. The spent fuel pool gates have been installed

4. RPV level is 430 inches above instrument zero and rising slowly due to ECCS manual initiation

5. Fuel pool level is at the 222' elevation and still lowering with maximum available emergency makeup injecting INITIATING CUES:

Determine the minimum required Emergency Action Level based on the above conditions Page 4 of 8 A A,.J.

0.UUU

EXELON NUCLEAR Critical Element(s) indicated by in Performance Checklist.

PERFORMANCE CHECKLIST:

STEP STANDARD SAT/UNSAT

1. Obtain ERP-101 ERP-101 obtained

2. Locate Section 1.2, Irradiated or New Section 1.2 located Fuel

3. *Classify the Event per Section 1.2 of EAL determined to be an ALERT ERP-101 (Evaluator copy is attached)

CUE: You have reached the termination point N/A N/A for this JPM. You may stop here.

A A L.....

A'f. 00c Page 5 of 8

.*'..*x A "4

.'

EXELON NUCLEAR *:'.K-

>;*'i.U Comments:

Note: Any grade of UNSAT requires a comment.

JPM Overall Rating:

I SAT/UNSAT A4.doc Page6of8 A4.dOC Page 6 of 8 !~"

TASK CONDITIONS:

1. Peach Bottom Unit 3 is in Mode 5

2. A breach of the bottom of the spent fuel pool liner has resulted in a loss of level in the reactor cavity and spent fuel pool

3. The spent fuel pool gates have been installed

4. RPV level is 430 inches above instrument zero and rising slowly due to ECCS manual initiation

5. Fuel pool level is at the 222' elevation and still lowering with maximum available emergency makeup injecting INITIATING CUES:

Determine the minimum required Emergency Action Level based on the-above conditions EXELON NUCLEAR CANDIDATE

o f--X.LN NUCLEAR 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL EVENT IC Unexpected Rise in Plant Radiation or Airborne Concentration.

1.2.1.a Applicable Modes: ALL Uncontrolled water level drop in the spent fuel pool with all irradiated fuel assemblies remaining covered by water 1.2.1.b Applicable Modes: ALL Unexpected Skimmer Surge Tank low level alarm AND Visual observation of an uncontrolled water level drop below the fuel pool skimmer surge tank inlet IC Unexpected Rise in Plant Radiation 1.2.1.c Applicable Modes: ALL Radiological readings exceed 600 mRlhr one foot away OR 1200 mR/hr at the external surface of any dry storage system ALERT IC Major Damage to Irradiated Fuel, or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel 1.2.2.a Applicable Modes: ALL Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1) 1.2.2.b Applicable Modes: ALL Report of visual observation of irradiated fuel uncovered 1.2.2.c Applicable Modes: 5 (With Reactor Refueling Cavity Flooded)

Water Level < 458" above RPV instrument zero for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering "1.2.2.d Applicable Modes: ALL 1 e Water Level < 232ft 3 inches plant elevation for the Spent Fuell Pool that wil rsl in Irradiated Fuel uncovering SITE AREA EMERGENCY GENERAL EMERGENCY I

None Table 1-1 Refuel Floor ARMs Steam Separator Pool Refuel Slot Fuel Pool Refueling Bridge ANSWER KEY 3-7 (7-9) 3-8 (7-10) 3-9(7-11) 3-10(7-12)

I I

Effective Date:

ERP-101, Rev.

23 Page 1 of 32 RES/res PECO NUCLEAR PEACH BOTTOM UNITS 2 AND 3

EMERGENCY RESPONSE PROCEDURE ERP-101 CLASSIFICATION OF EMERGENCIES 1.0 RESPONSIBILITIES 1.1 Shift Management:

1.1.1 Recognize and classify an event or condition.

1.1.2 Assume duties of Emergency Director (ED).

1.2 Plant Manager or designated alternate:

1.2.1 Relieve acting ED.

1.2.2 Assume duties of ED.

2.0 INITIAL ACTIONS NOTE THE JUDGMENT OF THE EMERGENCY DIRECTOR TAKES PRECEDENCE OVER GUIDANCE IN THE PROCEDURE.

NOTE IDENTIFICATION AND CLASSIFICATION OF EMERGENCIES SHOULD BE ACCOMPLISHED WITHIN 15 MINUTES AFTER THE APPLICABLE EMERGENCY ACTION LEVELS (EALs)

ARE MET.

2.1 Emergency Director shall:

2.1.1 Select categories appropriate for station events or conditions.

2.1.2 Review Emergency Action Level (EALs) for categories selected.

2.1.3 IF the event trigger is known to be spurious, THEN do not classify the event (i.e.,

false high reading, false radiation monitor readings, etc.)

2.1.4 Classify the event based on selected categories and most severe EALs.

2.1.5 IF the event or condition classifies as an emergency, THEN assume duties of ED and implement ERP-200.

ERP-101, Rev. 23Error! Main Document Only.

Page 2 of 32 RES/res 3.0 CONTINUING ACTIONS NOTE IT IS PREFERABLE TO OBTAIN EMERGENCY RESPONSE MANAGER (ERM)

CONCURRENCE PRIOR TO DE-ESCALATION.

3.1 IF emergency conditions dictate, THEN escalate or de escalate emergency classification.

4.0 FINAL CONDITIONS 4.1 Emergency conditions have been terminated, or ERP-C 1900, Recovery Phase Implementation has been implemented.

5.0 ATTACHMENTS AND APPENDICES 5.1 -

EAL Table of Contents and Tables 1 through 9.

CM-i, CM-2, CM-3, CM-5 5.2 - Terms and Definitions 6.0 SUPPORTING INFORMATION 6.1 Purpose 6.1.1 To provide the method for classifying an event or condition into one of four (4) emergency classifications described in the Nuclear Emergency Plan.

6.1.2 To provide pre-determined Protective Action Recommendations (PARs) for specific plant conditions whenever a General Emergency is declared.

6.2 Criteria For Use 6.2.1 Implement whenever conditions meet or exceed EALs listed in the Tables.

NOTE ISSUANCE OF A PAR REQUIRES A GENERAL EMERGENCY CLASSIFICATION AND CONVERSELY A GENERAL EMERGENCY CLASSIFICATION REQUIRES THE ISSUANCE OF A PAR.

6.2.2 PAR information in the tables, is expected to be used when an event rapidly progresses to a General Emergency or when the PAR is based only on plant conditions.

Dose Assessment based PAR information may be obtained from the Dose Assessment Coordinator or the Dose Assessment Team Leader.

In either case, the most conservative PAR available is to be used.

ERP-101, Rev. 23Error! Main Document Only.

Page 3,of 32 RES/res Whenever the Emergency Operations Facility (EOF) is activated, then all PAR information from the ED should be submitted to the ERM.

CM-4 6.3 Special Equipment None 6.4 References 6.4.1 6.4.2 6.4.3 6.4.4 6.4.5 6.4.6 6.4.7 6.4.8 6.4.9 6.4.10 6.4.11 6.4.12 6.4.13 6.4.14 6.4.15 6.4.16 6.4.17 6.4.18 6.2.3 EPA-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents ERP-200, Emergency Director (ED)

ERP-C-1900, Recovery Phase Implementation Nuclear Emergency Plan NUMARC/NESP-007, Methodology for Development of Emergency Action Levels NUREG 0654, FEMA-REP-I, Criteria for Preparations and Evaluation of Radiological Emergency Response Plans in Support of Nuclear Power Plants PBAPS Technical Specifications PBAPS Offsite Dose Calculation Manual PBAPS Updated Final Safety Analysis Report Reference Manual:

Identification and Evaluation of Potentially Reportable Items SE-I, Plant Shutdown from the Remote Shutdown Panel SE-5, Earthquake SE-10, Plant Shutdown from the Alternative Shutdown Panels T-101, Reactor Pressure Vessel Control T-102, Primary Containment Control T-103, Secondary Containment Control T-104, Radioactivity Release Control T-116, RPV Flooding

ERP-101, Rev. 23Error! Main Document Only.

Page 4 of 32 RES/res 6.4.19 T-200, Primary Containment Venting 6.4.20 SO 67.7A, Verification of Suspected Earthquake or Seismic System Activation 6.4.21 US NRC Regulatory Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors 6.4.22 US NRC Response Technical Manual 6.5 Commitment Annotation 6.5.1 CM-I, NRC Inspection Report 50-277, 278/

88 12/12 (T00349),

(see Attachment 1, tables 1 through 9) 6.5.2 CM-2, Event INV Report 3-90-031, corrective action #7, (T00826),

(see Attachment 1, table 1 for Reactor Fuel and table 3 for Fission Product Barrier) 6.5.3 CM-3, NRC URI 85-17-03, IN Inspection Report 86-06/06, (T01934),

(see Attachment 1, table

9) 6.5.4 CM-4, Peach Bottom Inspection Report 92-19/19 (T02540),

(see section 6.2.3) 6.5.5 CM-5, NRC Inspection 92-03/03, (T02541),

(see, table 3 for Fission Product Barrier)

ERP-101, Rev. 23 Page 5 of 32 RES/res EAL Table of Contents 1.0 Reactor Fuel 1.1 Coolant Activity................................................................................................

6 1.2 Irradiated Fuel or New Fuel............................................................................

7 2.0 Reactor Pressure Vessel 2.1 Reactor W ater Level.......................................................................................

8 2.2 Reactor Power...............................................................................................

9 3.0 Fission Product Barrier CM-2, CM-5 3.1 Initiating Condition Matrix...............................................................................

10 3.2 Fission Product Barrier Table........................................................................

11 4.0 Secondary Containment Bypass 4.1 Main Steam Line............................................................................................

13 5.0 Radioactivity Release 5.1 Effluent Release and Dose...........................................................................

14 5.2 In-Plant Radiation..........................................................................................

16 6.0 Loss of Power 6.1 Loss of AC or DC Power...............................................................................

18 7.0 Internal Events 7.1 Technical Specifications & Control Room Evacuation...................................

20 7.2 Loss of Decay Heat Removal Capabililty...........................

21 7.3 Loss of Assessment/Communications Capabililty.........................................

22 8.0 External Events 8.1 Security Events..............................................................................................

24 8.2 Fire/Explosion and Toxic/Flammable Gases.................................................

26 8.3 Man-Made Events..........................................................................................

28 8.4 Natural Events..............................................................................................

29 9.0 Other CM-3 9.1 G e n e ra l..............................................................................................................

3 1 MODE 11 Run 22 Startup 33 Shutdown (hot) 44 Shutdown (cold) 55 Refueling D6 Defueled CM-1, All Tables

ERP-101, Rev. 23 Page 6 of 32 RES/res 1.0 Reactor Fuel 1.1 Coolant Activity CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Fuel Clad Degradation EVENT 1.1.1.a Applicable Modes: ALL Reactor Coolant activity > 4 pCi/gm Dose Equivalent Iodine 131 1.1.1.b Applicable Modes: 1, 2, 3 SJAE Discharge Radiation > 2.5x103 mRlhr ALERT None SITE AREA None EMERGENCY GENERAL None EMERGENCY

ERP-101, Rev. 23 Page 7 of 32 RES/res 1.0 Reactor Fuel 1.2 Irradiated Fuel or New Fuel CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Unexpected Rise in Plant Radiation or Airborne Concentration.

EVENT 1.2.1.a Applicable Modes: ALL Uncontrolled water level drop in the spent fuel pool with all irradiated fuel assemblies remaining covered by water 1.2.1.b Applicable Modes: ALL Unexpected Skimmer Surge Tank low level alarm AND Visual observation of an uncontrolled water level drop below the fuel pool skimmer surge tank inlet IC Unexpected Rise in Plant Radiation 1.2.1.c Applicable Modes: ALL Radiological readings exceed 600 mR/hr one foot away OR 1200 mR/hr at the external surface of any dry storage system

.ALERT IC Major Damage to Irradiated Fuel, or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel 1.2.2.a Applicable Modes: ALL Unplanned general area radiation > 500 mR/hr on the refuel floor (Table 1-1) 1.2.2.b Applicable Modes: ALL Report of visual observation of irradiated fuel uncovered 1.2.2.c Applicable Modes: 5 (With Reactor Refueling Cavity Flooded)

Water Level < 458" above RPV instrument zero for the Reactor Refueling Cavity that will result in Irradiated Fuel uncovering 1.2.2.d Applicable Modes: ALL Water Level < 232ft 3 inches plant elevation for the Spent Fuel Pool that will result in Irradiated Fuel uncovering SITE AREA None EMERGENCY GENERAL None EMERGENCY I

Table 1-1 Refuel Floor ARMs 3-7 (7-9)

Steam Separator Pool 3-8 (7-10)

Refuel Slot 3-9(7-11)

Fuel Pool 3-10(7-12)

Refueling Bridge

2.0 Reactor Pressure Vessel 2.1 Reactor Water Level ERP-101, Rev. 23 Page 8 of 32 RESires CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Reactor Coolant System Leakage EVENT 2.1.1 Applicable Modes: 1, 2, 3, 4 The following conditions exist:

Unidentified Primary System Leakage > 10 gpm into the Drywell OR Identified Primary System Leakage > 25 gpm into the Drywell ALERT None SITE AREA IC Loss of Water Level in the Reactor Vessel That Has or Will Uncover fuel in EMERGENCY the Reactor Vessel 2.1.3 Applicable Modes: 4, 5 RPV level <-172 "

(,MERGENCY None

ý qNEA

2.0 Reactor Pressure Vessel ERP-101, Rev. 23 Page 9 of 32 RES/res 2.2 Reactor Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL None EVENT ALERT IC Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful 2.2.2 Applicable Modes: 1, 2 Automatic RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS SCRAM to make Reactor shutdown SITE AREA IC Failure of Reactor Protection System Instrumentation to Complete or Initiate EMERGENCY an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful 2.2.3 Applicable Modes: 1, 2 RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

IC Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core Applicable Modes: 1, 2 RPS SCRAM should occur due to RPS Setpoint being exceeded AND Failure of Automatic RPS, ARI AND Manual SCRAM to reduce reactor power < 4%

AND Torus Temperature is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-1 02, T/T-1) OR RPV level <-200

- - PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

GENERAL EMERGENCY 2.2.4

3.0 Fission Product Barrier Table ERP-101, Rev. 23 Page 10 of 32 RES/res 3.1 Initiating Condition Matrix USE TABLE 3.2, "FISSION PRODUCT BARRIER STATUS TABLE" FOR CLASSIFYING EVENT CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL 3.1.1 Applicable Modes: 1, 2, 3 EVENT ANY Loss OR ANY Potential Loss of Primary Containment ALERT 3.1.2 Applicable Modes: 1, 2, 3 ANY Loss OR ANY Potential Loss of EITHER Fuel Clad OR RCS SITE AREA 3.1.3 Applicable Modes: 1, 2, 3 Loss of BOTH Fuel Clad AND RCS EMERGENCY O OR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Clad OR RCS, AND Loss of ANY Additional Barrier "GENERAL 3.1.4 Applicable Modes: 1, 2, 3 Loss of ANY Two Barriers AND Potential Loss of Third Barrier

- - PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles. (See Fission Product Barrier Table 3.2 for exception based on extremely Hi Containment Radiation Levels.)

NOTES:

1.

If a "Loss" condition is satisfied, the "Potential Loss" category can be considered satisfied. This is accounted for in the matrix contained in the Fission Product Barrier Table 3.2 used to determine the proper classification based on Fission Product Barrier status.

2.

For all conditions listed in Fission Product Barrier Table 3.2, the barrier failure column is only satisfied if it fails when called upon to mitigate an accident. For example, failure of both containment isolation valves to isolate with a downstream pathway to the environment is only a concern during an accident.

If this condition exists during normal power operations, it will be an active Technical Specification Action Statement. However, during accident conditions, this will represent a breach of containment.

3.2 Fission ProL.

3arrier Status Table Applicable Modes: 1, 2, 3

p-101, Rev. 23 Page 11 of 32 RES/res Reactor Coolant System

rtmary uontainmern Parameter Loss Reactor Coolant Reactor Coolant activity >

Activity 300 pCi/gm Dose Equivalent Iodine 131 RPV Level RPV level <-200" RPV Level Unknown RCS Leak Rate Drywell Pressure Drywell Radiation Drywell Rad Monitor reading

> 8x10' R/hr V level % I RPV level cannot be restored above -200" within the time limit of the "SAFE" region of the Maximum Core Uncovery Time Limit Curve (T-1 16, RF-1)

RPV level cannot be determined RPV level cannot be determined AND RPV Flooding cannot be established as indicated by inability to maintain 5 ADS/SRVs open with RPV pressure at least 60 psig above Torus pressure per T-116 RCS leakage

>50 gpm Drywell Pressure

> 2.0 psg AND Indication of a leak inside drywell Drywell Rad Monitor reading

> 15 R/hr Drywell Pressure

> 49 psig and rising Drywell Hydrogen > 6%

AND Drywell Oxygen > 5%

Evacuate 5 mile radius, evacuate affected sector(s) plus I sector on each side of affected sector(s) for 5-10 miles.

Barrier Fuel Clad OR Drywell Rad Monitor reading

> 6x105 R/hr Primary Containment

r, 3.2 Fission Pr6o

, Barrier Status Table A..,..U

.1.,..

4 "3 '3 Fuel Clad Barrier Parameter Containment Isolation P-101, Rev. 23 Page 12 of 32 RES/res MI..ppical e Modes:I, I

s

£1 I

--A T.-A Or;.n~lr P.ntnfuinmn',n Reactor CooanaP eystelL s

Potential Loss Unisolable primary system leakage outside drywell as indicated by T-103, Temperature Action Level is exceeded in ONE area requiring a SCRAM OR Unisolable primary system leakage outside drywell as indicated by T-1 03, Radiation Action Level is exceeded in ONE area requiring a SCRAM Emergency Director Any condition in the judgment of the Emergency Director Judgment that indicates Loss or Potential Loss of the FUEL CLAD barrier Any condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the RCS barrier

+

Loss Failure of both valves in any one line to close AND downstream pathway to the environment exists OR Intentional venting per T-200 is required OR Unisolable primary system leakage outside drywell as indicated by T-103, Temperature Action Level is exceeded in ONE area requiring a SCRAM OR Unisolable primary system leakage outside drywell as indicated by a T-103, Radiation Action Level is exceeded in ONE area requiring a SCRAM Any condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the Primary In the table below, circle all of the appropriate X's in each applicable row for each Loss or Potential Loss of Fission Product Barrier as determined by the table above.

Classify the event as identified in the table heading if all X's in a column under that heading are circled.

Fission Product Barrier Status Unusual ALERT SITE AREA EMERGENCY GENERAL EMERGENCY Event Fuel Clad - Loss X

X X

Fuel Clad - Potential Loss X

X X

Reactor Coolant System - Loss X

X X

Reactor Coolant System-Potential Loss X

X X

Primary Containment - Loss X

X X

X Primary Containment - Potential Loss X

X

'*'*.PAR*"*.

Evacuate 2 mile radius, evacuate affected sector(s) plus I sector on each side of affected sector(s) for 2-5 miles. (Upgrade PAR for DMW Rad > 6x10 R/hr)

Containment barrier

I III1*1 y vvllL*UUmlWl*ww"

ERP-1 01, Rev. 23Error! Main Document Only.

Page 13 of 32 RES/res 4.0 Secondary Containment Bypass 4.1 Main Steam Line CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Fuel Clad Degradation EVENT 4.1.1 Applicable Modes: 1, 2, 3 Main Steam Line HiHi Radiation (10xNFPB)

ALERT IC RCS Leak Rate 4.1.2 Applicable Modes: 1,.2, 3 Indication of a Main Steam Line Break:

Hi Steam Flow Annunciator AND Hi Steam Tunnel Temperature Annunciator OR Direct report of steam release I:E AREA

,*ERGENCY GENERAL EMERGENCY None None

ERP-1 01, Rev. 23Error! Main Document Only.

Page 14 of 32 RES/res 5.0 Radioactivity Release 5.1 Effluent Release and Dose CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Technical Specifications for 60 Minutes or Longer 5.1.1.a Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds TWO TIMES the HiHi alarm setpoint value for > 60 minutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate using computer dose model exceeds 0.114 mRem/hr TPARD OR 0.342 mRem/hr child thyroid CDE based on a 60 minute average Note: If the required dose projections cannot be completed within the 60 minute period, then the declaration must be made based on the valid sustained monitor reading.

5.1.1.b Applicable Modes: ALL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO TIMES Tech Specs (Liquid Release ODCM 3.8.B.1

-J and Gaseous Release ODCM 3.8.C.1.b) for

> 60 minutes ALERT IC Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Radiological Technical Specifications for 15 Minutes or Longer 5.1.2.a Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds TWO HUNDRED TIMES the HiHi alarm setpoint value for > 15 minutes:

Main Stack, Vent Stack, Radwaste Discharge, Service Water Discharge AND Calculated maximum offsite dose rate exceeds 11.4 mRemlhr TPARD OR 34.2 mRem/hr child thyroid CDE based on a 15 minute average Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

5.1.2.b Applicable Modes: ALL Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates exceeding TWO HUNDRED TIMES Tech Specs (Liquid Release ODCM 3.8.B.1 and Gaseous Release ODCM 3.8.C.l.b) for

> 15 minutes

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Page 15 of 32 RES/res

-E AREA k-L-AlERGENCY GENERAL 1iERGENCY rE:

Cc IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR Whole Body or 500 mR Child Thyroid for the Actual or Projected Duration of the Release 5.1.3 Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:

Main Stack 5.84 I.Ci/cc Vent Stack 2.08E-3 p.Ci/cc Torus Vent 203 cpm Note:

If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

OR Projected offsite dose using computer dose model exceeds 100 mRem TPARD OR 500 mRem child thyroid CDE OR Analysis of Field Survey results indicate site boundary whole-body dose rate exceeds 100 mRemlhr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 500 mRem for one hour of inhalation IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity that Exceeds 1000 mR Whole Body or 5000 mR Child Thyroid for the Actual or Projected Duration of the Release Using Actual Meteorology 5.1.4 Applicable Modes: ALL A valid reading on one or more of the following radiation monitors that exceeds or is expected to exceed the value shown for > 15 minutes AND Dose Projections are not available:

Main Stack 58.4 ýiCi/cc Vent Stack 2.08E-2 0Ci/cc Torus Vent 2000 cpm Note: If the required dose projections cannot be completed within the 15 minute period, then the declaration must be made based on the valid sustained monitor reading.

OR Projected offsite dose using computer dose model exceeds 1000 mRem TPARD OR 5000 mRem child thyroid CDE OR Analysis of Field Survey results indicate site boundary whole body dose rate exceeds 1000 mRem/hr expected to continue for more than one hour, OR Analysis of Field Survey results indicate child thyroid dose commitment of 5000 mRem for one hour of inhalation

- - PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

E = Committed Dose Equivalent, TPARD = Total Protective Action Recommendation Dose

ERP-101, Rev. 23Error! Main Document Only.

Page 16 of 32 RES/res 5.0 Radioactivity Release 5.2 In-Plant Radiation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Unexpected Rise in Plant Radiation or Airborne Concentration EVENT 5.2.1 Applicable Modes: ALL Valid Direct Area Radiation Monitor readings rise by a factor of 1000 over normal*

levels Normal levels can be considered as the highest reading in the past twenty four hours excluding the current peak value.

ALERT IC Release of Radioactive Material or Rises in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown 5.2.2.a Applicable Modes: ALL Valid radiation level readings > 5000 mRPhr in areas requiring infrequent access to maintain plant safety functions as identified in procedure SE-1, SE-10 AND Access is required for safe plant operation, but is impeded, due to radiation dose "rates 5.2.2.b Applicable Modes: ALL Valid Control Room OR Central Alarm Station radiation reading > 15 mR/hr SITE AREA None EMERGENCY GENERAL None EMERGENCY

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Page 18 of 32 RES/res 6.0 Loss of Power 6.1 Loss of AC or DC Power CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes EVENT 6.1.1.a Applicable Modes: ALL The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND At least Two Diesel Generators are supplying power to their respective 4 KV emergency busses IC Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode for Greater than 15 Minutes 6.1.1.b Applicable Modes: 4, 5 Unplanned Loss of ALL safety related DC Power indicated by

< 107.5 VDC on DC Panels 2(3)0D21, 22, 23, 24 for >15 minutes LERT IC AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout 6.1.2.a Applicable Modes: 1, 2, 3 The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer for >15 minutes AND Only One 4 KV emergency bus powered from a Single Onsite Power Source due to the Loss of: Three of Four Division Diesel Generators, D/G Output Breakers, or 4 KV Emergency Busses as indicated by bus voltage IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode 6.1.2.b Applicable Modes: 4, 5, D The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power

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-"E AREA IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential K.tERGENCY Busses 6.1.3.a Applicable Modes: 1, 2, 3 The following conditions exist:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure to restore power to at least One 4 KV emergency bus within 15 minutes from the time of loss of both offsite and onsite AC IC Loss of All Vital DC Power 6.1.3.b Applicable Modes: 1,. 2, 3 Loss of ALL Safety Related DC Power indicated by < 107.5 VDC on DC Panels 2(3)0D21, 22, 23, 24 for > 15 minutes GENERAL IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC EMERGENCY Power 6.1.4 Applicable Modes: 1, 2, 3 Prolonged loss of all offsite and onsite AC power as indicated by:

Loss of Power to 2 and 3 Startup and Emergency Aux. Transformers and 343 Startup Transformer AND Failure of ALL Emergency Diesel Generators to supply power to 4 KV emergency busses AND At least one of the following conditions exist:

"* Restoration of at least One emergency bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is NOT likely OR

"* Reactor Water Level cannot be maintained > -172k OR

"* Torus temperature is on the "UNSAFE" side of the Heat Capacity Temperature Limit (HCTL) curve (T-102, T/T-1)

- - PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

ERP-1 01, Rev. 23Error! Main Document Only.

Page 20 of 32 RES/res 7.0 Internal Events 7.1 Technical Specification & Control Room Evacuation CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Inability to Reach Required Shutdown Mode Within Technical Specification EVENT Limits 7.1.1 Applicable Modes: 1, 2, 3 Inability to reach required shutdown mode within Tech. Spec. LCO required action completion time.

ALERT IC Control Room Evacuation Has Been Initiated 7.1.2 Applicable Modes: ALL Entry into SE-1 or SE-10 procedure for Control Room evacuation SITE AREA IC Control Room Evacuation Has Been Initiated and Plant Control Cannot Be EMERGENCY Established 7.1.3 Applicable Modes: ALL The following conditions exist:

Control room evacuation has been initiated AND Control of the plant cannot be established per SE-I or SE-10 within 15 minutes GENERAL None EMERGENCY

ER P-1 01, Rev. 23Error! Main Document Only.

Page 21 of 32 RES/res 7.0 Internal Events 7.2 Loss of Decay Heat Removal Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL None EVENT IC Inability to Maintain Plant in Cold Shutdown 7.2.2 Applicable Modes: 4, 5 The following conditions exist:

Unplanned Loss of ALL Tech Spec required systems available to provide Decay Heat Removal functions AND Uncontrolled Temperature rise that either:

Exceeds 212 OF (Excluding a <15 minute rise >2120 F with a heat removal function restored)

OR Results in temperature rise approaching 212 *F (with NO heat removal function restored)

SITE AREA IC Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EMERGENCY 7.2.3 Applicable Modes: 1, 2, 3 Loss of TORUS heat sink capabilities as evidenced by T-1 02 T/T legs directing a T 112 Emergency Blowdown GENERAL None EMERGENCY ALERT

ERP-101, Rev. 23Error! Main Document Only.

Page 22 of 32 RES/res 7.0 Internal Events 7.3 Loss of Assessment / Communication Capability CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Unplanned Loss of Most or All Safety System Annunciation or Indication in EVENT The Control Room for Greater Than 15 Minutes 7.3.1.a Applicable Modes: 1, 2, 3 Unplanned loss of most or all safety system annunciators (Table 7-1) OR indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit(s).

IC Unplanned Loss of All Onsite or Offsite Communications Capabilities 7.3.1.b Applicable Modes: ALL Loss of ALL Onsite communications (Table 7-3) affecting the ability to perform routine operations OR Loss of ALL Offsite communications (Table 7-3)

.ERT IC Unplanned Loss of Most or All Safety System Annunciation or Indication In Control Room With Either (1) a Significant Transient in Progress, or (2)

Compensatory Non-Alarming Indicators are Unavailable 7.3.2 Applicable Modes: 1, 2, 3 Unplanned loss of most or all safety system annunciators (Table 7-1) OR indicators (Table 7-2) for > 15 minutes requiring increased surveillance to safely operate the unit(s)

AND EITHER A significant plant transient is in progress (Table 7-4) OR the plant monitoring system (PMS) is unavailable.

SITE AREA IC Inability to Monitor a Significant Transient in Progress EMERGENCY 7.3.3 Applicable Modes: 1, 2, 3 Loss of safety system annunciators (Table 7-1)

AND indicators (Table 7-2)

AND PMS AND a significant plant transient is in progress. (Table 7-4)

`-ENERAL

..AERGENCY

1.

1 None

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Page 23 of 32 RES/res

(

Table 7-1 Safety System Annunciators ECCS Containment Isolation Reactor Trip Process Radiation Monitoring Table 7-2 Safety Function Indicators Reactor Power Decay Heat Removal Containment Safety Functions Table 7-3 Communications Onsite Offsite Site Phones (GTE System)

X X

OMNI System X

X Plant Public Address X

Station Radio X

NRC (FTS-2000)

X PA State Police Radio X

Load Dispatcher Radio X

PECO Dial Network X

Table 7-4 Significant Plant Transients SCRAM Recirc Runbacks > 25% thermal power Sustained power oscillations 25% peak to peak Stuck open relief valve(s)

ECCS injection

ERP-1 01, Rev. 23Error! Main Document Only.

Page 24 of 32 RES/res 8.0 External Events 8.1 Security Threats CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Confirmed Security Event Which Indicates a Potential Degradation in the EVENT Level of Safety of the Plant 8.1.1 Applicable Modes: ALL A credible threat to the station reported by the NRC.

OR An actual threat that meets ALL of the following criteria:

"* A credible threat reported by any other outside agency or determined per SY-AA-101-132; AND

"* Is specifically directed towards the station; AND

"* Is imminent (within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

OR Attempted intrusion and attack to the Protected Areas OR Attempted sabotage discovered within the Protected Areas OR "Hostage/Extortion situation that threatens normal plant operations ALERT IC Security Event in a Plant Protected Area 8.1.2 Applicable Modes: ALL Intrusion into plant protected areas by a hostile force OR Confirmed bomb, sabotage or sabotage device discovered in the Protected Areas SITE AREA IC Security Event in a Plant Vital Area EMERGENCY 8.1.3 Applicable Modes: ALL Intrusion into plant Vital area by a hostile force OR Confirmed bomb, sabotage or sabotage device discovered in a Vital Area GENERAL IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold EMERGENCY Shutdown 8.1.4 Applicable Modes: ALL Loss of physical control of the control room due to security event OR Loss of physical control of all remote shutdown capability due to security event

- - PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

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Page 26 of 32 RES/res 8.0 External Events 8.2 Fire / Explosion and Toxic / Flammable Gases CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Fire Within Protected Area Boundary Not Extinguished Within 15 Minutes of EVENT Detection 8.2.1.a Applicable Modes: ALL Fire within ON-1 14 Plant Vital Structures (Table 8-1) which is not extinguished within 15 minutes of control room notification or verification of a control room alarm IC Release of Toxic or Flammable Gasses Deemed Detrimental to Safe Operation of the Plant 8.2.1.b Applicable Modes: ALL Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant OR Report by Local, County or State Officials for potential evacuation of site personnel based on offsite event "IC Natural and Destructive Phenomena Affecting the Protected Area 8.2.1.c Applicable Modes: ALL Report by plant personnel of an unanticipated explosion within protected area boundary resulting in visible damage to permanent structure or equipment ALERT IC Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown 8.2.2.a Applicable Modes: ALL The following conditions exist:

Fire or explosion which potentially makes inoperable:

Two or More subsystems of a Safe Shutdown System (Table 8-2) OR Two or More Safe Shutdown Systems OR Plant Vital Structures containing Safe Shutdown Equipment AND Safe Shutdown System or Plant Vital Structure is required for the present Operational Mode 11 1

I 'C ERP-101, Rev. 23Error! Main Document Only.

Page 27 of 32 RES/res

ERT Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Applicable Modes: ALL 8.2.2.b Report or detection of toxic gases within Plant Vital Structures (Table 8-1) in concentrations that will be life threatening to plant personnel OR Report or detection of flammable gases within Plant Vital Structures (Table 8-1) in concentrations affecting the safe operation of the plant SITE AREA None EMERGENCY GENERAL None EMERGENCY I

Table 8-1 Plant Vital Structures Power Block Diesel Generator Building r

Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower Table 8-2 Safe Shutdown Systems Diesel Generators HPCI Core Spray SBGTS PCIS 4KV Safeguard Buses RCIC HPSW ECW Control Room Ventilation ADS RHR (All Modes)

ESW CAC/CAD

ERP-101, Rev. 23Errorl Main Document Only.

Page 28 of 32 RES/res 8.0 External Events 8.3 Man-Made Events Table 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Destructive Phenomena Affecting the Protected Area EVENT 8.3.1.a Applicable Modes: ALL Vehicle crash within protected area boundary that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.

8.3.1.b Applicable Modes: ALL Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

ALERT IC Destructive Phenomena Affecting the Plant Vital Area 8.3.2 Applicable Modes: ALL Vehicle crash affecting Plant Vital Structures (Table 8-1)

OR Turbine failure generated missiles result in any visible structural damage to or penetration of any Plant Vital Structures (Table 8-1)

SITE AREA None EMERGENCY GENERAL None EMERGENCY

ERP-101, Rev. 23Error! Main Document Only.

Page 29 of 32 RES/res 8.0 External Events 8.4 Natural Events CLASSIFICATION [ EMERGENCY ACTION LEVEL UNUSUAL EVENT ILERT SITE AREA EMERGENCY

_NERAL "MERGENCY IC Natural and Destructive Phenomena Affecting the Protected Area 8.4.1.a Applicable Modes: ALL Earthquake >.01 g as determined by procedure SO 67.7.A 8.4.1.b Applicable Modes: ALL Report by plant personnel of tornado striking within protected areas OR Wind speeds > 75 mph as indicated on site Meteorological data for > 15 minutes 8.4.1.c Applicable Modes: ALL Assessment by the control room that an event has occurred. (Natural and Destructive Phenomena Affecting the Protected Areas) 8.4.1.d Applicable Modes: All High River level > 112' OR Low River level < 98.5' IC Natural and Destructive Phenomena Affecting the Plant Vital Area 8.4.2.a Applicable Modes: ALL Earthquake >.05 g (Operating Basis Earthquake OBE) as determined by procedure SO 67.7.A 8.4.2.b Applicable Modes: ALL Tornado or wind speeds > 75 mph causing damage to Plant Vital Structures (Table 8-1) 8.4.2.c Applicable Modes: ALL Report of any visible structural damage to any Plant Vital Structure (Table 8-1) 8.4.2.d Applicable Modes: All High River level > 116' OR Low River level < 92.5' None None

ERP-1 01, Rev. 23Error! Main Document Only.

Page 30 of 32 RES/res ie 8-1 Plant Vital Structures Power Block Diesel Generator Building Emergency Pump Structure Inner Screen Structure Emergency Cooling Tower

ERP-101, Rev. 23Error! Main Document Only.

Page 31 of 32 I RESires 9.0 Other 9.1 General CLASSIFICATION EMERGENCY ACTION LEVEL UNUSUAL IC Other Conditions Existing Which in the Judgment of the Emergency Director EVENT Warrant Declaration of an Unusual Event 9.1.1 Applicable Modes: ALL Other conditions exist which in the judgment of the Emergency Director indicate a potential degradation of the level of safety of the plant ALERT IC Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert 9.1.2 Applicable Modes: ALL Other conditions exist which in the Judgment of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted

,,E AREA IC Other Conditions Existing Which in the Judgment of the Emergency Director ERGENCY Warrant Declaration of Site Area Emergency 9.1.3 Applicable Modes: ALL Other conditions exist which in the Judgment of the Emergency Director indicate actual or likely major failures of plant functions needed for protection of the public GENERAL IC Other Conditions Existing Which in the Judgment of the Emergency Director EMERGENCY Warrant Declaration of General Emergency 9.1.4 Applicable Modes: ALL Other conditions exist which in the Judgment of the Emergency Director indicate:

(1) actual or imminent substantial core degradation with potential for loss of containment, or (2) potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary

- - PAR***

Evacuate 2 mile radius, evacuate affected sector(s) plus 1 sector on each side of affected sector(s) for 2-5 miles.

Page 32 of 32 RES/res TERMS AND DEFINIT Events in progress or have occurred, that indicate a potential degradation of the level of safety of the plant.

No releases of radioactive material requiring off-site response or monitoring are expected unless further degradation of safety systems occurs.

Events in progress or have occurred that involve actual or potential substantial degradation of the level of safety of the plant.

Any releases of radioactive material are expected to be limited to small fractions of the Environmental Protective Agency (EPA)

Protective Action Guidelines (PAG) exposure levels.

Events in progress or which have occurred that involve actual or likely major failures of plant functions needed for protection of the public.

Any releases of radioactive material are not expected to exceed EPA PAG exposure levels except near site boundary.

Events in progress or which have occurred that involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity.

Releases of radioactive material can be reasonably expected to exceed EPA PAG exposure levels off-site for more than the immediate site or

ML021700671

Text

1.5 REFERENCES

REFERENCES:

REFERENCES:

Title:

Reference:

2.0 REFERENCES

REFERENCES:

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