Text

Initial Exam 2008-302 Draft Administrative JPMs
	ML091810066
Person / Time
Site:	Browns Ferry
Issue date:	06/30/2009
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-259/08-302, 50-260/08-302, 50-296/08-302, ES-301, ES-301-1
	Download: ML091810066 (181)
	v • d • e

") ES-301 Administrative Topics Outline Form ES-301-1 I Facility: BFN (NRC EXAM) Date of Examination: 12/08/08 Examination Level (circle one): RO/SRO Operating Test Number HLT0707 Administrative Topic Type Describe Activity to be performed (see Note) Code*

A. Conduct of Operations N Demonstrate correct method for Independent (550) (Generic 2.1.29) Verification (SPP-IO.3r1) (RO/SRO)

(Admin A)

B. Conduct of Operations M Overtime Eligibility (540) (Generic 2.1.5) (OPDP-l,OSIL 25, SPP-1.5) (RO/SRO)

(Admin B)

C. Equipment Control D Evaluate Recombiner Efficiency (3-01-66 r56)

(510) (Generic 2.2.44) (RO/SRO)

(Admin C)

D. Radiation Control D Determine Dose Limitations for Pregnant (511) (Generic 2.3.4) Employee (SPP-5.1 r6) (RO/SRO)

(Admin D)

E. Emergency Plan N/S Classify the Event per the REP (Loss of all Pwr to (480TC) (Generic 2.4.41) 4kv SID bds > 3 hrs) (SRO Only)

(Admin E) (EPIP-l r 431 EPIP-5 r37)

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

Type Codes & Criteria: (C)ontrol room (D)irect from bank (~3 for ROs; ~ 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (2: 1)

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

(S)imulator

) ES-301, Page 22 of27

PAGE 1 of6

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JPM NUMBER: 550 TITLE: DETERMINE CORRECT METHOD OF VERIFICATION ON A GIVEN SYSTEM ADMIN: Conduct of Operations PROVIDE CANDIDATE WITH A COPY OF: SPP-10.3 (If requested)

SUBMITTED BY: DATE: _ _ __

VALIDATED BY: DATE: _ _ __

APPROVED BY: DATE: _ _ __

TRAINING PLANTCONCURRENCE: _ _ _ _ _ _ _ _ _ ____ DATE: _ _ __

OPERATIONS

Examination JPMs Require Operations Training Manager Approval or Designee Approval and Plant Concurrence

)

PAGE 2 of6 REVISION LOG Revision Effective Pages Description Number Date Affected Of Revision a 08/07/08 All Initial issue

PAGE 30f6 OPERATOR: _____________________________________________

RO SRO DATE: ____________

JPM NUMBER: 550 TASK TITLE: DETERMINE CORRECT METHOD OF VERIFICATION ON A GIVEN SYSTEM KIA NUMBER: 2.1.29 KIA RATING: RO 4.1 SRO 4.0 TASK STANDARD: DETERMINE CORRECT METHOD OF VERIFICATION ON A GIVEN SYSTEM PERFORMANCE LOCATION: CLASSROOM: X REFERENCES/PROCEDURES NEEDED: SPP-10.3, Rev 1 VALIDATION TIME: CLASS ROOM: _____

MAX. TIME ALLOWED: _ _ _ _ _ (FOR TIME CRITICAL JPMs ONLY)

PERFORMANCE TIME:

COMMENTS:

ADDITIONAL COMMENT SHEETS ATTACHED? YES ___ NO RESULTS: SATISFACTORY UNSATISFACTORY EXAMINER SIGNATURE: DATE: ___________

)

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PAGE 4 orB Examiner Key A. Open Manual valve in a high Rad area - exposure rate 6 Rlhr.

(2) Perform Verification by alternate means, i.e. flow indication downstream of valve B. Closed Manual valve.

(5) Turn the valve handwheel in the Closed direction and verify the valve stem does Not move C. Locked Manual valve Throttled 3 turns Open.

(6) This valve cannot be Independently Verified (The 1st and 2 nd verification already performed was adequate)

D. Locked Closed Manual valve.

(8) Verify the chain is in place and the locking mechanism is intact E. Open Manual valve.

(4) Turn the valve handwheel in the Closed direction and verify the valve stem moves and return the valve to the Open position

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PAGE 5 Orl>

Candidate Handout A valve checklist was performed on system X. All valves were 1st and 2 nd party verified.

When the checklist was complete, the system was placed in service. Then the US decided that an Independent Verification should have been performed on some of the valves. Select the proper method for Independent Verification of the following valves.

(Assume each valve requires 2 ~ minutes to verify).

A. Open Manual valve in a high Rad area- 1. Perform Verification by alternate means, i.e. Red exposure rate 6 R/hr. light illuminated and Green light extinguished on B. Closed Manual valve. Control Room panel.

C. Locked Manual valve Throttled 3 turns Open. 2. Perform Verification by alternate means, i.e. flow D. Locked Closed Manual valve. indication downstream of valve.

E. Open Manual valve. 3. Turn the valve handwheel in the Open direction and verify valve stem does Not move.

4. Turn the valve handwheel in the Closed direction and verify the valve stem moves and return the A. For the valves (A - E), valve to the Open position.

Enter a number (1 - 8) that 5. Turn the valve handwheel in the Closed direction corresponds to the correct and verify the valve stem does Not move.

B. verification process used to verify the valve position 6. This valve cannot be Independently Verified (The and place it in the space 1sl and 2 nd verification already performed was C. provided. (The numbers to adequate).

the right may be used more than once or not at all). 7. Turn the valve handwheel in the Open direction D. and verify the valve stem moves and return the valve to the Closed position.

8. Verify the chain is in place and the locking E. mechanism is intact.

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Candidate Handout A valve checklist was performed on system X. All valves were 1st and 2 nd party verified.

When the checklist was complete, the system was placed in service. Then the US decided that an Independent Verification should have been performed on some of the valves. Select the proper method for Independent Verification of the following valves.

(Assume each valve requires 2 % minutes to verify).

A. Open Manual valve in a high Rad area - 1. Perform Verification by alternate means, i.e. Red exposure rate 6 Rlhr. light illuminated and Green light extinguished on B. Closed Manual valve. Control Room panel.

C. Locked Manual valve Throttled 3 turns Open. 2. Perform Verification by alternate means, i.e. flow D. Locked Closed Manual valve. indication downstream of valve.

) E. Open Manual valve. 3. Turn the valve handwheel in the Open direction and verify valve stem does Not move.

4. Turn the valve handwheel in the Closed direction and verify the valve stem moves and return the A. For the valves (A - E).

valve to the Open position.

Enter a number (1 - 8) that 5. Turn the valve handwheel in the Closed direction corresponds to the correct and verify the valve stem does Not move.

B. verification process used to verify the valve position 6. This valve cannot be Independently Verified (The and place it in the space 1st and 2nd verification already performed was C. provided. (The numbers to adequate).

the right may be used more than once or not at all). 7. Turn the valve handwheel in the Open direction D. and verify the valve stem moves and return the valve to the Closed position.

8. Verify the chain is in place and the locking E. mechanism is intact.

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Tennessee TITLE SPP-10.3 Valley Authority Rev. 1 Page 1 of 16 TVAN STANDARD VERIFICATION PROGRAM Quality Related ~Yes DNo PROGRAMS AND PORe Required ~Yes o No PROCESSES 10CFR50.59 Review DYes ~ No Effective Date 11/14/2003 RESPONSIBLE PEER TEAM: O~erations Organization

)

CONCURRENCES J. L. Lewis 7-15-03

Primary Sponsor Date WRL 9/16/03 8/11/03 W. R. Lagergren 7-23-03 Peer Team Mentor Date APPROVAL For Nuclear Assurance Sponsored SPPs NIA General Manager, NA Date Karl W. Singer 9/16/03

Senior Vice President, Nuclear Operations Date

Site-specific_changes are approved by Site Sponsor and Site Vice President (see PCF)

TVA 40480 [05-2000] Page 1 of 1 SPP-2.1-1 [05-30-2000]

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1 PROCESSES Page 2 of 16 REVISION LOG Revision Effective Pages Description Number Date Affected of Revision o 6/30/99 All Initial issue. This procedure replaces STO-12.6 (Corp.),

(COC & WBN) SSP-12.6 (BFN & SON), and SSP-12.06 (WBN).

7/2199 BFN YSO LATER SQN 813199 8/6/99 11114/03 2-7, 9-16 Revised to remove verification requirements for placing and removing clearances which have been incorporated into SPP-10.2. Added Section 3.5, and Subsections 3.5.1, 3.5.2, and 3.5.3. Added definitions to Section 5.0 for Peer-Checking and Self-Checking. Revised Appendix A to remove N/A from System 41 and System 84 (systems require verification). Removed SQN only requirement on Section 3.3.1.E.

}

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

') PROCESSES Page 3 of 16 TABLE OF CONTENTS Section Title Page Revision Log .........................................................................................................................................2 Table of Contents ................................................................................................................................. 3 1.0 PURPOSE ...............................................................................................................................4 2.0 SCOPE ....................................................................................................................................4 3.0 INSTRUCTIONS ......................................................................................................................4 3.1 Responsibilities .......................................................................................................................4 3.2 Qualifications ..........................................................................................................................5 3.3 Verification Techniques ........................................................................................................... 6 3.4 Verification Requirements ........................................................................................................ 8 3.5 Human Error Prevention ......................................................................................................... 11 4.0 RECORDS ............................................................................................................................. 13 4.1 QA Records ........................................................................................................................... 13

) 4.2 Non-QA Records .................................................................................................................... 13 5.0 DEFINITIONS ........................................................................................................................ 14 APPENDIXES Appendix A - Systems and Components Requiring Independent or Second-Party Verification .......................................................................... 15

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1 PROCESSES Page 4 of 16 1.0 PURPOSE This procedure establishes the requirements for and the criteria used to determine the applicable verification method for configuration control. The methods of verification are independent (IV) and second-party verification. For the purposes of this procedure and associated procedureslinstructions, the term second-party verification is considered synonymous with concurrent verification (CV).

2.0 SCOPE This procedure applies to all TVA Nuclear (TVAN) personnel and contractors performing activities affecting nuclear power plant systems. Self-checking techniques should be utilized to ensure the worker positively identifies the correct unit, train, and/or component, and reviews the intended action and expected response before performing the task.

This procedure does not apply to activities performed by the Quality organization or design verification activities.

Verifications required in association with clearance activities are performed in accordance with SPP-10.2, Clearance Program.

3.0 INSTRUCTIONS 3.1 Responsibilities Operations Manager A. The Operations Manager is responsible for the following:

1. Determining the verification method required and designating those systems and/or components requiring IV or second-party verification.

Appendix A provides the list of systems and components requiring IV or second-party verification.

2. Resolving disagreements between plant sections and making the final decision regarding the method of verification required.

Responsible Manager B. The responsible manager is responsible for the following:

1. Designating IV or second-party verification requirements in appropriate plant procedures/instructions and work documents.

2. Ensuring that plant procedures/instructions and work documents specify IV or second-party verification when required.

3. Ensuring that personnel assigned to perform IV and second-party verifications are qualified.

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1 PROCESSES Page 5 of 16 Procedure Preparers C. The preparers of site procedureslinstructions and work documents are responsible for the following:

1. Ensuring that IV or second-party verification requirements are specified as appropriate.

2. Ensuring the type of verification is clearly identified.

D. Shift Manager The shift manager (SM) shall be responsible for the following:

1. Determining the corrective actions to be taken when discrepancies are discovered.

2. Ensuring that personnel assigned to perform IV and second-party verification are qualified.

3. Authorizing deviations from normal verification practices if needed.

E. Training Manager Develop, conduct, and document training of personnel engaged in verification activities.

F. All Personnel Inform their respective foreman or supervisor if they have been assigned a verification which they do not feel qualified to perform. In the event their respective supervisor is not available, they will contact the SM for resolution before continuing the verification.

3.2 Qualifications Individuals assigned IV or second-party verification responsibilities shall meet the following qualification requirements:

A. Technically qualified to perform the assigned task (experience, position description, familiarity with the task, etc., should be considered) as determined by the responsible manager.

B. Completed training on verification program requirements.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

'\

) PROCESSES Page 6 of 16 3.3 Verification Techniques Second Party (Concurrent) Verification Standard:

1. Using 3-way communication, performer and peer agree on the action to take, on which component, and for what purpose, confirmed by the guiding document.

2. Using self-checking, the performer and verifier individually confirm the correct component, label, etc. Flag the component if desired.

3. Performer performs predetermined action and only that action.

4. Verifier watches the actions of the performer to verify the actions are correct.

5. Verifier, at the moment of performing the action and without being influenced by the performer, confirms the actions of the performer are correct, and ascertains the proper configuration matches required condition after action is performed using one or more of the following means:

Hands-on verification that configuration is correct (e.g., checking valve position)

Observing remote indication

Observing correct system/equipment/component response

6. Performer and verifier confirm the new configuration agrees with the guiding document and signs the appropriate spaces provided in the procedure.

Independent Verification Standard:

NOTE In the true meaning of Independent Verification, the performer and the verifier may receive the pre-job brief together but not be associated together for the activity. In trying to keep the integrity of the verification, the verifier cannot rely on any visual/audible ques of the performer. The object of this verification process is to not pollute the verification with any information from the performer.

1. Performer self-checks the component to be manipulated.

2. Performer performs predetermined action and only that action.

3. Performer confirms new configuration agrees with guiding document and signs his/her signature in the spaces provided in guiding document.

4. At a separate time and not in the presence of the performer, the verifier self-checks the component that was manipulated to verify component identification matches the component required to be verified.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 7 of 16

5. Verifier verifies the as-found configuration or condition matches the required position, without changing it, using one or more of the following means:

Hands-on verification that configuration is correct (e.g., manually checking valve position)

Observing remote indication

Observing correct system/equipment/component response

6. Verifier confirms new configuration or condition agrees with guiding document and signs his/her signature in the spaces provided in guiding document.

7. If as-found configuration or condition is incorrect, report the condition to supervision immediately.

3.3.1 Valves A. Valves that are to be verified open will be manipulated in the closed direction only as necessary to remove any slack from the operating mechanism and verify valve stem movement. The valve will then be fully opened, subject to normal precautions on backseating valves.

B. Valves that are to be verified closed will be manipulated in the closed

) direction only as necessary to verify the valve is fully closed, and not binding or difficult to operate. Care must be exercised, however, to avoid overtorquing the valve operator and damaging the valve seat. If any doubt exists, SM should be contacted for resolution.

C. To determine the position of a throttled valve, the total number of turns until the handwheel stops moving in the open/closed direction shall be counted. To set the position of a throttled valve, open/close the valve the required number of turns from the full closed/open position (handwheel will no longer move in the closed/open direction).

D. Reach rod valve position indicators will not be used as the sole method of position verification.

E. Locked valve and throttled valve position cannot be independently verified since these operations require the verifier to observe actions while they are being performed. Second-party verification shall be used to verify the position of locked and throttled valves in those cases where IV would normally be required.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 8 of 16 3.3.2 Alternate Verification Techniques Alternate verification techniques may be used by the verifier where specified by approved procedures, valve and breaker line-up checklists, or at the discretion of shift supervisory personnel. Examples include the following:

A. Use of remote position indicators. (Indicating lights in the control room, at the switchgear, or at local controls are the normal method of determining motor-operated and air-operated valve position.)

B. Use of process parameters (e.g., pressure, flow, vibration, current, voltage, potential lamps, etc.).

C. Observation of the valve stem to aid in determination of valve position if the valve stem is marked by paint (when fully closed) or other positive verification methods.

D. Authorized scribe marks on valve stems, properly labeled with the throttled position.

E. Functional mechanical position indicators.

F. A post maintenance/modification functional test provided the testing verifies each component under consideration.

) 3.3.3 Circuit Breakers Circuit breaker verification will include a local inspection of the breaker, control power switches or fuses, and other equipment as outlined below:

A. To verify a breaker is removed from service, the independent or second-party verifier will ensure control power is isolated (if required) by inspecting appropriate switches, fuses or fuse blocks, and ensure the breaker is racked out to the disconnected position, as applicable.

B. To verify a breaker is restored to service, the independent or second-party verifier will ensure control power is energized by inspecting appropriate switches, indicating lights, fuses or fuse blocks, and will ensure the breaker is fully racked in with closing springs charged as applicable. Where practical, the end device should be operated following the reinstallation of a breaker. The verifier will also ensure the cubicle door is in good condition with all fasteners tight.

3.4 Verification Requirements When determination of these requirements is not clear, the responsible manager will designate the requirements. If there is disagreement, the operations manager will designate the requirements.

3.4.1 IV or second-party verification is required for those systems listed in Appendix A and shall include the following as a minimum:

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 9 of 16 A. All valves, breakers, and other components in safety-related systems where an inappropriate positioning could adversely affect system/plant operation or containment integrity.

B. All valves, breakers, and other components in fire protection system major flow paths, including fire fighting water supply and storage, carbon dioxide storage systems, fire protection systems, and all components necessary for the system to function and supply extinguishing media to the fire.

C. All valves, breakers, and other components in gaseous and liquid radioactive waste handling and processing systems where an inappropriate positioning could result in radioactive material release to the environment.

3.4.2 Activities Exempt From Independent and Second-Party Verification Requirements A. Calculations performed by qualified computer software.

B. Activities for which verifications would be required and one or more of the following conditions exist:

Out-of-service systems/channels/components for which

) configuration control will not be maintained and will be verified to be in the proper configuration during the return to operable status.

Activities involving significant radiation exposure. As a guideline, an exposure greater than 10 mrem TEDE to perform the verification would be considered excessive.

Activities occurring during emergency conditions (imminent danger to plant or personnel) requiring rapid personnel action.

Activities that could jeopardize personnel safety.

Components located within locked/covered/controlled access areas provided access to the area has not occurred since the last documented verification.

For these instances, the decision not to perform a verification is to be documented on the procedurelinstruction or work document.

3.4.3 Independent Verification Requirements IV is used to confirm that an activity or condition has been implemented in conformance with specified requirements. The individual performing the IV must physically check the condition without relying on observation or verbal confirmation by the initial performer. However, the independent verifier may be involved in unrelated portions of the same activity. IV is required for the following:

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 10 of 16 A. Any critical activity that, if done improperly, could remain undetected until that structure, system, or component was called upon to mitigate an accident or transient as described in the FSAR, Fire Protection Plan, Security Plan, or ODCM. Critical implies the activity is absolutely necessary for Systems, Structures, and Components to function.

B. Initial system lineups, or restoring components to their required position/condition following an outage where the system status was not maintained.

C. Normal system line-up periodic checks conducted during operating conditions. In this case, the individual performing the periodic check of the original lineup is considered to be the independent verifier and an additional second check is not required. IV of locked components consists of checking that required locking devices are present and intact.

D. Installation and removal of temporary alterations covered by the TACF Program.

3.4.4 Second-Party Verification Requirements Second-party verification is used in lieu of IV for the activities listed below.

When performing a second-party verification, an agreement must be reached between the performer and the verifier that the activity/manipulation to be

) performed is correct before performance.

A. Activities where performing an IV would by itself invalidate the actions or conditions the performer is attempting to establish.

EXAMPLE Verification of throttled valve position, locked valve position, installation and removal of high voltage line or bus PT fuses, installation and removal of fuses in fuse blocks/clips which are normally hidden from view, etc.

B. Activities which, if improperly accomplished or incorrectly identified, may cause any of the following:

Immediate plant trip or transient

Safety system actuation

Start of equipment

Equipment failure/damage

Release of radioactive material

Personnel injury EXAMPLE Removal or installation of wires, jumpers, or other connections; valve, switch, or breaker manipulations; removal or installation of fuses or circuit cards; etc.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 11 of 16 3.5 Human Error Prevention Self-checking helps prevent errors when 'touching' plant equipment to change its status or even when revising a plant document important for plant safety and reliability. Self-checking is particularly important during skill based tasks that could be performed without much conscious thought. The tool is required to be used at all times for manipulation of any plant equipment.

Peer-checking is the act of checking the correct component identification and subsequent manipulation prior to action being taken so that the actions to be taken will be correct. Unlike Concurrent Verification, peer-check may involve audio and/or visual cues and does not require documentation. Peer-Checking is used as defined in the Pre-Job Briefing. Most common uses are when mis-identification, mis-operation, or improper installation or assembly can have undesirable impact on people's safety or plant equipment. Other uses include a history of error or unfavorable experience with a particular action, or requested by a Peer in the field. Peer-Check is not required when utilizing CV or IV.

3.5.1 Peer-Checking Peer-checking is collaborative tool performed by two individuals. One acts as the doer, and the second person, a qualified peer, acts as the checker. The purpose of peer-checking is to prevent human error for a specific action, especially for critical steps or during a series of steps. Peer-Checking is merely two persons (performer and checker) self-checking in parallel, agreeing together

) that the action is the correct action to be performed and on the correct component.

Peer-checking can be confused with concurrent verification. Although the purpose of both techniques is to prevent error for a specific action, concurrent verification has the added purpose of configuration control. That is why the concurrent verification is documented with signatures in the guiding document.

Peer-checking is a technique to avoid a mistake in the operation/manipulation of a component, while concurrent verification helps avoid placing an important component in an undesired configuration needed for either operability or functionality of the system, structure, or component.

What action(s) is to be performed

Why it is to be done

How it is to be done

When it is to be done

Who is involved

What can go wrong

How to stop/correct/prevent an error or event CUES:

When mis-identification, mis-operation, or improper installation assembly can have adverse impact on people or plant equipment.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 12 of 16

Pre-determined in the Pre-Job Briefing

Requested by a PEER in the field

When required by the plant policy

Adverse Operating experience with the particular action or series of actions.

Behavior Standard for Peer-Checking:

The performer verbalizes the intended action out loud, pauses for the peer providing the peer-check to mentally process the action plan.

The performer waits for verbal agreement from the peer providing the peer-check. The performer proceeds with the action only if the peer providing the peer-check verbally agreed with the intended action.

The peer-check will be in visual and/or audible range of the performer.

At Risk Practices to Avoid with Peer-Checking:

Used in place of Independent Verification or Second Party Verification.

)

Checker not qualified with the task or is not experienced with the activity.

Checker not paying close attention to the performer.

Believing the performer will not err because of the performers experience or proficiency.

Checker unable to view component to be manipulated.

Checker not prepared to prevent an incorrect action taken by the performer.

Asking for a PEER-Check without directing the request to a specific person by name.

3.5.2 Self-Checking This technique focuses attention at important points in an activity before a specific act is performed. Once attention is focused, the individual takes a moment to think about the intended action and its expected outcome. Self-checking is particularly effective during skill-based tasks that could be performed without much conscious thought. Important steps to self-check involve touching plant equipment to change its status or may involve revising a document important for plant safety and reliability. In some cases these steps are determined by the component involved, which can initiate undesired outcomes if performed incorrectly.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 13 of 16 Cues to Self-Check:

Critical step identified during pre-job briefing

Time pressure - feeling a being hurried

Task interruption

Impending change in system or equipment status (especially maintenance disassembly and reassembly)

Behavior Standards For Self-Checking:

1. Have controlling document in hand

2. Prior to manipulation use Touch-STAR when manipulating components:

.§.top - Pause before performing operation/manipulation.

Eliminate distractions, if necessary. Focus attention on the step to be performed.

Ihink - Verify the action is appropriate for equipment/system status. Anticipate expected result(s) of the action and its indications. Consider what actions to take should expected result not occur (contingency).

Act - Without loosing eye contact, touch the component, label, etc. Compare component, label, etc., with checklist, procedure step, or drawing. State the component name or UNID allowed.

) Without loosing physical contact established earlier, perform the action.

Review - Verify anticipated result obtained. Perform contingency, if expected result does not occur.

3. If distracted, involving loss of visual or physical contact, then repeat the process to verify the proper component is about to be manipulated.

4. Slow - deliberate pace when proceeding through critical steps.

5. Stop when questions or discrepancies are encountered.

At-Risk Behaviors To Avoid During Self-Checking:

Carrying on a conversation while self-checking

Self-checking without guiding document

Attempting to perform more than one action at a time; no two-handed operations

Continuing with the action when questions or discrepancies occur

Looking at something other than component to be manipulated

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 14 of 16 4.0 RECORDS 4.1 QA Records None 4.2 Non-QA Records None 5.0 DEFINITIONS Second-Party/Concurrent Verification (CV) - The act of verifying a condition, such as lifting a lead or installing a jumper, concurrent with the activities related to establishing the condition.

The individual performing the second-party verification and the performer must reach agreement that the activity/manipulation to be performed is correct before performance. The terms second-party/concurrent verification are synonymous and may be used interchangeably. For the verification process, the main focus is on Configuration/Status Control.

Independent Verification (IV) - The act of checking a condition, such as a component position, separately from the act of establishing the condition. The individual performing the IV must physically check the condition without relying on observation or verbal confirmation by the initial performer. The verifier must be physically independent as well as independent by time. In the true meaning of Independent Verification, the performer and the verifier may receive the pre-job

) brief together but not be associated together for the activity. In trying to keep the integrity of the verification, the verifier cannot rely on any visual/audible ques of the performer. The object of this verification process is to not pollute the verification with any information from the performer.

Peer-Checking - The act of checking the correct component identification and subsequent manipulation prior to action being taken so that the actions to be taken will be correct. Unlike Concurrent Verification, peer-check may involve audio and/or visual cues and does not require documentation. Peer-Check is not required when utilizing CV or IV. When PEER-Checking, the main focus is on the action itself, not as much "Configuration Control."

Self-Checking - This technique focuses attention at important points in an activity before a specific act is performed. Once attention is focused, the individual takes a moment to think about the intended action and its expected outcome. Self-checking is particularly effective during skill-based tasks that could be performed without much conscious thought. Important steps to self-check involve touching plant equipment to change its status or may involve revising a document important for plant safety and reliability. In some cases these steps are determined by the component involved, which can initiate undesired outcomes if performed incorrectly.

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 15 of 16 APPENDIX A Page 1 of2 SYSTEMS AND COMPONENTS REQUIRING INDEPENDENT OR SECOND-PARTY VERIFICATION SYS WBN SQN BFN 01 Main Steam (safety-related Main Steam System (safety- Main Steam (safety-related portion) related portion) portion) 03 Main Feedwater System Main Feedwater System Reactor Feedwater System (safety-related portion) (safety-related portion) (safety-related portion) 03 Auxiliary Feedwater System Auxiliary Feedwater System N/A 13 Fire Detection System Fire Detection System Fire Detection System 23 N/A N/A RHR Service Water System 26 High Pressure Fire Protection High Pressure Fire Protection High Pressure Fire Protection 30 Auxiliary Building Gas Auxiliary Building Gas Treatment HVAC (Refueling Zone, Reactor Treatment System, Lower System, MCR Ventilation System, Zone, Turbine Bldg., Radwaste Compartment Cooler Fans, Lower Compartment Cooler Fans, Bldg.)

Containment Air Return Fans Containment Air Return Fans 31 MCR Ventilation Covered by System 30. Control Bay and Off-Gas Building HVAC(CREV) 32 Essential Air System Essential Air System Control Air System (Reactor Bldg.

And Drywell) 39 C02 Storage and Fire Protection C02 Storage and Fire Protection C02 Storage and Fire Protection

) 41 Layup Water Treatment System Layup Water Treatment System Halon Fire Protection System 43 Post Accident Sampling System Post Accident Sampling System - Sampling and Water Quality

- Those parts of the system that Those parts of the system that System - Those parts that isolate isolate RCS, RHR, Containment isolate RCS, RHR, Containment releases to the environment and and flush water and flush water establish primary and secondary containment 62 Chemical Volume & Control Chemical Volume & Control N/A System System 63 Emergency Core Cooling Emergency Core Cooling System Standby Liquid Control System System I 64 N/A N/A Reactor Building Heating and Ventilation System, Primary Containment and Isolation System 65 Emergency Gas Treatment Emergency Gas Treatment Standby Gas Treatment System System System 66 N/A N/A Offgas System 67 Essential Raw Cooling Water Essential Raw Cooling Water Emergency Equipment Cooling System System Water System 68 Reactor Coolant System Reactor Coolant System Reactor Recirculation System 69 N/A N/A Reactor Water Cleanup System (RWCU)

)

TVAN STANDARD SPP-10.3 PROGRAMS AND VERIFICATION PROGRAM Rev. 1

) PROCESSES Page 16 of 16 APPENDIX A Page 2 of2 SYSTEMS AND COMPONENTS REQUIRING INDEPENDENT OR SECOND-PARTY VERIFICATION SYS WBN SQN BFN I 70 Component Cooling Water Component Cooling Water Rx Bldg Closed Cooling Water System System System (RBCCW) 71 N/A N/A Reactor Core Isolation Cooling System (RCIC) 72 Containment Spray System Containment Spray System Auxiliary Decay Heat Removal 73 N/A N/A High Pressure Coolant Injection System (HPCI) 74 Residual Heat Removal (RHR) Residual Heat Removal (RHR) Residual Heat Removal System System System (RHR) 75 N/A N/A Core Spray System 76 N/A N/A Containment InertinQ System 77 Radwaste Systems - Those Radwaste Systems - Those parts Liquid Radwaste System (Floor parts that isolate releases to that isolate releases to the and Equipment Drains) - Those the environment environment parts that isolate releases to the environment and establish primary and secondary containment 78 Spent Fuel Pit Cooling System Spent Fuel Pit Cooling System Fuel Pool Cooling and Cleanup System 82 EmerQency Diesel Generators EmerQency Diesel Generators Standby Diesel Generators 83 Hydrogen Recombination Hydrogen Recombination N/A 84 Flood Mode Boration System Flood Mode Boration System Containment Atmosphere Dilution System 85 Rod Control Rod Control Control Rod Drive Hydraulics 86 Refer to System 82 Refer to System 82 Diesel Air Start System 88 Containment Isolation System - Containment Isolation System - N/A including valves/components including valves/components that that provide a containment provide a containment isolation isolation function function 90 Radiation monitoring systems - Radiation monitoring systems - Radiation monitoring system -

Those parts of the systems that Those parts of the systems that Those parts that isolate releases provide isolation functions to provide isolation functions to to the environment and establish effluent pathways effluent pathways primary and secondary containment 99 ESFAS & RPS ESFAS & RPS Reactor Protection System 268 Permanent Hydrogen Mitigation Permanent Hydrogen Mitigation N/A

-- Reactor core (Fuel and Reactor core (Fuel and Reactor core (Fuel and component locations) component locations) component locations)

-- Class 1E Electrical Distribution Class 1E Electrical Distribution Class 1 E Electrical Distribution System. System. System.

NOTE N/A indicates that the associated system number is not utilized at the referenced plant.

JPM NO. 540 REV. NO.3 PAGE 1 of 8

)

JPM NUMBER: 540 TITLE: DETERMINATION OF OVERTIME ELIGIBILITY ADMIN: Conduct of Operations PROVIDE CANDIDATE WITH A COPY OF: OSIL 25 & SPP-1.5 (Only If Requested)

SUBMITTED BY: DATE:--'--_ __

VALIDATED BY: DATE: _ _ __

APPROVED BY: DATE: _ _ __

TRAINING PLANTCONCURRENCE: _ _ _ _ _ _ _ _ ___ DATE: _ _ __

OPERATIONS

Examination JPMs Require Operations Training Manager Approval or Designee Approval and Plant Concurrence

)

JPM NO. 540 REV. NO.3 PAGE 2 of 8 REVISION LOG Revision Effective Pages Description Number Date Affected Of Revision 0 08/28/05 All Initial issue 1 02/16/06 All Procedure revision 2 06/15/07 All Procedure revision 3 08/08/08 All General revision & re-format, Modified for 0707 NRC exam

)

JPM NO. 540 REV. NO.3 PAGE 3 of 8 OPERATOR: _____________________________________________

RO SRO DATE: __________

JPM NUMBER: 540 TASK TITLE: DETERMINATION OF OVERTIME ELIGIBILITY KIA NUMBER: 2.1.1 KIA RATING: RO 3.7 SRO 3.8 TASK STANDARD: GIVEN APPROPRIATE INFORMATION, DETERMINE OPERATOR OVERTIME ELIGIBILITY.

PERFORMANCE LOCATION: CLASSROOM: X REFERENCES/PROCEDURES NEEDED: OSIL 25 8-29, SPP-1.5 Rev 5 VALIDATION TIME: CLASSROOM: 15:00 MAX. TIME ALLOWED: _ _ _ _ _ (FOR TIME CRITICAL JPMs ONLY)

PERFORMANCE TIME:

COMMENTS:

ADDITIONAL COMMENT SHEETS ATTACHED? YES _ _ NO RESULTS: SATISFACTORY UNSATISFACTORY EXAMINER SIGNATURE: DATE: _ _ _ _ _ ___

JPM NO. 540 REV. NO. 3 PAGE 4of8

')

Classroom INITIAL CONDITIONS: A startup is planned for the following shift. One Reactor Operator must be held over two hours for startup. The following is the work history (excluding shift turnover time) of the available reactor operators on shift (hours reflect those worked PRIOR to the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months holdover). A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months occurred between all work periods.

INITIATING CUES: Evaluate the work history for all 5 operators. Determine for each Operator if; A. They can be held over for two hours WITHOUT a waiver, AND; B. They can be held over for two hours WITH a waiver, AND; C. They cannot be held over WITH or WITHOUT a waver

)

JPM NO. 540 REV. NO.3 PAGE 5 of 8 EXAMINER'S SOLUTION DO NOT GIVE TO STUDENT Step Description Standard SAT/UNSAT Reference SPP-1.5, Current Revision SPP-1.5 and OSIL OSIL 25 25 (If requested) 1 Evaluate Operator 1 Determine Operator #1 would ~-NO exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period 8 - YES and would require overtime C- NA autho rizatio n 2 Evaluate Operator 2 Determine Operator #2 would ~-NO exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period 8 - YES and exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months C- NA period and would require overtime authorization 3 Evaluate Operator 3 Determine Operator #3 is already A-NO on waiver for greater than 72 in 7 8-NO days and would exceed 85 hours9.837963e-4 days 0.0236 hours 1.405423e-4 weeks 3.23425e-5 months in C-YES a 7 day period which cannot be

~aivered per OSIL-25 I

4 Evaluate Operator 4 Determine that Operator #4 would A-YES not exceed any overtime guidelines 8 - NA C- NA 5 Evaluate Operator 5 Determine Operator #5 would ~-NO exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a 7 day period 8 - YES and would require overtime C- NA authorization ALL STEPS ARE CRITICAL - 4 of 5 CORRECT TO PASS JPM (80%).

JPM NO. 54.1f REV. NO.3 PAGE 6 of 8 Examiner's Copy A startup is planned for the following shift. One Reactor Operator must be held over two hours for startup. The following is the work history (excluding shift turnover time) of the available reactor operators on shift (hours reflect those worked PRIOR to the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months holdover). A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months occurred between all work periods can work can work with Cannot work DAY 1 2 3 4 5 6 7 8 without waiver waiver with or without Today waver Operator #1 yes no/na yes no/na yes no/na 12 12 12 12 0 0 12 12 Operator #2 yes no/na yes no/na yes no/na 0 12 12 12 0 12 12 12 Operator #3 yes no/na yes no/na yes no/na 0 12 12 12 12 12 12 12 Operator #4 yes no/na yes no/na yes no/na 12 12 12 0 12 12 10 12 Operator #5 yes no/na yes no/na yes no/na 0 10 8 8 12 12 12 10 INITIATING CUES: Evaluate the work history for all 5 operators. Determine for each Operator if; A. They can be held over for two hours WITHOUT a waiver, AND; B. They can be held over for two hours WITH a waiver, AND; C. They cannot be held over WITH or WITHOUT a waver Circle the correct responses above for each operator ALL STEPS ARE CRITICAL - 4 of 5 CORRECT TO PASS JPM (80%).

PAGE 7of8 CANDIDATE HANDOUT INITIAL CONDITIONS: A startup is planned for the following shift. One Reactor Operator must be held over two hours for startup. The following is the work history (excluding shift turnover time) of the available reactor operators on shift (hours reflect those worked PRIOR to the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months holdover). A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months occurred between all work periods.

INITIATING CUES: Evaluate the work history for all 5 operators. Determine for each Operator if; A. They can be held over for two hours WITHOUT a waiver, AND; B. They can be held over for two hours WITH a waiver, AND; C. They cannot be held over WITH or WITHOUT a waver

\'~"

PAGE 8 of8 CANDIDATE HANDOUT TASK CONDITIONS:

A startup is planned for the following shift. One Reactor Operator must be held over two hours for startup. The following is the work history (excluding shift turnover time) of the available reactor operators on shift (hours reflect those worked PRIOR to the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months holdover). A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months occurred between all work periods can work can work with Cannot work DAY 1 2 3 4 5 6 7 8 without waiver waiver with or without Today waver Operator #1 yes no/na yes no/na yes no/na 12 12 12 12 0 0 12 12 Operator #2 yes no/na yes no/na yes no/na 0 12 12 12 0 12 12 12 Operator #3 yes no/na yes no/na yes no/na 0 12 12 12 12 12 12 12 I

I Operator #4 yes no/na yes no/na yes no/na 12 12 12 0 12 12 10 12 Operator #5 yes no/na yes no/na yes no/na 0 10 8 8 12 12 12 10 INITIATING CUES: Evaluate the work history for all 5 operators. Determine for each Operator if; A. They can be held over for two hours WITHOUT a waiver, AND; B. They can be held over for two hours WITH a waiver, AND; C. They cannot be held over WITH or WITHOUT a waver Circle the correct responses above for each operator

(

( I

TITLE SPP-1.S 1m Overtime Restrictions (Regulatory)

Rev.OOOS Page 1 of 10 Quality Related o Yes o No TVAN Standard pORe Required o Yes o No Programs and Processes Effective Date 05-04-2007 I

Responsible Peer Team: Plant Managers Concurred by: David A. Kulisek 4/24/07 Primary Sponsor Date Concurred by:

N/A N/A Peer Team Mentor Date Approved by:

N/A N/A I

General Manager, NA Date Yahya Sadre for PDS 4/30107 Approved by: *Senior Vice President, Nuclear Operations Date

Site-specific changes are approved by Site Sponsor and Site Vice President (see PCF)

TVAN Standard Overtime Restrictions SPP-1.S Programs and (Regulatory) Rev.OOOS Processes Page 2 of 10 Revision Log Revision or Affected Change Effective Page Number Date Numbers Description of Revision/Change 0 10/8/97 All Replaces STD-1.7, SSP-1.7 at SQN and BFN; and SSP-1.07 at WBN.

1 5/20/99 2,4-10 Revised to allow Plant Manager or Site Vice President to delegate approval of exception to overtime during outages, and modified requirements for review of the monthly overtime exception report.

2 11/15/99 2,5,8,9 Added exception to paragraph 3.B.3 and adjusted Forms SPP-1.5-1 and SPP-1.5-2 accordingly (minor/editorial change).

3 10/25/01 2-10 Revised to implement Browns Ferry (BFN)

Technical Specifications change 403. The following BFN specific changes were made: removed allowance for Site Vice President to delegate

,) approval of exception to overtime during outages, deleted requirements for Plant Manager review of the monthly overtime exception report, and added a requirement to conduct a periodic independent review of overtime use by Human Resources. Also, incorporated actions from SQN and WBN PERs including: added statement concerning FFD Program to scope. Paragraph 3.B.3 example was modified for the 8-hour break period.

Responsibilities for individuals, supervisors and management were added. Turnover Time, Break /

8-hour break, and Work Time and were added to the definitions.

TVAN Standard Overtime Restrictions SPP-1.5 Programs and (Regulatory) Rev. 0005

.. _.~oce!:>sE!s ___ Page 3 of 10 Revision Log Revision or Affected Change Effective Page Number Date Numbers Description of Revision/Change 4 7/7/04 2-8 Revised to reflect management expectations for supervisors to ask if an individual will exceed any OT limits when or prior to: (1) holding over anyone to perform safety related work; (2) calling anyone in I to perform or directly supervise safety related work; and, (3) during outages, prior to starting safety related work. (WBN PER 02-003508-000 CA2),

and to eliminate scheduling personnel in excess of overtime limits (WBN PER 02-003508-000 CA3) and to clarify, and make consistent with Tech Specs, which positions require exception forms be completed (WBN PER 02-003508-000 CA4). Additionally, eliminated the use of blanket authorizations and clarified that exemptions were to be used only in "very unusual circumstances" as identified in NRC GL 82-12. These were cited in a NRC violation issued to ANO on 7/19/02.

5 05/04/07 3,9,10 This document has been converted from Word 95 to Word 2002 (XP) using Rev. 4.

Minor/editorial change: Added Requirements and References section 6.0.

Minor/editorial change: Revised Form SPP-1.5-1 to allow for Employee Identification Number (EIN) instead of a Social Security Number (SON PER 94551-001 ).

)

TVAN Standard Overtime Restrictions SPP-1.S Programs and (Regulatory) Rev.OOOS Processes Page 4 of 10 Table of Contents 1.0 PURPOSE ................................................................................................................................. S 2.0 SCOPE ...................................................................................................................................... S 3.0 INSTRUCTIONS ........................................................................................................................ S 3.1 Requirements ............................................................................................................................. 5 3.2 Responsibilities .......................................................................................................................... 7 4.0 RECORDS ................................................................................................................................. 8 4.1 QA Records ............................................................................................................................... 8 4.2 Non-QA Records ........................................................................................................................ 8 S.O DEFINITIONS ............................................................................................................................ 8 6.0 REQUIREMENTS AND REFERENCES .................................................................................... 9 Appendix A: SPP-1.S-1 Overtime Limitation Exception Report ............................................ 10

)

TVAN Standard Overtime Restrictions SPP-1.5 Programs and (Regulatory) Rev. 0005 Processes Page 5 of 10 1.0 PURPOSE A. This SPP establishes TVAN's regulatory overtime program to meet regulatory requirements and as specified in individual plant Technical Specifications. This procedure does not eliminate adherence to any of the requirements of the TVAN Fitness for Duty Program (ref. SPP-1.2).

B. The intent of these controls are to prevent situations where fatigue could reduce the ability of operating personnel to keep the reactor in a safe condition, and to assure that critical plant operating personnel are not assigned to shift duties while in a fatigued condition which could significantly reduce their mental alertness or decision making ability.

2.0 SCOPE A. The following personnel are required to meet the limitations specified in this SPP:

senior reactor operators [SROs], licensed reactor operators [ROs], radiological control technicians (health physicists), auxiliary operators (assistant unit operators [AU Os]) ,

and key maintenance personnel (see 5.0 Definitions), including craft and contractors performing safety-related activities in the capacity of the TVA positions listed.

B. Although not specifically restricted by Technical Specifications, the limitations specified in this SPP should be used as a guide for all personnel performing work.

3.0 INSTRUCTIONS 3.1 Requirements A. Adequate shift coverage shall be maintained without routine heavy use of overtime.

The objective shall be to have personnel within the scope of this SPP, work an 8,10, or 12-hour day, a nominal 40-hour week while the plant is operating.

B. On a temporary basis, substantial amounts of overtime may be used as required:

To support unforeseen problems, or

During extended periods of shutdown for refueling outages or

Major maintenance projects or

Major plant modifications In these cases the following restrictions apply:

1. An employee may work no more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time (see 5.0 Definitions). For exceptions, see 3.1 D below.

As an example an employee has worked on shift 16 continuous hours. At the end of the shift it takes one hour for shift turnover which means the employee has

) worked 17 continuous hours. Since the one hour over 16 was for shift turnover, an authorization for deviation from the overtime limitation is not required.

TVAN Standard Overtime Restrictions SPP-1.5 Programs and (Regulatory) Rev. 0005 Processes Page 6 of 10 3.1 Requirements (continued)

2. An employee may work no more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48-hour period, or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period, excluding shift turnover time (see 5.0 Definitions). For exceptions, see 3.1 D below.

For example a 7-day period is any combination of 7 consecutive days or rolling days (such as Monday through Sunday or Tuesday through Monday). Hours worked over 72 require authorization for exception to the limitation, excluding shift turnover (for exceptions, see 3.1 D below).

3. Employees are to receive at least an 8-hour break (see 5.0 Definitions) between work periods, including shift turnover time (see 5.0 Definitions). For exceptions, see 3.1C below.

The intent of this requirement is to ensure workers have at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months off work for recuperation to prevent fatigue. A work period is the normal working shift each day such as 5 eight hour shifts or 4 ten hour shifts etc. and does include both work time and turnover time as defined in Section 5.0 Definitions. Therefore, the limitations for the 8-hour break can be applied by the following:

a. An employee must have had at least a continuous 8-hour break within the previous rolling 24-hour period. Turnover time is not included in the break time.

b. If the employee has not had a continuous break of at least 8-hours within the previous rolling 24-hour period, an approved Exception Report must be obtained prior to exceeding 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months past the end of the last 8-hour continuous break.

c. For example, an employee has been working his normal 8-hour shift starting at 0730 to 1600 and then has a 112-hour turnover (turnover ends at 1630).

The end of the last 8-hour break was at 0730. At 2000 he is called back to work and reports at 2100. His break time has been 4-1/2 hours. He does not require an approved exception since he had greater than an 8-hour break within the previous 24-hours (2100 the day before to 0730 that morning). However, if this same employee was called and reported in at 2400, an approved Exception Report would be required for not having had an 8-hour continuous break. (Neither the break from 2400 to 0730 that morning or from 1630 to 2400 that night provides a continuous 8-hour break.) Continuing with the first scenario, the employee reports to work at 2100 and works to 2400 and then has a 1/2-hour turnover to 0030. At 2330, an Exception Report for failing to have a 8-hour break must be approved (prior to exceeding 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from the end of the last 8-hour break).

Additionally, if the employee reports to work at his normal time of 0730 an approved Exception Report must be completed prior to his return. The employee can not return to work prior to 0830 without having an approved Exception Report for not having the continuous 8-hour break.

)

TVAN Standard Overtime Restrictions SPP-1.5 Programs and (Regulatory) Rev. 0005

') Processes Page 7 of 10 3.1 Requirements (continued)

4. When substantial overtime is required, it shall be assessed and monitored to ensure the above-listed restrictions are met. Assignment of overtime is made after consideration of such factors as on going activities, expected duration, and personnel involved, with the intent to minimize potential impact on safety-related activities.

5. When a control board operator must work overtime, every effort must be made to remove the operator from the responsibility of operating the control room board.

C. Personnel performing safety-related activities (in the capacity of the TVA positions listed herein) should not be scheduled to exceed any overtime limit.

D. The Plant Manager or Plant Manager's designees may authorize deviations from the overtime restrictions on the Overtime Limitation Exception Report (Form SPP-1.5-1).

1. The Plant Manager's designees may approve deviations other than exceeding 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight.

2. During outages, the Plant Manager may delegate the authority to exceed 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months to the Outage Manager or the Assistant Plant Manager.

3. The deviation will be based on the following criteria.

a. Very unusual circumstances exist.

b. Significant reduction in personnel effectiveness would be highly unlikely.

E. The exact work to be performed is specified in the Overtime Limitation Exception Report in sufficient detail for the authorizing manager to review and conclude that significant reduction in personnel effectiveness would be highly unlikely.

F. The form must be filled out and approved before the individual(s) exceeds the overtime limit. If approval is received by telecon, the preparer shall document that approval was via telecon, initial, date, and given by whom.

G. The Plant Manager shall designate periodic independent reviews to monitor program compliance and to ensure excessive hours have not been assigned, and that Overtime Limitation deviations are being requested and authorized when required.

3.2 Responsibilities A. Employees are responsible for notifying their supervisor of the potential for exceeding the limits specified in this procedure well in advance of such occurrence. Employees may not exceed the limits specified by this procedure or exceed the hours as approved by an Exception Report. Employees must have an 8-hour break as defined in this procedure or an Exception Report must be approved prior to starting work and/or a pre-work turnover.

B. Supervisors will ask if an individual will exceed any overtime limits prior to holding anyone over to perform safety-related work.

TVAN Standard Overtime Restrictions SPP-1.S Programs and (Regulatory) Rev. 0005 Processes Page 8 of 10 3.2 Responsibilities (continued)

c. Anyone calling someone in to perform or directly supervise safety-related work will ask if the individual will exceed any overtime limits.

D. During outages, prior to starting safety-related work, the supervisor will ensure that each individual will not exceed any overtime limit.

E. Supervisors or designee are responsible for preparing the applicable Exception Report, ensuring that a significant reduction in personnel effectiveness would be highly unlikely, and obtaining management approval prior to an employee exceeding the limits as specified in this procedure. This includes obtaining required authorization for an employee not having an 8-hour break as defined in this procedure. The Supervisor or designee is responsible for distribution of approved Exception Reports as shown on the report.

F. Management (as limited by this procedure) is responsible for assessing the need for exceeding overtime/break limits including the justification that a significant reduction in personnel effectiveness would be highly unlikely. If appropriate, authorization is provided on the Exception Report.

4.0 RECORDS 4.1 QA Records None 4.2 Non-QA Records Form SPP-1.5-1, Overtime Limitation Exception Report 5.0 DEFINITIONS EMPLOYEE STATUS:

Every 24-hour day can be broken into three categories of time, Turnover, Work, or Break.

The sum of these three times will always be 24. Therefore, for the purposes of this procedure, all employees are always in one of the three statuses, either working or in turnover or on break as defined by:

Break I 8-hour Break - For the purposes of this procedure, an 8-hour break is considered to be continuous. The initiation of the break period is after the completion of work time or any turnover time following work time. The break period ends at the time the employee starts work time or a pre-work turnover time.

Turnover Time - For the purposes of this procedure, turnover time is neither counted as work time nor break time. Turnover time is conducted prior to work time and/or at the end of work time. Therefore, the 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24,24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48-hour period, or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period does not include time spent in turnover. Likewise, hours spent in turnover are not included as part of the 8-hour break period.

TVAN Standard Overtime Restrictions SPP-1.5 Programs and (Regulatory) Rev. 0005 Processes Page 9 of 10 5.0 DEFINITIONS (continued)

Work Time - For the purposes of this procedure, work time is the time spent working (including lunch etc.) and DOES NOT include time in Turnover (both pre-work or at the end of work). The limitations of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in a 48-hour period, and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period apply to work time.

Key Maintenance Personnel - The term applies to all TVAN and contractor personnel who are subject to performing maintenance, repair, calibration or testing of safety-related structures, systems or components or personnel who are directly supervising such activities.

6.0 REQUIREMENTS AND REFERENCES Requirements and References are contained in the "SPP-1.S REQ & REF" document.

TVAN Standard Overtime Restrictions SPP-1.S Programs and (Regulatory) Rev.OOOS Processes Page 10 of 10 Appendix A (Page 1 of 1)

SPP-1.S-1 Overtime Limitation Exception Report OVERTIME LI.MITATIONEXQEPTION RePORT The following ~mployees (three or less) are authorized to work overtime in excess oftne overtime limits specified in paragraphs

......... o.iU.....-*

v, ........................ ~ ....... "- .............. , ....... ~ ." ..... , .... 'Y~.~ ., ..................... I* ...... ,t ...... " ....... ~*,... ...... .... 'Il~" .... ~-."'" .* ~ * ........ _ I .... ".-~ ..... ,.~*

Name SSN/EIN Organization Hours ExceedIng Requirements O~ # HQUf;S*. TimeUI.nit WiIl.Be EXCE!eded

a. Time:

b, .*. Time:

c. Time;

1. Specify specific reason for cause ofovertime (i.~') another employee on SIL, unexpected jab, etc.).

2. Specify requlrement{s) for exception; D 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months continuous D &-hourwork break o 16 hrs. in 24 0 24 hrs. in 48 0 72 hrs.in 7-day period

3. Specify exact work to beperfotmed and brief desoription (include procedure number, workplan number. work request number., or support of otilJ:!rspJ:!cific actiVity):

4. State Justification for exceeding the overtime limits. Justification should address considerations and actions t"ken to minimize potential impacts on safety-related activities (Le., rescheduling of task, asslgnment of alternate employee to task, etc.) and reasoning for the determination that a significant reduction in effectlvJ:!ness oftM personnel involved will not res.ult. This analysis should indude considerations stich as the total amount of time workedJanticipated, break periods takenfplanned,* type of actiVity to be performed, etc., fQrpersonn!;l! t() ensure thatfatigUe is notlwill*not be a factor.

Prepared by Print Name Prepared by Date _~--...._.,---" Time .,..-~_ _~ Ext. _ _ __

Signature THEABoVEACTIVITY.,SCONSIOEREOANUNUSUACC(RCUMSTANCEANDWARRANfsE:X.CEEDINGTHEOVERTtME RESTRICTIONS.

Approved b.y Print Name/Posltion Approved by Date _ _ ~ _ __

PIi:!fJfM~nf:igerJAYthorti~(\~~$ignee. !~ure FOR ADDITIONALSPACE; USE>~s.VERSeSIDEOF tHIs FORM.

Distrib.utedto: Plant Manager's Office ..

Originator (Employees Section Manager)

TVA 40551 [05-2007] Page 1 of 1 SPP-1.5-1 [05-04-2007J

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TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE 1 OF 8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY 1.0 PURPOSE Provide for uniformity in filling vacant overtime shifts, implementing regulatory restrictions with regard to hours worked, instructions for work in excess of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , establish Operations leave policy, and establish the Operator Relief policy.

2.0 REFERENCES

SPP-1.5, Overtime Restrictions (Regulatory)

NUREG-0737 LA. 1.3 (IE Circular No. 80-02)

NRC Generic Letters 82-12 and 83-14 BP-110 and 107 R30 940316 859 SPP-l.2, Fitness For Duty BFPER 99-012810-000 BFPER 99-012812-000 3.0 DEFINITIONS NONE 4.0 INSTRUCTIONS 4.1 Filling Overtime for Employees on Continuous Shifts 4.1.1 Employees in temporary classifications are considered as permanent employees for the purpose of filling overtime.

4.1.2 When the overtime position to be filled requires documented qualifications or training, employees selected to work overtime shall meet these requirements in addition to being in the appropriate classification. Reference Attachment 2 when calling in for overtime.

4.1.3 The Shift Manager will be responsible for designating the required overtime positions, and these positions will be filled in the order listed below, provided that employees may be called without exceeding the overtime limits described in Standard Programs and Practices (SPP-1.5).

Vacancies of normally assigned positions caused by pre-approved annual leave will be filled by use of a sign up sheet for employees in classification (reference Attachment 3).

The Shift Manager is responsible for uniquely identifying the required positions to be filled by overtime in the Master Leave Book. The Unit 1 Operator or the Operations Clerk will refer to the Master Leave Book to obtain the number of required positions, which have been uniquely identified by the Shift Manager, for the sign up sheet. Once the sign up sheet has been filled out, the Shift Manager will review and approve the sign up sheet then the Unit Operator will start the calling process at approximately 1300 on Sunday.

The sign up sheet will begin on Monday at 0700 for the beginning of each week. The sign up sheet will list the date and shift for each vacancy. The Unit 1 Operator will call the low overtime person on the sign up sheet on Sunday beginning at approximately 1300 to inform him/her what shift he/she is to work during the next week. If you sign up and you are the low person, you are expected to work that shift. All other overtime will be filled by the following:

Origanl Signed by Robert Marsh 08/2912005 Operations Superintendent DATE

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE20F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05

.\ II OVERTIME, LEAVE, AND RELIEF POLICY 4.1 Filling Overtime for Employees on Continuous Shifts (Continued) 4.1.3.1 Employees in classification in which the overtime shifts are to be worked are first offered the shift on their off-days in order of their overtime hours, provided it will not require them to work 16 consecutive hours.

4.1.3.2 If no one is available in group 1 above, employees in classification who have had one 8-hour offshift and who will not have to work 16 consecutive hours are offered the shift in order of their overtime hours.

4.1.3.3 If no one is available in groups 1 or 2 above, employees onshift in classification who are working overtime on their offdays are offered the shift in order of their overtime hours.

NOTE: *Agreed to by Operations Superintendent and Job Steward: All trainees will be offered overtime in accordance with the overtime procedures, but SHALL NOT BE FORCED OVER.

4.1.3.4 If no one is available in groups 1, 2, or 3 above, employees onshift in classification are offered the shift in order of their overtime hours. If the employee is involved in a shift swap that could cause him to violate nuclear staff work hours as outlined in SPP-1.5, this will require the Plant Manager's (or Duty Plant Manager's) approval. Complete Form SPP 1.5-1.

4.1.3.5 If no one is available in groups 1-4 above, those employees who were bypassed initially because they would have exceeded 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a seven-day period will be offered the shift on a voluntary basis following steps 1-4 above; provided that no employee may work more than 85 hours9.837963e-4 days 0.0236 hours 1.405423e-4 weeks 3.23425e-5 months in any seven-day period. Exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, or 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in a seven-day period would require a deviation from the overtime limits set forth in SPP-1.5. This is documented using Form SPP 1.5-1.

4.1.3.6 If no one is available in groups 1-5 above, the employee low on overtime onshift in classification and who is not on overtime is held over. (If a person is working evening shift and is on annual leave (AIL) the following day, he or she cannot be forced over to work a midnight shift. The person may elect to have their shift changed and save the AIL.) Normally, an employee will not be required to double over 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , more than every other day. If the employee is involved in a shift swap that would cause him to violate nuclear staffwork hours as outlined in SPP-1.5, this will require the Plant Manager's (or Duty Plant Manager's) approval. Complete Form SPP 1.5-1. Employees not exceeding the overtime limits set forth in SPP-1.5 would be given priority over those requiring deviations, provided that no employee may work more than 85 hours9.837963e-4 days 0.0236 hours 1.405423e-4 weeks 3.23425e-5 months in a seven-day period.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE30F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY 4.1 Filling Overtime for Employees on Continuous Shifts (Continued) 4.1.3.7 If the employees held over requests relief, the employees on the incoming shift in classification are called at home. It is not necessary that he/she wait or that the Shift Manager (SM) wait until the shift actually starts to obtain a relief.

4.1.3.8 An employee on scheduled AIL is not called in until employees onshift and those on offdays are required to work more than one shift per day.

4.1.3.9 Use the Following guidance to fill upcoming vacancies in the Work Control Center SRO (WCC SRO) and the Shift Support Tagging (SST) positions, as deemed necessary by the Shift Manager.

1. WCC SRO

a. Call out for Unit Supervisors that are currently on their off days.

b. Ifno one accepts; call out for Shift Managers currently on their off days.

2. SST

a. Call the qualified Unit Operator (UO) per the call-out procedure, do not skip over non step 3 qualified people (arrange position swaps as necessary), if no one accepts;

b. Force UO on shift per call-out procedure (arrange position swaps as necessary), force qualified UO on-shift if position swaps will not meet minimum manning requirements.

4.1.3.10 To ensure administrative requirements are met for having 2 active licenses in the Unit 1 and 2 spaces, Unit 1 is to be manned with an Active Licensed Unit Operator. This is an Administrative appointment and not required by the Code of Federal Regulations or Technical Specifications when the Unit is de-Fueled.

4.2 Filling Overtime on Holidays Vacancies are filled in overtime order of those individuals that were "Holidayed Off' on that shift (Not for all shifts that day, but just for the shift in question). Persons holidayed off for the day shift will be called for the day shift in overtime order, then any additional vacancies will be filled in accordance with the overtime procedure. The same would apply for all shifts on that holiday.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE40F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05

) II OVERTIME, LEAVE, AND RELIEF POLICY 4.3 Filling Overtime for Operations Employees not on Continuous Shifts 4.3.1 Overtime for Operations employees not on continuous shifts is filled in accordance with the General Agreement, which states that overtime shall be distributed among the qualified employees in the group in which the overtime is worked.

4.3.2 Mandatory overtime is filled with the qualified employee in the group with the lowest hours of overtime worked.

4.4 Recording Overtime for all Operations Employees 4.4.1 Records of paid overtime worked or refused are kept by the Operations clerk on a biweekly basis and are made available upon request to labor representatives.

4.4.2 For Operations employees, overtime hours refused are not considered in detennining overtime hours.

4.4.3 Overtime hours are zeroed for all employees at the end of the pay year.

4.4.4 Operations employees who are hired or transferred between locations are recorded with an

') amount of overtime equal to the average of the new classification and location.

Reclassified Operations employees at the same location retain their accumulated overtime hours.

4.5 Regulatory Restrictions With Regard to Work Hours Fatigue, especially if due to loss of sleep, results in a marked deterioration of a person's response to visual signals, increases the time for a person to make a decision, and result in more personnel errors.

Additionally, as a person becomes more fatigued the person tends to ignore more and more signals.

In recognition of this, limitations have been established in the references for personnel work hours.

INSTRUCTIONS

The controls shall apply to the plant staffwho perfonn safety-related functions (Senior Reactor Operators, Reactor Operators, and Assistant Unit Operators (AUOs)).

a. Work no more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight (not including shift turnover time).

b. Have a break of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or more (which can include shift turnover time) between all work periods.

c. Work no more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period (excluding shift turnover time).

d. Work no more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period (excluding shift turnover time).

For shift swaps which exceed the guidelines stated above, a written request must be submitted to the Operations office as far in advance as possible and not less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in advance, except in emergencies. The request must include the shifts requested to be swapped and the reason the shift swap is needed. Swaps will be allowed if the reason is detennined to be legitimate, pending approval ofthe Plant Manager or Duty Plant Manager.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE50F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY

These authorized deviations to the working hour guidelines shall be documented utilizing Form SPP 1.5-1 and available for NRC review.

Only when it is necessary, due to emergencies or critical load situations, should supervisors permit any employee to work more than 16 continuous hours without a non work period.

Work in Excess of 16 Hours Anyone required to work 16 continuous hours or more shall, upon being relieved, report to the SM for instructions on when to return to work. These instructions will provide a non work period of at least eight continuous hours before returning to work. He/she will be paid his/her regular rate for that part of the rest period which falls within the hours of his/her regularly scheduled straight-time shift.

If the work period is followed by a scheduled rest period of greater than eight hours, it is not necessary to contact the SM.

The Plant Manager or Site Vice President may authorize deviations from the overtime restrictions on the Overtime Limitation Exception Report (Form 1.5-1). Their designees may approve deviations other than exceeding 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight. Form 1.5-2 will be used during periods of extended shutdown for refueling major maintenance, or major plant modification as determined by Plant Manager. The deviation will be based on the following criteria:

1. Unusual circumstances exist.

2. Significant reduction in personnel effectiveness would be highly unlikely.

The exact work to be performed is specified in the Overtime Limitation Exception Report in sufficient detail for the authorizing manager to review and conclude that significant reduction in personnel effectiveness would be highly unlikely.

The form must be filled out and approved before the individual( s) exceeds the overtime limit. If approval is received by telecon, the Preparer shall document that approval was via telecon, initial, date, and given by whom.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE60F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY

5.0 Leave

General Requirements All leave requests, personal annual leave (PAL), sick leave (SIL) and annual leave (AIL), shall be made through the Shift Manager. The SM Operations Manager or Operations Superintendent SHALL approve all leave. In accordance with the rules of the General Agreement, employees requesting PAL must notify the SM at least four (4) hours before the beginning of the PAL shift.

This will be strictly enforced. The Operations Clerk will only record leave as directed by the Shift Manager. The SM shall also complete Attachment 1.

5.1 Annual Leave for Trades and Labor, Salary Policy and Management Personnel in Operations 5.1.1 If one feels he/she must be off and can justify leave to the Operations Manager's/

Superintendent's office, he/she may request Absent Without Leave (AWOL) which the SM may grant. AWOL status should be used as this allows proper Operations management leave disposition. If placed on AWOL, Operations Manager/Superintendent's office must be contacted for disposition of the AWOL. The AWOL may be converted to AIL or unapproved absence, based upon Operations Management evaluation. Leave Without Pay (LWOP) is not to be granted by the SM.

5.1.2 Each shift shall have a minimum of six (6) AUOs for emergency response duties, five (5) qualified UOs, three (3) USs, and one (1) SM. This requirement can be met with regularly scheduled on-shift personnel or filled by calling overtime. PAL should be granted as long as this requirement is met.

5.1.3 IfPAL is granted, the Operations clerk will be notified to list it as such on the time sheet and the accumulated PAL listing. PAL shall not be combined with (either before or after) a TVA paid holiday.

5.2 Sick Leave for Trades and Labor Salary Policy and Management Personnel in Operations 5.2.1 Sick leave is an excellent form of insurance for employees to use for valid medical reasons. In order to ensure equitable administration of this benefit, the following guidelines should be followed.

5.2.2 The Operations Superintendent should review on a monthly basis the SIL used by all employees in Operations. Each individual's situation should be reviewed on a case-by-case basis to determine if any S/L abuse exists. Some of the items to be taken into consideration include:

1. Any effect being on a rotating shift has on an individual.

2. The locations of the employee's residence and the doctor's office.

3. The negative impact of an employee feeling compelled to report for work when it is not prudent; i.e., the employee may have a contagious condition.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE70F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY 5.2.3 Patterns of SIL use which indicate potential misuse should be discussed with the employee in an informal meeting to determine the actual cause. If, after this meeting to obtain information, additional discussions are needed, then a formal meeting will be held in which the employee's representative should be present. If these meetings do not achieve the desired result, formal disciplinary action may be initiated. Formal disciplinary action shall not begin until after these two meetings have taken place. SIL which is substantiated by appropriate medical evidence is not a basis for disciplinary action.

5.2.4 One of the greatest benefits to employees of prudent use of S/L is having an adequate reserve of leave to ensure continuity of regular pay in the event of a serious illness or accident. An employee's leave record may affect approval of advanced S/L.

5.2.5 S/L is granted for use consistent with the instructions in BP-114 and Personnel Manual Instruction, Section 3, which states that S/L is granted when an employee:

a. is incapacitated for duties because of sickness, injury, or pregnancy and confinement;

b. receives medical, dental, or optical exam or treatment;

c. is required to give care and attendance to a member of his or her immediate family who is afflicted with a quarantinable communicable disease; or

) d. would jeopardize the health of others by being present at his or her post of duty because of exposure to a quarantinable communicable disease. (Quarantinable communicable disease is identified by Executive Order 12452 and currently includes such diseases as Cholera, Diphtheria, Infectious Tuberculosis, Plague, Smallpox, Yellow Fever and suspected viral hemorrhagic fevers such as Lassa Margurg, Ebola, Congo Crimean and others not yet isolated or named.)

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE80F8 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY 6.0 Operator Relief When no qualified Unit Operator is available to relieve the Unit 1 Control Room, Unit 2 or 3 Board or Desk positions, the following guidelines shall be used for relieflbreaks:

6.1 ReliefslBreaks should only be for a short duration, 15 minutes or less.

6.2 There will always be 2 Reactor Operators in Unit's 1 and 2 Control Rooms and 2 Reactor Operators in Unit 3 Control Room (except as noted in 6.7).

6.3 The Unit 2 Desk Operator will be the relief Operator. The Unit 2 Desk Operator will relieve the Unit 1 Operator, the Unit 2 Board Operator, and the Unit 3 Desk Operator.

The expectation is that there will always be a Unit 1 Board Operator unless he/she is required to report to Unit 2 during a transient. When the Unit 2 Desk Operator relieves the Unit 1 Operator, the Operator will be assigned to the Unit 1 board with the expectation that the Operator will be available to report to Unit 2 as needed. There will always be a Unit 2 Board Operator.

6.4 To maintain consistency on the Unit 3 board, the Unit 3 Desk Operator is the relief for the Unit 3 Board Operator. This only happens when the Unit 2 Desk Operator relieves the Unit 3 Desk Operator, and in turn, the Unit 3 Desk Operator then relieves the Unit 3 Board Operator.

6.5 Turnovers shall be thorough, complete, and include the standard announcement.

6.6 The Unit 2 and 3 Control Room SROs will approve of the relief prior to it's initiation, and can deny the relief as he deems necessary.

6.7 Managers are not to relieve Reactor Operators unless an emergency exists. However, managers may be assigned to the Control Room for additional oversight if it becomes necessary to reduce the Control Room Reactor Operator compliment by one Reactor Operator without having a relief as previously described. ill this case, there would be no official assumption of control manipulation, or log entry duties, by the manager.

6.8 The relieved operator will make the affected Control Room SRO aware of his destination, and will report back to his assigned work station immediately upon being summoned, or hearing a Public Address (P A) announcement that their Unit has scrammed or the unit conditions have degraded.

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'~,- ...

OPERATIONS OSIL-25 ATTACHMENT 1 DAILY LEAVE REPORT Page 1 of 1 08/29/05 DATE SHIFT (check one) LEAVE (check one)

NAME MID DA EVE AIL PAL SIL AWOL REMARKS

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE 1 OF3 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY Attachment 2 Instructions for filling out the Call-in Request Sheet

1) The Unit Operator and / or the Operations Clerk will assign the number of positions required to be filled for the shift in question. This will encompass the required positions and number required in each position including extra personnel required to support shift activities.

2) Shift Manager signs (signature) the call-in request sheet prior to initiating the call-in signifying he concurs with the positions and the number of persons required to fill the shift compliment. This can include any additional personnel required to support extra shift tasks.

3) Columns will be filled out in "YESINO" format using the following criteria;

WORK., "Do you want to work the required shift"? This is to determine whether the individual wants to work the entire shift.

WAIVER, "Will you require a waiver to work the entire shift"?

FIT FOR DUTY, "Are you fit for duty"? (See Fitness For Duty Below)

ALCOHOL, "Have you consumed alcohol in the past 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months "?( See Fitness For Duty Below)

INITIALS, The Unit Operator (caller) initials in the row for the individual which has been called.

4) The Shift Manager and the Unit Operator (or Operations Clerk) will then sign the bottom ofthe Call-in request and forward the sheet to the Operations Clerk. The Operations clerk will file the Call-in request in a fire proof cabinet for the required retention period.

FITNESS FOR DUTY The criteria listed below is to be used to determine the correct methodology in determining Fit For Duty status.

An employee is expected to not consume alcohol 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months prior to reporting for SCHEDULED work and to report fit and within FFD guidelines. If called for unscheduled work the employee's suitability for work must be determined.

A. The following must be done whenever a worker is being called in for unscheduled work.

1. The caller will ask and will document on a call-in request sheet the worker's response to the following two questions.

a. Are you fit to report to work?

b. Have you consumed alcohol within the past 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ?

If the answer to the last question is yes, and the individual is called in, document how much alcohol was consumed and when.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE20F3 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY Attachment 2 FITNESS FOR DUTY (Cont))

2. The worker must advise the caller and the supervisor if he or she believes that he/she is unfit to report for work.

3. The caller will then decide whether or not to have the person report to work.

4. If the answer to the alcohol consumption question is "yes" then Nuclear Security on site should be notified and be requested to administer a saliva test. This test must be administered as soon as the person arrives on site.

5. If the test results are 0.039 or below, the supervisor shall determine if the employee can be permitted to work. The employee will not be subject to disciplinary action.

6. If the results are 0.040 or above the employee will not be permitted to work. The worker can be paid callout pay. This will NOT be considered a positive test for FFD purposes.

B. Emergency Response Center Personnel Emergency Response Center Personnel who are called by an automated electronic system are responsible for:

1. Advising the center if he/she believes that they are unfit to report for duty.

2. Advising the center on reporting if alcohol has been consumed within the past five hours.

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE 3 OF3 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05

') OVERTIME, LEAVE, AND RELIEF POLICY Attachment 2 Call-in Request Form Shift Date,_ _ _ _ __

Number of positions US UO AUO STA-r-,- - - - - SSSr------- 1st res-p-o-n-dT"e-r-s------

SM

---~(~Sl~*g-n-a~tu-r-e)~----

List T &L for call-in by OT hours (list those requiring a waiver last)

If the answer to the alcohol consumption question is "yes" Nuclear Security on site should be notified and requested to administer a saliva test to the employee. Document how much alcohol was consumed and when. This test should be administered when the person arrives on-site.

Du~OfficiaIComments: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

SM Review: _ _ _ _ _ _ __ Operator Performing: Page _of_ _

(Signature) (Signature)

Retention Period:1 year Responsibility: Operations Shift Clerk

'""--,,,/ '~

TENNESsEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE 1 OF4 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY Attachment 3 AUO Dav Shift Overtime Sign Up Sheet Week Beginning: and Ending: _ _ __

Monday Tuesday Wednesday Thursday Friday Saturday Sunday Date Date Date Date Date Date Date

Needed # Needed # Needed # Needed # Needed # Needed # Needed Shift Manager Check if this is a Non-Training Week: _ _ _ _ _ _ __ Date:- - - - -

Shift Manager

TENNE~'0:5EE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE20F4 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY Attachment 3 AUO Night Shift Overtime Sign Up Sheet Week Beginning: and Ending: _ _ __

Monday Tuesday Wednesday Thursday Friday Saturday Sunday Date Date Date Date Date Date Date

Needed # Needed # Needed # Needed # Needed # Needed # Needed -

Shift Manager Check if this is a Non-Training Week: _ _ _ _ _ _ __ Date:, _ _ _ __

Shift Manager

TENNESsEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE30F4 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY Attachment 3 UO Day Shift Overtime Sign Up Sheet Week Beginning: and Ending: _ _ __

Monday Tuesday Wednesday Thursday Friday Saturday Sunday Date Date Date Date Date Date Date

Needed # Needed # Needed # Needed # Needed # Needed # Needed Shift Manager Check if this is a Non-Training Week: _ _ _ _ _ _ __ Date:. _ _ _ __

Shift Manager

TENNESSEE VALLEY AUTHORITY OSIL-25 BROWNS FERRY NUCLEAR PLANT PAGE 4 OF4 OPERATIONS SECTION INSTRUCTION LETTER 08/29/05 OVERTIME, LEAVE, AND RELIEF POLICY Attachment 3 UO Night Shift Overtime Sign Up Sheet Week Beginning: and Ending: _ _ __

Monday Tuesday Wednesday Thursday Friday Saturday Sunday Date Date Date Date Date Date Date

Needed # Needed # Needed # Needed # Needed # Needed # Needed Shift Manager Check if this is a Non-Training Week: _-----::-:--:-::-::-::--_ __ Date: _ _ _ __

Shift Manager

JPM NO. 510 REV. NO.3 PAGE 1 of 14 JPM NUMBER: 510 TITLE: EVALUATE RECOMBINER PERFORMANCE ADMIN: Equipment Control PROVIDE CANDIDATE WITH A COPY OF: 3-01-66, Rev 56 SUBMITTED BY: DATE: _ _ __

VALIDATED BY: DATE: _ _ __

APPROVED BY: DATE: _ _ __

TRAINING PLANTCONCURRENCE: _ _ _ _ _ _ _ _ _ ____ . DATE: _ _ _~

OPERATIONS

Examination JPMs Require Operations Training Manager Approval or Designee Approval and Plant Concurrence

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JPM NO. 510 REV. NO.3 PAGE 2 of 14 REVISION LOG Revision Effective Pages Description Number Date Affected Of Revision 0 08/08/02 All Initial issue 1 06/22/07 All Procedure revision 2 01/02/08 All General revision 3 08/07108 All General revision & re-format

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JPM NO. 510 REV. NO.3 PAGE 3 of 14 OPERATOR: _____________________________________________

RO SRO DATE: ____________

JPM NUMBER: 510 TASK TITLE: EVALUATE RECOMBINER PERFORMANCE KIA NUMBER: 2.1.7 KIA RATING: RO 4.4 SRO*4.7 TASK STANDARD: EVALUATE OFF-GAS RECOMBINER PERFORMANCE PERFORMANCE LOCATION: CLASSROOM: X REFERENCES/PROCEDURES NEEDED: 3-01-66, Rev 56 VALIDATION TIME: CLASSROOM: 12:00 MAX. TIME ALLOWED: _______ (FOR TIME CRITICAL JPMs ONLY)

PERFORMANCE TIME:

COMMENTS:

ADDITIONAL COMMENT SHEETS ATTACHED? YES ___ NO RESULTS: SATISFACTORY UNSATISFACTORY EXAMINER SIGNATURE: DATE: ____________

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JPM NO. 510 REV. NO. 3 PAGE 4 of 14 Classroom INITIAL CONDITIONS: You are the desk operator. A startup is in progress on Unit 3 and reactor power has b~en raised to 99% rated thermal power. The Hydrogen Water Chemistry System is out of service lAW 3-01-4. Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with Section 5.0. The operating steam jet is operating properly.

INITIATING CUES: The Shift Operations Supervisor directs you to evaluate Off-Gas Recombiner 3A performance in accordance with 3-01-66, Section 6.1.

3-TI-66-75A 393°F 3-TI-66-75B 320 OF 3-TRS-66-77 A Center temp 618 OF 3-TRS-66-77B Center temp 380 OF Rx Power Thermal 3430 mwth 3-H2R-66-96 operable both pens reading .24% H2

JPM NO. 510 REV. NO.3 PAGE 5 of 14 START TIME _ _ _ _ __

6.1 Recombiner Performance Evaluation NOTES

1) The production of hydrogen and oxygen in the reactor is dependent upon reactor power level and upon the amount of hydrogen injected by the Hydrogen Water Chemistry System if in service. Since the recombination of hydrogen and oxygen is exothermic, the operating temperature of the recombiner is also dependent upon power level and the status of the HWC System.

2) Following startup, while still at low power, recombiner performance and hydrogen concentration should be closely monitored.

PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[1] PERFORM a recombiner performance evaluation as follows:

[1.1] DETERMINE the in-service recombiner inlet temperature as indicated on RECOMBINER 3A(3B), INLET TEMP 3-TI-66-75A(B), Panel 3-9-53.

STANDARD:

Determined Recombiner 3A inlet temp 3-TI-66-75A, Panel 3-9-53 from handout.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 510 REV. NO.3 PAGE 6 of 14

')

PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[1.2] DETERMINE the in-service recombiner operating (center) temperature as indicated on RECOMBINER 3A13B TEMPERATURE recorder, 3-TRS-66-77, Panel 3-9-53.

STANDARD:

Determined the in-service recombiner operating (center) temperature as indicated on Recombiner 3A temperature recorder, 3-TRS-66-77, Panel 3-9-53 (from handout).

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

PERFORMANCE STEP: CRITICAL x NOT CRITICAL

[1.3] CALCULATE the temperature difference (/1T) between

)

/ the values obtained in Steps 6.1 [1] and 6.1 [2].

STANDARD:

Calculated Recombiner 3A inlet/center ~t and determined ~t is 225 of SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

)

,f

JPM NO. 510 REV. NO.3 PAGE 7 of 14 PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[1.4] DETERMINE the reactor thermal power (MWt) from process computer.

STANDARD:

Determined reactor thermal power from the handout.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[1.5] USING Illustration 1, PLOT the corresponding point of reactor power in MWt and IJ.T.

STANDARD:

Using illustration 1, Determined llt corresponding to 3430 MWT is 240.1 OF.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 510 REV. NO.3 PAGE 8 of 14 PERFORMANCE STEP: CRITICAL x NOT CRITICAL

[1.6] VERIFY point on illustration 1 is above or equal to the appropriate line (HWC in service or HWC out of service)

STANDARD:

Determines from Illustration 1 that calculated i1t vs MWt plots BELOW the HWC Out of Service (solid) line (Critical).

Stops task performance and informs SRO that acceptance criteria is NOT met. (Not Critical)

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ _"'--

JPM NO. 510 REV. NO.3 PAGE 9 of 14 PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[2] IF the in-service recombiner performance is below the minimum allowable, THEN:

[2.1] CHECK Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with Section 5.0.

STANDARD:

N/A - Given in initial conditions.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[2.2] MONITOR the OFFGAS HYDROGEN ANALYZER recorder, 3-H2R-66-96 on Panel 3-9-53.

STANDARD:

Monitors the Offgas Hydrogen Analyzer recorder, 3-H2R-66-96 on Panel 3-9-53 (from handout both points operable and reading .24% H2.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 510 REV. NO.3 PAGE 10 of 14 PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[2.3] IF both hydrogen analyzers are inoperable, THEN NOTIFY Chemistry to obtain a grab sample to determine hydrogen concentration.

STANDARD:

Determined Step [2.3] is N/A.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

PERFORMANCE STEP: CRITICAL x NOT CRITICAL

[2.4] IF a malfunction of the SJAE is suspected, THEN

,)

REFER TO Section 8.4 and TRANSFER SJAEs.

STANDARD:

Given in initial conditions that in service steam jet is operating properly and does NOT transfer SJAEs (Not Critical Unless determines to swap SJAE).

SAT UNSAT N/A _ _ COMMENTS:. _ _ _ _ __

)

JPM NO. 510 REV. NO.3 PAGE11of14 PERFORMANCE STEP: CRITICAL NOT CRITICAL x

[3] IF off-gas hydrogen rises above 1%, THEN REFER TO 3-AOI-66-1.

STANDARD:

Determines off-gas hydrogen has NOT risen above 1% from handout.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

CAUTION Off-Gas System valves are potentially spark-producing when operated; therefore, WHEN hydrogen concentration is suspected of being greater than 4%, THEN DO NOT TAKE any action that will change off-gas valve positions until after the unit is shut down.

JPM NO. 510 REV. NO. 3 PAGE 12 of 14

')

PERFORMANCE STEP: CRITICAL x NOT CRITICAL

[4] IF analysis or hydrogen analyzers show hydrogen concentration is below 4%, THEN PLACE standby recombiner in operation.

REFER TO Section 8.3.

STANDARD:

Determines the required action is to PLACE standby recombiner in operation lAW Section 8.3.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

)

END OF TASK STOP TIME_ _ _ __

)

JPM NO. 510 REV. NO.3 PAGE 13 of 14 11.lustratlon1

(:Page 101 1, Recombine Pefforma:nce Eva:l:ruation - AT to Reactor Power

~+i------~------+-----~-------r------

100+1----+---~--

DeJtlt.iPnmr*~

Mm

)

dSeniice ...... ,* HWC Evaluation satisfactory when intersedlon point of AT Reador Is above the appropriate line.

For 3458m'Nt HVtJClnservlce AT ;;:1 HV'/C out of service AT ;;:

CUBYEEACTORS Normal Water Chemistry (NWC) AT =O.070 F per MWt rt rt ...... R"' ... V.Jater Chemistry (HWC) AT:::: U.U;;¥;;I per M\llJt

PAGE 14 of 14 Classroom INITIAL CONDITIONS: You are the desk operator. A startup is in progress on Unit 3 and reactor power has been raised to 99% rated thermal power. The Hydrogen Water Chemistry System is out of service lAW 3-01-4. Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with Section 5.0. The operating steam jet is operating properly.

INITIATING CUES: The Shift Operations Supervisor directs you to evaluate Off-Gas Recombiner 3A performance in accordance with 3-01-66, Section 6.1.

3-TI-66-75A 393 of 3-TI-66-75B 320 of 3-TRS-66-77 A Center temp 618 of 3-TRS-66-77B Center temp 380 of Rx Power Thermal 3430 mwth 3-H2R-66-96 operable both pens reading .24% H2

(

BFN Off-Gas System 3-01-66 Unit 3 Rev. 0056 P~9~_ 50 of ~ ~ § _______

6.0 SYSTEM OPERATIONS 6.1 Recombiner Performance Evaluation NOTES

1) The production of hydrogen and oxygen in the reactor is dependent upon reactor power level and upon the amount of hydrogen injected by the Hydrogen Water Chemistry System if in service. Since the recombination of hydrogen and oxygen is exothermic, the operating temperature of the recombiner is also dependent upon power level and the status of the HWC System.

2) Following startup, while still at low power, recombiner performance and hydrogen concentration should be closely monitored.

[1] PERFORM a recombiner performance evaluation as follows:

[1.1 ] DETERMINE the in-service recombiner inlet temperature as indicated on RECOMBINER 3A(3B), INLET TEMP 3-TI-66-75A(B), Panel 3-9-53. o

[1.2] DETERMINE the in-service recombiner operating (center) temperature as indicated on RECOMBINER 3A13B TEMPERATURE recorder, 3-TRS-66-77, Panel 3-9-53. o

[1.3] CALCULATE the temperature difference (~T) between the values obtained in Steps 6.1 [1] and 6.1 [2]. o

[1.4] DETERMINE the reactor thermal power (MWt) from process computer. o

[1.5] USING Illustration 1, PLOT the corresponding point of reactor power in MWt and ~T. o

[1.6] VERIFY point on illustration 1 is above or equal to the appropriate line (HWC in service or HWC out of service) o

BFN Off-Gas System 3-01-66 Unit 3 Rev. 0056 Page 51 of 115 6.1 Recombiner Performance Evaluation (continued)

[2] IF the in-service recombiner performance is below the minimum allowable, THEN:

[2.1 ] CHECK Off-Gas Preheater, Recombiner and SJAEs are in operation in accordance with Section 5.0. D

[2.2] MONITOR the OFFGAS HYDROGEN ANALYZER recorder, 3-H2R-66-96 on Panel 3-9-53. D

[2.3] IF both hydrogen analyzers are inoperable, THEN NOTIFY Chemistry to obtain a grab sample to determine hydrogen concentration. D

[2.4] IF a malfunction of the SJAE is suspected, THEN REFER TO Section 8.4 and TRANSFER SJAEs. D

[3] IF off-gas hydrogen rises above 1%, THEN REFER TO 3-AOI-66-1. D CAUTION Off-Gas System valves are potentially spark-producing when operated; therefore, WHEN hydrogen concentration is suspected of being greater than 4%, THEN DO NOT TAKE any action that will change off-gas valve positions until after the unit is shut down.

[4] IF analysis or hydrogen analyzers show hydrogen concentration is below 4%, THEN PLACE standby recombiner in operation.

REFER TO Section 8.3. D

JPM NO. 511 REV. NO.2 PAGE 1 of?

JPM NUMBER: 511 TITLE: RADCON DOSE LIMITS ADMIN: Radiation Control PROVIDE CANDIDATE WITH A COPY OF: SPP-5.1 (Only If Requested)

SUBMITTED BY: DATE: _ _ __

VALIDATED BY: DATE: _ _ __

APPROVED BY: DATE: _ _ __

TRAINING PLANTCONCURRENCE: _____________________ DATE: _____

OPERATIONS

Examination JPMs Require Operation,s Training Manager Approval or Designee Approval and Plant Concurrence

JPM NO. 511 REV. NO.2 PAGE 2 of?

REVISION LOG Revision Effective Pages Description Number Date Affected Of Revision 0 10/31/02 All Initial issue 1 09/01/05 3 Procedure revision 2 08/08/08 All General revision and re-format

)

JPM NO. 511 REV. NO.2 PAGE 3 of 7 OPERATOR: _____________________________________________

RO SRO DATE: ___________

JPM NUMBER: 511 TASK TITLE: RADCON DOSE LIMITS KIA NUMBER: 2.3.4 KIA RATING: RO 3.2 SRO 3.7 TASK STANDARD: Given circumstances, determine the dose limitation for declared and undeclared pregnant female employees and their eligibility for overtime.

PERFORMANCE LOCATION: CLASSROOM: X REFERENCES/PROCEDURES NEEDED: SPP-5.1, Rev 6 VALIDATION TIME: CLASSROOM: _ _ __

MAX. TIME ALLOWED: _ _ _ _ _ (FOR TIME CRITICAL JPMs ONLY)

PERFORMANCE TIME:

COMMENTS:

ADDITIONAL COMMENT SHEETS ATTACHED? YES _ _ NO RESULTS: SATISFACTORY UNSATISFACTORY EXAMINER SIGNATURE: DATE: _ _ _ _ _ ___

JPM NO. 511 REV. NO.2 PAGE 4 of 7 Classroom INITIAL CONDITIONS: You are the Shift Manager (SRO) in charge for the current shift OR you are the Unit 1 Operator (RO) making the callout.

Two AUO overtime slots are available on the next shift and the first two eligible individuals are female employees. The overtime slot involves a high exposure area of the turbine building, RADCON reports the job will require approximately 100 mrem for each of the two employees. One of the two individual (AUO 1) reports that she is pregnant and has just found out. She wishes to declare her pregnancy and request counseling by RADCON. She also informs you that ""the other female AUO (AUO 2) is also pregnant" but you need to talk with her because of a reluctance to declare her pregnancy because she needs the money. When you talk with her (AUO 2), she confirms she is pregnant and does not wish to participate in the RADCON program for pregnant women.

INITIATING CUES: The examiner will ask a series of questions about the situation above, (provide participant handout.)

JPM NO. 511 REV. NO.2 PAGE50f7 EXAMINER COPY KEY DO NOT HANDOUT TO STUDENT You are the Shift Manager or Unit 1 Unit Operator for the current shift. Two AUO overtime slots are available on the next shift and the first two eligible individuals are female employees. The overtime slot involves a high exposure area of the turbine building, RADCON reports the job will require approximately 100 mrem for each of the two employees. One of the two individual (AUO 1) reports that she is pregnant and has just found out. She wishes to declare her pregnancy and request counseling by RADCON. She also informs you that 'the other female AUO (AUO 2) is also pregnant" but you need to talk with her because of a reluctance to declare her pregnancy because she needed the money. When you talk with her (AUO 2), she confirms she is pregnant and does not wish to participate in the RADCON program for pregnant women.

What will be the dose limit for AU01?

500 mrem /9 month gestation, 50 mrem / month (Critical)

Given the situation, can AU01 be hired for the expected job?

No, the exposure would be too high on the job she would be expected to do i.e.,

exceed the monthly limit (Critical). However, she may be hired for another job of low dose, replacing a regular scheduled AUO who can be moved to the High Rad job (Not Critical).

What will be the dose limit for AU02?

Normal exposure limits apply (Critical)

Given the situation, can AU02 be hired for the expected job?

Yes (Critical)

NOTE to Examiner: May have to ask directed question to assess the knowledge items above especially for the 9 month and monthly limits on the first question, 3 of the 4 are required for successful completion. The source of this requirement is SPP 5.1 and 10CFR20, Prenatal Exposure and Declaration of Pregnancy definition. Exact wording not required.

PAGE 6 of?

Classroom INITIAL CONDITIONS: You are the Shift Manager (SRO) in charge for the current shift OR you are the Unit 1 Operator (RO) making the callout.

Two AUOovertime slots are available on the next shift and the first two eligible individuals are female employees. The overtime slot involves a high exposure area of the turbine building, RADCON reports the job will require approximately 100 mrem for each of the two employees. One of the two individual (AUO 1) reports that she is pregnant and has just found out. She wishes to declare her pregnancy and request counseling by RADCON. She also informs you that ""the other female AUO (AUO 2) is also pregnant" but you need to talk with her because of a reluctance to declare her pregnancy because she needs the money. When you talk with her (AUO 2), she confirms she is pregnant and does not wish to participate in the RADCON program for pregnant women.

) INITIATING CUES: The examiner will ask a series of questions about the situation above, (provide participant handout.)

PAGE? of?

Handout to Applicant You are the Shift Manager or the Unit 1 Unit Operator for the currentshift. Two AUO overtime slots are available on the next shift and the first two eligible individuals are female employees. The overtime slot involves a high exposure area of the turbine building, RADCON reports the job will require approximately 100 mrem for each of the two employees. One of the two individual (AUO 1) reports that she is pregnant and has just found out. She wishes to declare her pregnancy and request counseling by RADCON.

She also informs you that the other female AUO (AUO 2) is also pregnant but you need to talk with her because of a reluctance to declare her pregnancy because she needs the money. When you talk with her (AUO 2), she confirms she is pregnant and does not wish to participate in the RADCON program for pregnant women.

What will be the dose limit for AU01 ?

Given the situation, can AU01 be hired for the expected job? (If not, Why?)

) What will be the dose limit for AU02?

Given the situation, can AU02 be hired for the expected job? (If not, Why?)

)

(

TITLE SPP-S.1 mil Radiological Controls Rev. 0006 Page 1 of 46 Quality Related o Yes D No NPG Standard PORC Required DYes 0No Programs and Processes Effective Date 10-17-2007 (COC,BFN,WBN) hatef: (SQN) 03-03-2008 Responsible Peer Team: Radiological and Chemistry Control Concurred by: J. M. Corey 10/1/07

Primary Sponsor Date Timothy P. Cleary 10/4/07 Concurred by:

Peer Team Mentor Date N/A N/A Approved by:

General Manager, NA Date James R. Douet 10/9/07 Approved by: *Vice President, Nuclear Support Date

Site-specific changes are approved by Site Sponsor and Site Vice President (see PCF) --I

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 2 of 46 Revision Log Revision or Affected Change Effective Page Number Date Numbers Description of Revision/Change 0 N/A All Initial issue. Replaces TVAN STD-S.1, BFN SSP-S.1, and WBN SSP-S.01. This revision was not issued and therefore has never been effective at any site.

1 07-31-98 2,4-7, 10-13, SubSection 3.4.1.S.A modified to reflect these 1S, 17-19,21, values are controlled through quality-related code.

23, 26, 28-29 Added SubSection 3.4.1.6 to include Administrative Dose Levels and added definition for Administrative Dose Levels. SubSection 3.3.1.C.2 changed site-specific to TVAN-specific. SubSection 3.S.3 General Note changed for clarification and S.O Contaminated Area definition corrected. Updated position titles and made minor editorial changes.

2 10-30-98 2,4-16, Changes made to incorporate revision to 10 CFR 18-24,26-27, 20 [Federal Register: July 23, 1998 (Volume 63, 29-32 Number 141 )]. Corrective actions for CHPER98006S addressing the requirements of the Privacy Act with respect to Radiological Exposure Records and Land Use Surveys and minor editorial changes for clarification.

3 1/10/2000 2,20,23,27 Changes made: to make internal dose reporting consistent with HIS-20, remove reference to the CINDY code, and indicate that RWP waivers are controlled by RCDP-3.

4 6/4/02 2, 10, 1S, 32 Changes made to reflect programmatic changes in the method of use for reporting and documenting Personnel Contamination. Implement revision to 10CFR20.1 003, 20.1201 (a)(2)(ii), and 20.1201 (c) regarding the shallow dose equivalent limit.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 3 of 46 Revision Log Revision or Affected Change Effective Page Number Date Numbers Description of Revision/Change 5 11/12/03 2, 3, 5-23, 25, Changes made to delete references to the site 28,32 Radiological and Chemistry Manager, conferring responsibilities to RADCON Superintendents/RSO.

Clarify that training requirements are implemented in TRN-2. Address corrective action to PER 03-000227-000 to include responsibilities, expectations, and documentation requirements for fulfilling the 10CFR20.11 01 (c) requirement for i annual program review. Address the corrective action to PER 02-000327-000 with regard to internal dose assessments and thresholds for bioassays. Specified a 1 DAC threshold for tracking DAC-hrs. Clarified controls for contamination area. Made requirements for prior occupational dose documentation consistent with RCDP-4. Clarified response to dosimeter alarms, how dosimetry is to be worn, and program elements of the respiratory protection program. Deleted requirements for pre-natal counseling and program withdrawal statements to be included in woman's exposure history record. Deleted statement on method of contamination determination on respirators. Deleted RADCON requirement to evaluate non-rad (e.g., psychological, industrial) hazards for respiratory protection equipment.

Deleted statement that doses determined by DIR are reported at a 10 mrem threshold.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes -----------

E~g~ 4of46 Revision Log Revision or Affected Change Effective Page Number Date Numbers Description of Revision/Change 6 10/17/07 All This document has been converted from Word 95 (COC,BFN, to Word 2002 (XP) using Rev. 5. Removed PORC WBN) review requirement.

4, 7-28, 30-46 Administrative changes to reflect organizational hateF (SON) revision, corrective actions following WBN INPO 03/03/08 E&A and NA Audit, and Strategic Objective 1.8 Dose Reduction Team recommendations. Clarify labeling requirement (PER Action 73866-001).

Revise required actions based upon bioassay results section incorporating a Screening Level and an Evaluation Level. Clarify Annual review of RP program (PER Action 91438-002). Added review of program elements in accordance with INPO 05-008 (PER Action 93786-007). Incorporating aspects of INPO 05-008. Added section 6.0 references; canceled SPP-5.1 REO & REF document and returned contents to the procedure. Added PC recording criteria of EPRI document 1011740 (PER Action 92762 -006) and PC Tracking (PER Action 78100-001). Added guidance for free release of materials and individuals (PER Actions 86078-009 and 86078-010). Changed Radiological Program Manager to RP peer team under the responsibilities section. Added requirement for secondary dosimeter evaluation with each multi-badge where possible (PER Action 115069-003). Minor editorial changes. Replaced "TVAN" with Nuclear Power Group (NPG).

)

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page S of 4_6_

Table of Contents 1.0 PURPOSE ................................................................................................................................. 7 2.0 SCOPE ...................................................................................................................................... 7 3.0 INSTRUCTIONS ........................................................................................................................ 7 3.1 Responsibilities .......................................................................................................................... 7 3.1.1 Chief Nuclear Officer and Executive Vice President .................................................. 7 3.1.2 Vice President, Nuclear Support ................................................................................ 7 3.1.3 Radiation Protection Peer Team ................................................................................ 7 3.1.4 NPG Organization ...................................................................................................... 7 3.1.5 Radiation Protection Manager (RPM) IRadiation Safety Officer (RSO)* ........................................................................................................................ 8 3.1.6 Supervisors* ............................................................................................................... 8 3.1.7 NPG Employees* ........................................................................................................ 8 3.1.8 RP Personnel ............................................................................................................. 8 3.2 Program Monitoring Evaluation and Oversight* ......................................................................... 8 3.3 Radiation Protection Qualifications and Training ....................................................................... 9 3.3.1 Radiation Protection Training and Retraining* ........................................................... 9 3.3.2 Qualifications and Training for RP Personnel ............................................................. 9 3.4 Exposure Control ..................................................................................................................... 10 3.4.1 External Exposure Control* ...................................................................................... 10 3.4.2 Emergency Exposure Guidance and Planned Special Exposures (PSE)* ....................................................................................................................... 20 3.4.3 Internal Exposure Control* ....................................................................................... 21 3.4.4 Radiation Exposure Tracking, Recording, and Reporting ......................................... 26 3.5 Radiologically Controlled Areas (RCAs)* ................................................................................. 28 3.5.1 Posting* .................................................................................................................... 28 3.5.2 Surveys* ................................................................................................................... 28 3.5.3 Limits* ....................................................................................................................... 29 3.5.4 General Requirements* ............................................................................................ 32 3.5.5 Area-Specific Requirements* ................................................................................... 33 3.6 Radiation Work Permit System ................................................................................................ 34 3.6.1 Control of Work in the RCA ...................................................................................... 34

) 3.6.2 RWP Criteria ............................................................................................................. 34 3.6.3 Waived Requirements for RWPs .............................................................................. 34

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 6 of 46 Table of Contents (continued) 3.6.4 RWP Compliance ..................................................................................................... 34 3.7 Radioactive Material Control* .................................................................................................. 35 3.7.1 Packaging, Handling, and Transfer .......................................................................... 35 3.7.2 Labeling* ................................................................................................................... 35 3.7.3 Receipt* .................................................................................................................... 35 3.7.4 General Requirements* ............................................................................................ 35 3.8 Incident Reporting and Investigation ........................................................................................ 35 3.9 RP Instruments ........................................................................................................................ 36 3.9.1 Calibration* ............................................................................................................... 36 3.9.2 Procurement* ............................................................................................................ 36 3.9.3 Control and Use* ...................................................................................................... 37 3.10 Radiation Protection Quality Assurance Requirements ........................................................... 37 4.0 RECORDS ............................................................................................................................... 37 s.o DEFINITIONS* ......................................................................................................................... 37 5.1 Acronyms/Abbreviations .......................................................................................................... 42

6.0 REFERENCES

........................................................................................................................ 43 6.1 Source Documents .................................................................................................................. 43 6.2 Developmental References ...................................................................................................... 44

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 L-~ocesses_ .. Page 7 of 46 1.0 PURPOSE This procedure details specific requirements and management controls for radiation protection.

2.0 SCOPE The requirements of this document apply to all aspects of radiation protection associated with TVA Nuclear Power Group (NPG) 10 CFR Part 50 licensed nuclear facilities. This document states the minimum requirements for the Radiation Protection (RP) program at the nuclear sites. The NPG RP Program fulfills the requirements of 10 CFR 19, 20, and 30 through 34. The RP Program is further established to meet, to the extent practicable, the guidelines contained in INPO 05-008, and ANI Inspection Criteria 8.1 through 8.10. In addition, NPG non-nuclear plant facilities/operations are subject to the requirements in this document marked with an asterisk (*).

3.0 INSTRUCTIONS 3.1 Responsibilities 3.1.1 Chief Nuclear Officer and Executive Vice President The Chief Nuclear Officer and Executive Vice President is responsible for establishing policy, requirements, and management controls for radiation protection within the NPG consistent with TVA policies, and providing resources necessary for program development, maintenance, and implementation. The Chief Nuclear Officer and Executive Vice President shall periodically review the status of the program for conformance with established policy and performance goals. If problem areas are identified, the Chief Nuclear Officer and Executive Vice President will initiate corrective action.

3.1.2 Vice President, Nuclear Support The Vice President, Nuclear Support in this procedure, issues radiation protection requirements, and management controls.

3.1.3 Radiation Protection Peer Team The Radiation Protection Peer Team is responsible for planning, developing, reviewing and coordinating radiation protection policy, requirements, and management controls for the NPG. The peer team shall interpret policy and provide clarification as necessary. The peer team shall conduct assessments and shall provide technical assistance and operational support to the nuclear sites as requested.

3.1.4 NPG Organization Each organization within NPG shall implement the requirements of this procedure as applicable to the design, construction, operation, maintenance, and modification of NPG's power plants and facilities.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 8 of 46 3.1.5 Radiation Protection Manager (RPM) IRadiation Safety Officer (RSO)*

The RPM/RSO or designee is responsible for implementation, development and direction of the site/facility Radiation Protection program in accordance with applicable regulations and NPG requirements. They shall use, to the extent practicable, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (ALARA).

3.1.6 Supervisors*

All supervisors of personnel who work in radiologically-controlled areas must ensure that employees are aware of and comply with radiation protection requirements. Supervisors are responsible to coach workers in dose reduction techniques, equalizing work group dose, and monitoring personnel in-field performance.

3.1.7 NPG Employees*

Each NPG employee is responsible for performing his / her work in a manner that will protect themselves, fellow employees, and the general public from unnecessary exposure to radiation and radioactive material.

3.1.8 RP Personnel RP personnel shall have the responsibility and authority to stop or prevent the initiation of a job, test, or any work activity involving radiological protection if continued performance of the work would result in the violation of regulation, policy, or plant procedure; or would endanger the safety of personnel and such actions are consistent with plant safety. RP personnel shall immediately inform the Shift Manager and the Radiation Protection Manager, or his /

her designated alternate, of their actions.

3.2 Program Monitoring Evaluation and Oversight*

The RPM /Radiation Safety Officer (RSO) shall ensure that the Radiation Protection Program is reviewed annually. Non nuclear plant facilities licensees shall conduct the program review in accordance with applicable NUREG and/or procedure requirements.

Nuclear power plant radiation protection programs shall conduct the review as follows:

A. The review shall be conducted at least once every 12 months, on a schedule that covers all phases of the program on a 2-3 year review cycle.

B. Major phases of the program should include:

1. Leadership and management

2. Dose Control

3. Cumulative Dose Reduction

4. Contamination Monitoring and Control

5. Control of Radioactive Material

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 9 of 46 3.2 Program Monitoring Evaluation and Oversight* (continued)

6. Trained, Skilled and Experienced Radiation Protection Personnel and Radiation Workers C. The review shall be conducted by qualified persons who are knowledgeable of the onsite program and when practical, should be performed by personnel who do not have direct responsibility over the program.

D. The review shall, as a minimum, cover regulatory requirements, procedural compliance, technical adequacy, implementation, and effectiveness of each reviewed program phase.

E. Regulatory Guides, EPRI guidance, NEI, ANI, and INPO documents are sources that should be reviewed when assessing the program elements.

F. The review should look at Self-Assessments, INPO reviews, Integrated Trend Reports (ITR) or other documented reviews.

G. The review shall be documented and retained as a Facility-based Radiological Control Program record (SPP-5.9, "Radiological Control and Radioactive Material Shipment Augmented Quality Assurance Program").

3.3 Radiation Protection Qualifications and Training 3.3.1 Radiation Protection Training and Retraining*

A. A radiation protection training program shall be developed, documented, and administered consistent with expectations as outlined in NEI 95-04, "Guideline for General Access Training". This program is implemented in TRN 2, "General Employee Training" for NPG power plant facilities.

B. NPG non-nuclear plant facilities I operations shall implement a radiation protection training program in accordance with applicable license and procedure requirements.

C. All individuals who in the course of employment are likely to receive an occupational exposure to radiation from licensed and unlicensed radiation.sources under the control of the licensee in excess of 100 mrem in a year shall receive radiation protection training commensurate with their duties and responsibilities (10 CFR 19.12) and instructions on U.S. NRC Regulatory Guides 8.13 and 8.29.

3.3.2 Qualifications and Training for RP Personnel A. The education, training, and experience of RP personnel, excluding the Radiation Protection Manager, shall be in accordance with ANSI N18.1-1971.

B. The site Radiation Protection Manager shall have the education and experience as described in Regulatory Guide 1.8, Revisions 1 and 2 in the context of Regulatory Guide 1.8 and the endorsed ANSI N18.1-1971 and ANSI/ANS-3.1-1981. Because of TVA's commitment to both documents, the Radiation Protection Manager must meet the more restrictive of the composite qualifications and training of both documents.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 10 of 46 3.3.2 Qualifications and Training for RP Personnel {continued}

C. The Radiation Protection Manager shall have a bachelor's degree in a science or engineering subject, including formal training in radiation protection. At the time of initial core loading or appointment to the active position, whichever is later, the responsible individual shall have five years of experience in applied radiation protection. At least three of the five years shall be professional-level experience in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power plants, preferably in a nuclear power plant. During the three years, the individual shall participate in the radiation protection section of an operating nuclear power plant during the following periods: (1) routine refueling outage (one to two months); and (2) two months operation above 20 percent power. The Radiation Protection Manager shall have at least six months experience onsite (See Section 5.0 Definitions for clarification). Individuals who do not fully meet the literal requirements for the position may be temporarily assigned to fill that position.

Such assignments shall be justified and a time for the temporary assignment specified and documented. Temporary assignments shall not reduce the collective experience requirements specified for the level.

D. If the Radiation Protection Manager is temporarily replaced, the following shall apply:

The individual who temporarily replaces the Radiation Protection Manager shall have a bachelor's degree in a science or engineering subject and two years experience, one of which shall be nuclear power plant experience. Six months experience shall be onsite (See Section 5.0 Definitions for clarification).

E. Training for RP Personnel A training program for RP personnel shall be developed by Nuclear Training. Nuclear Training shall issue procedures detailing the program. The Program Manager of Radiological Services, will concur with the initial issuance and any change to procedures for the training of RP personnel. The National Voluntary Laboratory Accreditation Program (NVLAP), Technical Director will concur with the training requirements and procedures involving NVLAP accredited activities.

3.4 Exposure Control Information regarding an individual's occupational radiation exposure is maintained pursuant to and in accordance with the Privacy Act of 1974,5 U.S.C. 552a and TVA's Privacy Act regulations (18 CFR 1301 Subpart 8). They are designated as TVA-23.

3.4.1 External Exposure Control*

A determination of prior dose (Le., current calendar year dose and previous years dose) using NRC FORM 4, or equivalent, will be made for each individual who is likely to receive an annual occupational dose at TVA that requires monitoring pursuant to 10CFR20.1502.

A signed affidavit disclosing the current year and previous years dose is acceptable for individuals requiring an administrative dose level (ADL) of less than or equal to 500 mrem per year (less than or equal to 100 mrem per year for minors and declared pregnant women) for occupational exposure to radiation from licensed and unlicensed radiation sources under the control of the licensee.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 11 of 46 3.4.1 External Exposure Control* (continued)

A. Minimum dosimetry requirements are*:

1. All individuals who are expected to work in a RCA shall process through RP (or RSO for non-nuclear facilities) when arriving, transferring, or terminating at the NPG. In addition, monitored and NPG staff individuals who will visit another licensee or TVA plant, and require a thermoluminescent dosimeter (TLD) on that visit, must check out prior to leaving their respective sites unless exempted by RP (or RSO). If an employee is assigned to work at a non-TVA installation where an exposure to radiation is incurred, the employee shall inform RP (or RSO) of this assignment. The employee shall turn in their dosimetry, obtain any required bioassays, and complete any requested documentation. When the employee returns, they must report to RP (or RSO) to obtain any required bioassay and update their exposure records.

2. Dosimetry may be issued by RP (or RSO) after appropriate previous occupational exposure documentation is obtained and all applicable requirements have been met (Le., bioassay, training, etc.).

3. Individuals are responsible for maintaining possession of their TLD and ensuring that it is worn in accordance with the requirements of this instruction.

4. Dosimetry processing equipment, TLDs, pocket chambers, and electronic dosimetry shall be calibrated in accordance with approved procedures.

5. As a minimum, all assigned TLDs are read at least semi-annually. TLDs may be processed as necessary.

6. Provide each worker entering an RCA with dosimetry capable of measuring the worker's dose. Accomplish this by using a dosimeter of record (for example, a TLD), appropriate for the radiological environment, provided by a National Voluntary Laboratory Accreditations Program (NVLAP) certified processor (utility or vendor).

7. NPG nuclear plant facilities shall provide each worker entering an RCA with dosimetry giving the worker the means of tracking individual dose. Accomplish this by using a self-reading dosimeter such as an electronic dosimeter or a pocket ion chamber.

8. All individuals (facility staff, temporary workers, etc.) who enter the RCA and are not expected to receive measurable occupational exposure to radiation from licensed and unlicensed radiation sources under the control of the licensee (Le.,

~1 0 mrem) during the monitoring period shall, as a minimum, be provided with a secondary monitoring device and be escorted by an individual provided with a NVLAP accredited dosimetry device appropriate for the radiological environment.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 12 of 46 3.4.1 External Exposure Control* (continued)

B. Restricted Area*

Any individual who enters a restricted area (in accordance with 10 CFR 20.1003) should be monitored for occupational exposure to radiation from licensed and unlicensed radiation sources under the control of the licensee, using a NVLAP accredited dosimeter(s), unless it is demonstrated through prospective determination that such monitoring is not necessary. When monitoring is not required workplace monitoring is appropriate. Minors and declared pregnant women likely to receive, in 1 year, a total effective dose equivalent in excess of 100 mrem shall be individually monitored.

C. Radiologically Controlled Area*

A system shall be implemented to track and control worker radiation exposures. The assigned deep-dose equivalent (DDE) and shallow-dose equivalent (SDE) must be for the part of the body receiving the highest exposure.

1. Dosimetry should normally be worn on the front of the person between the neck and beltline. It should be in a clearly visible position. When worn in combination, the secondary dosimetry should be located near the TLD. The TLD beta window side of the TLD should face outward. When in a radiation area or high radiation area, the pocket chamber and/or electronic dosimeter should be placed in a location that will allow the user to frequently read the pocket chamber and/or electronic dosimeter.

2. Individuals shall report to RP (or RSO) when a secondary electronic dosimeter being used for dose control alarms or pocket chamber reads in excess of 3/4 scale.

3. In the case of single dosimeter usage, placement of the dosimeter may be varied at RP's (or RSO's) discretion. Repositioning of the whole-body dosimeter should be considered when it is likely that a location on the whole-body will exceed the chest dose by more than 50 percent, and the whole body dose rate is >100 mrem/hr.

4. If the highest dose location on the whole-body is not known, the work area dose-rate gradients make it likely that a location on the whole-body will exceed the chest dose by more than 50 percent, a whole-body dose in excess of 300 mrem is expected during the task, and the whole body dose rates in the immediate work area are at least 100 mrem per hour, then multiple TLD badges should be worn on those whole-body locations that might receive the highest dose. Secondary dosimetry should be evaluated to be worn with each TLD ; with the primary secondary dosimeter (log in dosimeter) worn on the body part that is likely to receive the highest DDE).

5. The deep-dose equivalent, lens dose equivalent, and shallow-dose equivalent may be assessed from surveys or other radiation measurements for the purpose of demonstrating compliance with the occupational dose limits, if the individual

) monitoring device was not in the region of highest potential exposure, or the results of individual monitoring are unavailable.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 13 of 46 3.4.1 External Exposure Control* (continued)

6. Extremity monitors should be provided when an extremity could receive 500 mrem (SDE) or more and greater than two times the whole body dose, or if it is determined that extremity monitoring is required under 10 CFR 20.1502.

D. NVLAP Programs shall be established to obtain and maintain accreditation of dosimetry processing and evaluation under the NVLAP. The Authorized Representative shall be responsible for this program and shall serve as the official contact between TVA and the accrediting organization.

E. Dose Calculationsllnvestigations*

In some instances, due to the loss or damage of a monitoring device or the inability of the monitoring device to measure certain types of radiation, it will be necessary to calculate an individual's dose as appropriate. All calculations / investigations shall be documented and retained as an Individual-based Radiological Control Program record (SPP-5.9). Dose calculations/investigations are reviewed and approved by radiation protection supervision. When available, the individual signs the assessment verifying that the information provided is accurate and that the individual understands the dose that was assigned as a result of the assessment.

1. RP (or RSO) shall investigate TLD and secondary dosimeter reading discrepancies.

2. Individuals shall inform RP (or RSO) if they lose their TLD or their secondary dosimeter is lost, damaged, exhibits an unexpected response (such as alarms due to integrated dose or a dose rate) or is off-scale.

3. TLD results are normally used as the official record of radiation exposure.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 14 of 46 3.4.1 External Exposure Control* (continued)

F. Administrative Dose Levels*

1. Occupational radiation dose limits at NPG facilities are consistent with the limits given in 10 CFR 20.

2. Administrative dose levels (ADLs) to be used as guidelines for maintaining doses below regulatory limits have been established within the NPG and shall be observed for routine work. This program is not applicable to minors or declared pregnant women. Obtain appropriate station supervision and radiation protection management approval to increase a worker'S administrative dose level.

Examples of a bona fide need for a dose extension are that 1) the unique ability or experience of the individual will minimize collective dose; and 2) other qualified individuals with lower doses are not available. However, do not repeatedly give the same individuals higher dose for the sake of reducing overall collective dose.

The NPG Administrative Dose Level Program is summarized in Table 1 below:

TABLE 1 ADMINISTRATIVE DOSE LEVEL PROGRAM Dose Equivalent (Rem) Requirement Authorization to exceed i (signatures)

Up to 0.5 TEDE Statement of current year dose Not applicable (or 1.5 LDE or 5.0 SDE) at and previous years dose signed TVA by individual Up to 1.0 TEDE NRC FORM-4 or equivalent to Not applicable (or 3.0 LDE or 10 SDE) all document current year and sources previous years dose equivalent To exceed 1.0 TEDE Same as above RPM/RSO (or 3.0 LDE or 10 SDE) all sources To exceed Form-4 information must be RPM/RSO, Plant 5.0 TEDE3 all sources verified and a Planned Special Manager1, and Site vp2 Exposure initiated or SED as appropriate.

To exceed 1N" all Form-4 must be verified RPM/RSO, Plant sources Manager1 , and Site VP2 or SEDas_apj)r~riate_._

At non-nuclear plant sites, this will be the RSO's immediate supervisor.

2 At non-nuclear plant sites, this will be the applicable TVA VP.

3 Authorizations for a planned special exposure will only be considered in an exceptional situation when alternatives that might avoid the dose estimated to result from the planned special exposure are unavailable or impractical.

4 Total effective dose equivalent should not exceed 1N rem, where N equals the individual's age in years at last birthday, without the authorization signatures delineated.

3. ADLs are based on dosimeter(s) used in determining the reported dose. Results which exceed an administrative level, based on other dosimeter data, do not violate the ADL.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 15 of 46 3.4.1 External Exposure Control* (continued)

NOTE Exceeding an ADL as a result of an incorrect signed documentation of current year or previous years dose does not constitute a violation of the intent of the ADL.

4. When visitor or contract personnel have more restrictive dose limits than TVA - the more restrictive limits will be used. It is the responsibility of the contractor to provide written notification to RP (or RSO) of any company administrative limit.

5. To ensure that ADLs are not exceeded, an administrative control system shall be maintained.

6. Individuals under the age of 18 shall not enter radiologically controlled areas.

7. Individuals whose lifetime accumulated total effective dose equivalent (TEDE) is equal to or exceeds 1N rem will normally be limited to 1,000 mrem per year at TVA. The first four administrative controls of Table 1 are applicable.

8. The RPM shall prepare a report for the TVA Chief Nuclear Officer and Executive Vice President for submittal within 30 days to INPO's Radiological Protection and Emergency Preparedness Division and the NRC (10 CFR 20.2105) if a regulatory limit is exceeded or a Planned Special Exposure (PSE) is used (10 CFR 20.2203,20.2204, and 20.2205).

9. Any worker who exceeds a regulatory dose limit shall not be permitted to enter any RCA until all investigations surrounding the event are completed. The RPM/RSO or designee must approve reentry.

10. Any personnel exposure received which is in excess of the limits of 10 CFR 20.1201 shall be reported by the RPM/RSO to Radiation Effects Advisory Group (REAG) and the appropriate area chief physician for an examination.

A medical examination and authorization from the Chief Nuclear Officer and Executive Vice President are required before resumption of duties in RCAs for individuals who have received five times the annual limit of 10 CFR 20.1201.

11 . ADL Program Records All ADL extension evaluations (approved and disapproved) shall be documented and retained as an Individual-based Radiological Control Program record (SPP-5.9).

)

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006

_~r~cesses Page 16 of 46 3.4.1 External Exposure Control* (continued)

12. Prenatal Exposure Program Information regarding a woman's participation in the Prenatal Exposure Program is maintained pursuant to and in accordance with the Privacy Act of 1974, 5 U.S.C. 552a and TVA's Privacy Act regulations (18 CFR 1301 Subpart B).

They are designated as TVA-23.

NPG shall ensure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to the occupational exposure does not exceed 500 mrem (10 CFR 20.1208(a)).

The dose equivalent to the embryo/fetus is the sum of the deep dose equivalent to the declared pregnant woman, the dose equivalent to the embryo/fetus resulting from radionuclides in the embryo/fetus and radionuclides in the declared pregnant woman.

If the dose equivalent to the embryo/fetus is found to have exceeded 500 mrem, or is within 50 mrem of this dose, by the time the woman declares the pregnancy to NPG, NPG shall be deemed to be in compliance with 10 CFR 20.1208(a) if the additional dose equivalent to the embryo/fetus does not exceed 50 mrem during the remainder of the pregnancy.

a. Training (1) It is the recommendation of Radiation Protection and the RSO's that prenatal radiation exposure will be controlled to ensure that the embryo/fetus is not subjected to any undue risk.

Accordingly, individuals who may be exposed to radiation during their employment with TVA will be given information on the potential hazards to the embryo/fetus from radiation exposure based on the best current scientific knowledge and on the current exposure limits recommended for pregnant women and women who intend to become pregnant. This information will be provided in the standard RP training provided to all employees and will be provided in all required updates of such training.

Records will be maintained on the attendance of employees at this training. These records shall be maintained in such a fashion as to allow timely retrieval of an individual's attendance record.

(2) In addition, counseling on the potential radiation hazard to an embryo/fetus will be provided by RP (or RSO) to any woman who requests it apart from the standard training sessions. A written record of this counseling shall be made and maintained as an "Individual Radiation Exposure History Record" in accordance with the provisions of the Radiological Control records management program.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev.OOOG Processes Page 17 of 4G 3.4.1 External Exposure Control* (continued)

b. Voluntary Prenatal Exposure Program This program is strictly voluntary. It is available to women who are pregnant or are planning to become pregnant, at their sole discretion. Request to participate in the program shall be in writing. In addition, women who elect to participate in this program may choose to leave the program at any time by submitting a written statement to the effect. Requests to participate or withdraw will be maintained as an "Individual Radiation Exposure History Record" in accordance with the provisions of the Radiological Control records management program.

Participants in this program will be provided with counseling by RP (or RSO).

The counseling will be the same as that discussed in the section on Training (3.4.1 F .12.a), with records of participation made and maintained in the same manner.

The following actions shall be implemented to ensure prenatal radiation exposure is kept to a minimum for those women who have declared their pregnancy or their intent to become pregnant and have requested to participate in this program:

(1) The deep dose equivalent to the embryo/fetus because of occupational exposure of a woman who has declared that she is pregnant shall:

Be maintained ALARA by the pregnant individual, by the individual's supervisor, and by the facility's RPM/RSO.

Be limited to a value that would not let the woman exceed 50 mrem in a single month.

The total effective dose equivalent received by the declared pregnant woman shall be controlled to ensure compliance with dose equivalent limits for the embryo/fetus.

Further, a declared pregnant worker shall be excluded from Planned Special Exposure activities

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(2) For the case of a woman who has voluntarily declared her intent to become pregnant:

Her occupational radiation exposure should be maintained ALARA by the individual, by the individual's supervisor, and by the facilities RPM/RSO or designee.

The RPM/RSO or designee should limit the woman to a value that would not let her exceed 50 mrem total effective dose equivalent in a single month.

The RPM/RSO or designee will inform the woman that she will confirm her intent to become pregnant in writing every two months until she either declares her pregnancy, states she no longer intends to become pregnant, or chooses to leave the program.

Each of these actions shall be confirmed in writing and the statements shall be maintained as an "Individual Radiation Exposure History Record" in accordance with the provisions of SPP-5.9.

(3) Women participating in the prenatal radiation exposure program shall be monitored by a dosimeter, processed and evaluated by a NVLAP accredited dosimetry processor, if they enter or work in an area where they could exceed a deep dose equivalent of 100 mrem DOE in a year at TVA.

(4) Because of the uncertainties in assigning dose to the embryo/fetus due to the uptake of radionuclides, women participating in the prenatal radiation exposure program shall not enter posted contamination or airborne activity areas.

(5) Reasonable efforts will be made by management to retain participants in the program in their current job status, subject to the needs of the facility and the provisions of the applicable negotiated agreement.

Retention of current job status cannot be guaranteed.

(6) Any exception to the above exposure recommendations for a woman who has declared pregnancy or her intent to become pregnant and has requested to participate in this program shall be forwarded to the Chair, Radiation Effects Advisory Group (REAG).

13. Employees shall report to their local TVA medical facility and the site RP organization (or RSO) whenever they receive medical external radiation therapy or internal radionuclides for diagnosis or treatment. Routine diagnostic X-rays need not be reported. RP (or RSO) may suspend access, if necessary, to the restricted area or radiologically controlled area if radiological control of these areas would be compromised by entry of these persons. Additionally, access may be suspended if consultations regarding additional exposure are pending.

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14. The ADLs for individuals receiving therapeutic medical radiation exposures and individuals with radiologically-related medical restrictions should be evaluated on a case-by-case basis. It is recommended that the opinion and recommendations of the individual's treating specialist be solicited. The treating specialist would be most aware of the individual diagnosis, specific therapy, the attendant risks, as well as any unusual susceptibility or precautions necessary regarding workplace radiation exposure. The individual and his or her supervisor will be counseled by Medical Services. A written record of this counseling shall be made and maintained along with all other supporting documentation. For individuals receiving therapeutic medical radiation exposures the individual should have risks clearly explained and be encouraged, but not required, to be placed on a lower ADL.

If the individual chooses to be placed on a lower ADL, the individual shall be informed that reasonable accommodations will be made to retain him/her in his/her present job status.

However, his/her present job status cannot be guaranteed. An annual ADL of 500 mrem for the whole body appears to be a reasonable limit, absent other circumstances which warrant a higher or lower ADL. For individuals with radiologically-related medical restrictions, Medical Services, in consultation with the RPM/RSO or designee, will determine if occupational exposure should be administratively restricted.

G. Skin Dose from Contamination*

1. An assessment of a worker's dose should be performed whenever a worker has been contaminated with >54,000 cpm-hrs as measured with a standard 15.5 cm 2 frisker (See RCTP-106 for skin dose calculation).

2. Dose to the skin will be averaged over an area of 10 cm 2 (contiguous) receiving the highest dose.

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Page 20 of 46 3.4.2 Emergency Exposure Guidance and Planned Special Exposures (PSE)*

A. It is consistent with the risk concept to accept exposures leading to doses in excess of those appropriate for routine operation when recovery from an accident or major operational difficulty is necessary. Saving of a life, measures to circumvent substantial exposure to the general public, or the preservation of valuable installations may be sufficient cause for accepting above normal exposures. Dose limits for an emergency cannot be specified, but they should be commensurate with the significance of the objective and held to the lowest practicable level that the emergency permits.

Guidance for emergency exposure is specified in the TVA Radiological Emergency Plan.

Any decision to embark on emergency operations which would result in exposures in excess of 10 CFR 20.1201 should be done in consultation with the most senior member of RP who is available on a timely basis. Personnel must be made aware of possible consequences of such an exposure and selected on a voluntary basis.

Emergency team members who are expected to respond to a radiological emergency must be made aware of the consequences of such exposure.

B. TVA may authorize an individual to receive doses in addition to and accounted for separately from the yearly dose limits prescribed in 10 CFR 20.1201 provided that certain conditions are satisfied. These conditions are delineated in 10 CFR 20.1206.

Authorizations for a planned special exposure will only be considered in an exceptional situation when alternatives that might avoid the dose estimated to result from the planned special exposure are unavailable or impractical. Individuals required to authorize planned special exposures are indicated on Table 1.

Personnel must be made aware of possible consequences of such an exposure. PSEs are to be conducted in accordance with 10 CFR 20.1206, 20.2105, 20.2106, and with guidance from Regulatory Guide 8.35.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 21 of 46 3.4.3 Internal Exposure Control*

A. Confirmatory Internal Exposure Monitoring*

Internal occupational dose is controlled though facility design, engineering controls, confinement and reduction of contaminated areas, limiting access to radiological controlled areas, and the use of respiratory protective equipment. Personnel are not routinely monitored for internal deposited radioactive material. Confirmatory monitoring (by licensee) is performed for individuals through the assessment and tracking of DAC-h. Radio-bioassay (in vitro and in vivo measurement and analysis) is employed to confirm and/or evaluate probable intake.

B. Respiratory Protection Program*

A respiratory protection program shall be established and maintained in accordance with 10 CFR 20. Workers shall have respiratory protection training before wearing respiratory protection equipment.

1. TVA is responsible for providing a workplace environment in which individuals are adequately protected from hazards, including hazards from exposure to ionizing radiation. As part of TVA's program to maintain exposures ALARA, the TEDE is to be ALARA for activities subject to the 10 CFR 20 "Standards for Protection Against Radiation." These requirements allow intakes of radioactive material by workers if such intakes result in lower external dose and maintains TEDE ALARA.

Under these requirements intakes of radioactive material are permissible if evaluations predict that use of respiratory protection equipment will result in a higher TEDE. Additionally, other factors may be considered in the evaluation for maintaining TEDE ALARA. These factors may include, but are not limited to, environmental conditions, safety conditions, accessibility conditions, worker comfort, wear times, and the type of respiratory equipment specified or available.

All TEDE ALARA evaluations shall be documented and retained as a Facility-based Radiological Control Program record (SPP-S.9). Dose calculations/investigations are reviewed and approved by radiation protection supervision.

Unplanned intakes (no documented TEDE ALARA evaluation) of radioactive material by workers that result in a internal dose of 10 mrem or greater shall be documented in the Corrective Action Program (SPP-3.1).

2. The primary means to minimize the intake of airborne radioactive materials is to control the generation of airborne radioactivity. This is best accomplished at its source and by process or other engineering controls. These controls include identification and repair of leaks, process modification, decontamination, containment, and ventilation control. Routine and special tasks should be planned such that potential sources of airborne radioactive material are managed by repair, decontamination, process, or other engineering controls.

3. If it is impractical to repair, decontaminate, apply process or other engineering controls or while these processes are being implemented, other measures should be taken to limit the uptake of radioactive materials. These measures include increased surveillance, limitation of working times, use of respiratory protective devices, or combination thereof.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 22 of 46 3.4.3 Internal Exposure Control* (continued)

4. Regardless of how effective respiratory protective equipment may be in minimizing the inhalation of airborne radioactive materials, the use of respiratory equipment is less desirable (than proper containment, process, or engineering controls) due to physical stress and increased constraints on vision, freedom of movement, and ability to communicate.

Because the work environment may pose more of a health risk to individuals than an uptake of radioactive material, a risk evaluation should be performed when use of respiratory protection equipment is being considered.

5. Reasonable pre-job estimates of potential intakes and internal dose rates, the desire to maximize worker comfort and safety, and adequate protection of workers without unnecessary restrictions should be the basis by which the decision to utilize respiratory protection equipment is made. As there is difficulty in quantifying efficiency factors, general application of efficiency factors may be used based upon reasonable evaluation and documentation (of specific job types and expected conditions) for use of respiratory protection equipment application.

Evaluations should focus on the basic considerations of potential risk, and worker comfort and efficiency. Optimization of these considerations in performing the evaluation is consistent with ALARA objectives and should provide assurance that the TEDE is maintained ALARA.

If respirators are selected as the means to minimize internal exposure, the following applies:

a. Respirators are only to be issued to personnel who are appropriately trained, fitted, and medically qualified. Persons may wear contact lenses or special glasses.

b. Positive control of respirators is to be maintained for issue and use.

c. Respirator users may leave areas where respirators are in use for relief in case of equipment malfunctions, undue physical distress, procedural or communication failure, significant deterioration of operational conditions, or any other condition that might require such relief.

d. Periods of respirator use are to be appropriately managed. That is, the task length, protective clothing required, and environmental conditions (Le.,

temperature, humidity) should be considered.

e. Assignment and use of respiratory protective devices for routine and special work evaluations are to be controlled by Radiation Work Permits.

f. Assignment and use of respiratory protective devices for emergency situations are to be in accordance with TVA emergency procedures.

6. Personnel reasonably expected to be required to wear respiratory protection during emergency situations shall maintain required training, fit, and medical qualifications.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 23 of 46 3.4.3 Internal Exposure Control* (continued)

C. Program Elements*

Program elements, at a minimum, are to include:

1. Air sampling sufficient to identify the potential hazard, permit proper equipment selection, and estimate exposures;

2. Surveys and bioassays, as appropriate, to evaluate actual intakes;

3. Testing of respirators for operability immediately prior to each use;

4. Written procedures shall be established that address: selection, fitting, issuance, maintenance, and testing of respirators, including testing for operability immediately prior to each use; program audits; minimum qualifications of program supervisors and implementing personnel; limitations on periods of respirator use and relief from respirator use; maintaining TEDE ALARA and performing evaluations; supervision and training of personnel; monitoring (including air sampling and bioassays), and recordkeeping; a description of the applications of respirators for routine, non-routine, and emergency respirator use; and periodic medical evaluation (NRC Regulatory Guide 8.15).

5. Determination by a physician prior to the initial fitting of respirators, and annually (quarter ending) thereafter or periodically at a frequency determined by a physician, that the individual user is medically fit to use the respiratory protection equipment.

D. Respiratory Equipment Selection*

Respiratory equipment selection shall be in accordance with 10 CFR 20 with the exception that half-face respirators shall not be used for protection against airborne radioactive materials. Individuals using respiratory protection devices must, in addition to having a valid mask fit, be clean shaven in the areas of the face which provide a sealing surface for the mask before using the mask.

E. Respiratory Equipment Contamination Limits*

Respirators shall not be used if smearable contamination is detected in excess of 20 dpm/100 cm 2 alpha or 1,000 dpm/1 00 cm 2 beta-gamma, or if fixed contamination levels on respirator surfaces exceed 30 cpm alpha or 500 cpm beta-gamma. Reasonable efforts should be made to reduce contamination levels on inside surfaces.

NOTE It is not necessary to survey for alpha emitters if the beta-gamma to alpha activity ratio is greater than 50 to 1 as determined through plant surveys.

F. ALls and DACs

)

ALls and DACs for occupational exposure are given in 10 CFR 20.1204 and Table 1 of Appendix B to Part 20.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 24 of 46 3.4.3 Internal Exposure Control* (continued)

G. Internal Dose Monitoring*

NOTE Noble gases are not included in this monitoring tracking.

If CEDE >one rem, CDE and TODE must also be calculated.

1. Internal dose monitoring (DAC-hr tracking including bioassay) is required for:

a. Adult workers that are likely to receive an occupational intake in excess of 0.1 ALI or 200 DAC-h in a year.

b. Declared pregnant women likely to receive a CEDE of 100 mrem or 40 DAC-h in a year.

TVA's historical operational experience has shown that internal dose monitoring, to meet the criteria above, is not warranted unless there is a significant potential for internal exposure. Therefore, based on this prospective determination, DAC-hr tracking will not be performed for work in atmospheres less than 1 DAC, unless a case-specific assessment of airborne activities and exposure times indicates that these criteria would be met.

2. A bioassay shall be performed when initial contamination on the face (excluding noble gases) indicates that an uptake may have occurred.

3. A bioassay shall be required whenever nasal contamination (excluding noble gases) is detected by nasal smears or other detection methods.

4. A bioassay shall be required for ingestion or suspected ingestion of radioactive material.

5. A bioassay shall be required whenever contamination of an open wound is detected.

6. A bioassay shall be required for evidence of damage to or failure of a respiratory protective device during use.

7. An initial bioassay shall be required before initial entry into contamination or airborne radioactivity areas.

8. A Whole Body Count (WBC) shall be performed whenever an individual exceeds 4 DAC-h (total) in a calendar year unless it consists solely of non-gamma emitting nuclides. In addition, a species-specific bioassay (e.g., urinalysis, fecal) will be performed when the DAC-h accumulation for a species (e.g., 3H, alpha) is 4. =

The species-specific bioassay may be waived for certain protracted exposures, based on retention/ excretion characteristics. Waivers shall be documented, approved by the RPM/RSO, and retained as a Individual-based Radiological Control Program record (SPP-5.9).

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 25 of 46 3.4.3 Internal Exposure Control* (continued)

9. Whole body counts indicating the presence of licensed radioactivity above background (Critical level of Detection (Lc) at the 95% confidence level) shall be counted as positive.

10. A termination bioassay (Le., exiting from TVA) shall be required of individuals who have entered bioassay areas.

11. Random bioassays shall be conducted periodically to evaluate the effectiveness of internal exposure control measures. Frequency will be based on plant conditions and maintenance activities.

12. A bioassay shall be given with greater frequency based on the judgment of RP.

H. Required Actions Based Upon Bioassay Results

1. Screening Level (SL) - the level below which a follow-up bioassay measurement need not be considered for investigation of intake and assignment of dose. The SL is based on a committed effective dose equivalent of 10 mrem. Calculated committed effective dose equivalent equal to or greater than the SL are recorded and included in an individual's Total Effective Dose Equivalent. Bioassay measurements indicating above the SL should have sufficient follow-up bioassays to accurately define the amount of radioactive material intake and the appropriate

) biokinetic pathway.

2. Evaluation Level (EL) - the level above which a bioassay or air monitoring result shall be investigated, to the extent reasonable, to determine actual conditions and parameters for dose evaluation. An investigation may involve special measurements, work history review, determination of material form, and modification of biokinetic parameters. Investigate incident, as appropriate, to determine probable cause and the extent of condition. As appropriate, initiate any possible corrective actions or additional measures to prevent recurrence. The EL is 50-mrem committed effective dose equivalent.

3. Internal Dose Assessments - Because of differences in physical properties and metabolic processes, each individual's dose resulting from an internal exposure is unique. In other words, the same radionuclide intake to multiple individuals will likely cause different doses to each individual. However, for very small intakes the use of reference man physiological data and biokinetic modeling provide a generally conservative screening tool to estimate CEDE, demonstrate compliance with regulatory requirements, and to provide assurance of an appropriate level of protection to workers with respect to internal radiation exposure. Base internal dose assessments on radiological and biological parameters, regulatory requirements and guidance, recognized national standards, and dosimetric models of ICRP and NCRP. Qualified radiation protection individuals shall independently review dose assessment calculations to ensure accuracy of results.

Workers acknowledge the dose assessment, verifying that information they provided is accurate. Explain to the worker the significance of the dose received.

)

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 26 of 46 3.4.3 Internal Exposure Control* (continued)

4. RP may take exception to certain requirements for bioassays involving individuals undergoing medical treatments with radioactive materials or initial screening before job assignment. Actions to be taken will be determined on a case by case basis and may involve work restrictions or delays as appropriate.

5. Bioassays and Internal Dose Assessments shall be documented and retained as a Individual-based Radiological Control Program record (SPP-5.9).

3.4.4 Radiation Exposure Tracking, Recording, and Reporting A tracking system shall be implemented which will track radiation exposure for purposes of trend analysis and work planning, and provide data for management evaluations of the ALARA program.

A. Exposure Control System*

An exposure control system will be implemented which will:

1. Keep up-to-date exposure data from dosimeters, calculated doses, and DAC-hr.

2. Compare individual dose data with TVA ADLs and regulatory limits.

3. Keep the supervisor informed of workers' exposure.

4. Keep employees informed of their own exposure.

B. Dose Record System*

A dose record system shall be implemented by RP for purposes of maintaining historical dose records for all persons for whom personnel monitoring or dose calculations are performed. These records are collected and maintained pursuant to and in accordance with the Privacy Act of 1974,5 U.S.C. 552a and TVA's Privacy Act regulations (18 CFR 1301 Subpart B). They are designated as TVA-23. The records maintained shall include: the deep-dose equivalent to the whole body, lens dose equivalent, shallow-dose equivalent to the skin, and shallow-dose equivalent to the extremities; the estimated intake of radionuclides; the committed effective dose equivalent assigned to the intake of radionuclides; and the specific information used to assess the committed effective dose equivalent pursuant to 10 CFR 20.1204(a) and (c), and when required by 10 CFR 20.2106.

Deep Dose Equivalent, Lens Dose Equivalent, Shallow Dose Equivalent (Whole-body),

Shallow Dose Equivalent (Maximum extremity), Committed Effective Dose Equivalent, Committed Dose Equivalent, Total Effective Dose Equivalent, and Total Organ Dose Equivalent dose information shall be calculated, maintained, and reported to the NRC and individuals according to NRC Regulatory Guides 8.7 and 8.34 and NRC Technical Communication RADIATION RECORDS DATA COLLECTION AND ANALYSIS to TVA dated January 4, 1994. The dose record system shall make a clear distinction among the quantities entered on the records (e.g., total effective dose equivalent, shallow-dose equivalent, lens dose equivalent, deep-dose equivalent, committed effective dose equivalent).

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006

') Processes - ---- -- ----- ---- --_. - - _ . ---_ .. - -----

Page 27 of 46 3.4.4 Radiation Exposure Tracking, Recording, and Reporting (continued)

1. The system includes:

a. All official dose records for each individual, including externally measured or calculated doses, whole body counting results and internal dose commitment calculation, personnel contamination reports, and investigation reports as appropriate.

b. Means to store and retrieve records in accordance with NPG's quality assurance program requirements.

c. Means to retrieve individual dose records by name or employee identification number.

d. Means for RP personnel to obtain individual records.

e. Means to generate all required reports.

C. Dose Record Reporting*

1. Those individuals who receive occupational exposure and require monitoring per '

10 CFR 20.1502 shall have their doses reported annually to the NRC and the individuals on an NRC FORM-5 or an electronic record containing all the information required by a FORM-5.

2. These reports are generated and reported by licensee as required by 10 CFR 20.2206.

3. In addition, a copy of any exposure report, required by 10 CFR 20 to be submitted to the NRC shall also be sent to the occupationally exposed individual or member of the public.

4. Individuals that participate in the confirmatory monitoring program may request the results of their confirmatory monitoring.

5. External exposures as measured with a NVLAP accredited device will be recorded and reported at a 10 mrem threshold value.

6. When determining the dose from airborne radioactive material, NPG shall include the contribution to the deep-dose equivalent, lens dose equivalent, and shallow-dose equivalent from external exposure to the radioactive cloud. External exposures as calculated for noble gas submersion dose will be integrated in the Radiation Protection Records system. Doses calculated by the RP Computer system will be reported at a 1 mrem monitoring period threshold value.

7. Internal exposures as calculated for derived air concentration (DAC-hrs) exposures and/or bioassay data will be integrated in the Radiation Protection Records system. Doses calculated by the Radiation Protection Computer system are reported at a 1 mrem threshold.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 28 of 46 3.S Radiologically Controlled Areas (RCAs)*

3.S.1 Posting*

In addition to the posting requirements of 10 CFR 20, the following shall apply:

A. Each RCA shall be posted by yellow and magenta signs bearing the standard radiation warning symbol and the words "Caution - Radiologically Controlled Area." The posting shall also state that a monitoring device is required (unless it has been determined that monitoring is not required).

B. Contamination areas shall have conspicuous boundaries consisting of such items as rad-ribbon, rad-rope, rad-tape, and step-off pads and be posted by yellow and magenta signs bearing the standard radiation warning symbol and the words "Caution-Contaminated Area" or "Caution-Contamination Area." Where, due to physical space limitations, it is impractical to post a contaminated area as described above, the area may be noted with radiation tape and/or radiation hazard tags.

Physical space limitation is intended to apply to such areas as floor drains, electrical panels, sample sinks, etc.

Radiological postings shall be displayed with yellow and magenta colors in accordance with 10 CFR 20.1901 and TVA's Final Position on Implementation (L61 931029800) of 10 CFR 20 implementation.

) 3.S.2 Surveys*

A. Surveys of the RCA equipment, materials, and personnel shall be of sufficient type and frequency to properly define the magnitude and extent of radiation levels; and the potential radiological hazards.

B. Clean areas adjacent to a RCA shall be routinely surveyed to ensure that radioactive material is not present above the limits of SubSection 3.S.3A. Should radioactive material be found above the limits, immediate steps shall be taken to contain the material and have it removed and posted properly.

C. Surveys are documented and include information specified by ANI/MAELU Information Bulletin 80-1A, Revision 4. Surveys are retained as a Facility-based Radiological Control Program record in accordance with SPP-S.9.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 29 of 46 3.5.3 Limits*

A. Cle?ln areas or material including equipment and tools released for unrestricted use shall have radioactivity below the following values:

Total Radioactivity Removable Radioactivity (Refer to Note 3) I Alpha Beta/Gamma Alpha Beta/Gamma I Refer to Note 2 Refer to Note 1 Refer to Note 2 30 cpm 100 cpm 20 dpm/1 00cm2 1000 dpm/100 cm 2 NOTES

1) Equal to 100 cpm above background, as measured by a frisker type instrument (Ludlum Model 177 with a HP-210 probe or equivalent with nominal 10 percent efficiency), at about 1/2 inch from the surface of the material being monitored in a background radiation field not exceeding 300 cpm and moving the detector probe at a slow rate. Material may be surveyed using other monitoring devices and techniques which are as sensitive as the frisker.

2) As measured by a scintillation type instrument (e.g., Bicron Surveyor M with an A50 probe or equivalent and a nominal 20 percent efficiency), at the surface of the material being monitored and holding the probe stationary. Material may be surveyed using other monitoring devices and techniques which are as sensitive as the Bicron.

It is not necessary to survey for alpha emitters if the beta-gamma to alpha activity ratio is greater than 50 to 1 as determined through facility surveys.

3) As evidenced by removal of radioactive materials from the surface by paper/cloth smears or wipes using appropriate techniques.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 30 of 46 3.5.3 Limits* (continued)

B. Release of material from the RCA NOTES

1) Material released for unrestricted use that has been in the RCA but not in contamination areas shall be monitored.

2) The whole body frisk monitoring conducted by individuals while exiting the RCA is considered sufficient for personal items (pens, keys, flashlights, pagers, blackberries, cell phones, etc.)

which have not been in a contaminated area.

3) Material released from RCAs established for radiation hazards only (no contaminated or airborne areas) do not require contamination surveys for release.

The following table indicates the items that individuals may wear through a personnel contamination monitor or may place in an article monitor for release from the RCA.

ALL OTHER ITEMS MUST BE SURVEYED BY RP PRIOR TO REMOVAL FROM THE RCA.

PCM

[Only if being worn]

Personal Clothing, Shoes, Watches, Jewelry, Eye Glasses, Hearing Aids, Ear Plugs, Hard Hat, Safety Glasses, Cell Phone, Pager, flashlight, and other similar personal devices.

Lanyards - Typical items worn on lanyard may remain attached.

Security Officer Weapons and other Security Officer Specific Equipment.

Small Personal Items in pockets (coins, wallet, keys, pens, pencils, etc.).

SAM Gloves, paperwork and notebooks (less than 1 inch thick), radios, flashlights, and items listed above not being worn.

C. As directed by RP, personnel exiting the RCA will be expected to remove all materials from their pockets to verify no items, other than those listed in the table above, are inadvertently brought through the Personal Contamination Monitors.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 31 of 46 3.5.3 Limits* (continued)

D. Detectable personnel contamination of either skin or clothing shall be documented for levels :::::100 cpm with a pancake probe equipped frisker or equivalent (considering hard-to-detect radionuclides). Protective clothing shall not be used if fixed contamination levels on the clothing exceed 4,500 cpm on a frisker or equivalent detection system.

E. The approval of site Plant Manager or designee shall be required before the release of any individuals with skin contamination in excess of these limits. If it may be reasonably determined that naturally occurring radionuclide and daughter products caused the contamination, the person may be released after documenting the contamination and cause. This documentation shall be retained as an Individual-based Radiological Control Program record in accordance with SPP-5.9.

F. The following levels shall be used for the tracking and trending of Personnel Contaminations (PCs)

1. Level 1 (100 cpm to 5,000 cpm) personnel contaminations shall be documented in the site's HIS-20 database and a Problem Evaluation Report initiated in the Corrective Action Program.

2. Level 2 (> 5000 cpm to 50,000 cpm) personnel contaminations may require follow-up evaluation depending on contamination exposure and radioactivity level and shall be documented in the site's HIS-20 database and a Problem Evaluation Report initiated in the Corrective Action Program.

3. Level 3 (> 50,000 cpm) personnel contaminations are significant, requiring expedited management attention and shall be investigated to determine if the contamination exposure period and radioactivity level will result in an estimated dose that will require a formal dose calculation (e.g., VARSKIN). They shall be documented in the site's HIS-20 database and a Problem Evaluation Report initiated in the Corrective Program.

G. Skin and non-skin contamination should be counted as a personnel contamination if the contamination is detected on a worker's skin or any item or clothing worn by a worker that is not protective clothing and is found to be ::::: 100 cpm. "Anticipated" or "TEDE ALARA" PCs should also be counted and documented. Contamination found to be attributed solely to radon or thoron and their progeny or noble gas should be exempted from being counted as a PC.

H. Any instruments used to release material from the RCA should have a functional check performed each week using a source with activity near the desired setpoint for alarm and reasonably approximating the station isotopic mix.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 32 of 46 3.5.3 Limits* (continued)

I. Automatic whole body Beta contamination monitors at the RCA exits should detect levels at the average beta energy of the station radionuclide mix equivalent ~5,000 dpm at a distance from the detector equivalent to the location of the individual being monitored based on the configuration of the whole body contamination monitor, the detector location, and the body part being monitored. Establish alarm setpoints as low as practical, considering the presence of difficult-to-detect isotopes in the station radionuclide mix. At least weekly, perform a functional check using a source with activity near the desired set pOint for alarm and reasonably approximating the station isotopic mix.

J. Gamma-sensitive portal monitors installed at the RCA exit should be used in a pause mode to optimize their detection capability. Portal monitors located at the RCA exit should have alarm setpoints that correspond to 30-35 nanocuries of Co-60 or about 75 nanocuries of Cs-137. At least each week when a portal monitor is in use, perform a functional check using a radioactive source with an activity appropriate for the alarm set point.

K. Install gamma-sensitive portal monitors at the protected area exit as a final barrier, increasing the likelihood of detecting contamination that may have been inadvertently released from the RCA. Establish the alarm set point as low as practical, considering ambient background radiation, the negative consequences of false-positive alarms, and reasonable egress times, but not greater than the alarm set points at the RCA exit.

3.5.4 General Requirements*

A. Personnel shall wear all required dosimetry and radiological protective equipment while in a RCA.

B. Personnel shall not eat, drink, smoke, chew gum or tobacco, or take anything orally in a RCA. If necessary, techniques for providing liquids or other consumables to personnel in a RCA can be used if precautions are taken to ensure that personnel do not ingest radioactivity. The transport of any consumable through or into any RCA will require appropriate Radiation Protection precautions to prevent contamination of the items.

C. Individuals (evaluated on a case-by-case basis by TVA Medical) are permitted to transport and use prescribed medications in the RCA, except for airborne or contaminated areas.

D. Personnel shall monitor themselves for contamination upon leaving any contaminated area and whenever required by Radiation Protection. Any indication of contamination shall be reported to RP.

E. Entry to and exit from the RCA shall be via access points authorized by RP.

F. All work in the RCA will require prior notification of RP.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 33 of 46 3.5.4 General Requirements* (continued)

G. Personnel shall monitor their electronic dosimeter dose carefully. Ensure that the electronic dosimeter is readily visible and exit the RCA BEFORE A DOSE ALARM IS RECEIVED. Valid electronic dosimeter dose alarms and unanticipated dose rate alarms received in the RCA shall be documented in the Corrective Action Program.

Investigate and document unplanned dose rate alarms and all accumulated dose alarms to determine if the worker was monitoring his or her exposure, radiological conditions changed, the scope of planned work was altered, or the dosimeter malfunctioned.

When in radiation areas workers shall monitor their electronic dosimeters frequently, once or twice per hour, and more frequently in high radiation areas to ensure dose received is consistent with expectations.

Dose rate alarm setpoints shall be established low enough to provide the worker with a warning of higher-than-expected work area dose rates for that specific job. They should also be set high enough to enable the worker to perform the specific job without receiving an unplanned dose rate alarm.

H. Open cuts, wounds, skin rashes, and infections shall be adequately protected and RP contacted for evaluation and authorization before entering the RCA.

3.5.5 Area-Specific Requirements*

A. Contaminated Area

1. Protective clothing as required shall be worn in contaminated areas.

2. Open cuts, wounds, skin rashes, and infections shall be adequately protected before entering a contamination area.

3. Three approved methods of removing equipment from a contamination zone are:

a. Have RP perform a survey of the material as it is removed. Every reasonable attempt should be made to utilize this method when removing equipment or material from a contamination zone (C-zone).

b. Properly bag and tag the material, minimize personnel exposure, transport it to the designated location.

c. Properly bag the material as it crosses the C-zone boundary and deliver to Radiation Protection.

B. High and Very High Radiation Area High and very high radiation areas shall be administratively controlled per license requirements.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 !

Processes Page 34 of 46 3.6 Radiation Work Permit System 3.6.1 Control of Work in the RCA A RWP system shall be established to document radiological conditions and prescribe appropriate protective requirements for work in radiologically controlled areas.

A. Site Radiation Protection shall be responsible for establishing entry requirements for radiological areas via the RWP.

B. The area in which the work is to be performed is surveyed for radiological hazards before the start of work and/or as appropriate during work to ensure that radiological hazards are properly identified.

C. Protective clothing and equipment, dosimetry, and work limitation requirements are specified for all workers entering the area.

3.6.2 RWP Criteria RWPs will normally be required for all work in radiologically controlled areas. RWPs shall always be required for areas where radiological conditions meet or exceed the criteria listed below.

A. Entering a "Radiation Area."

B. Entering a "High Radiation Area" or "Very High Radiation Area."

C. Entering a "Contaminated Area."

D. Entering an "Airborne Radioactivity Area."

E. Breaching a contaminated system or component.

F. RP discretion to provide adequate radiological control.

G. For radiographic examinations conducted at licensed nuclear facilities.

H. Entering an area or component where radiological conditions are unknown.

3.6.3 Waived Requirements for RWPs Waivers for prior issuance of an RWP during a declared emergency shall be controlled in accordance with RCDP-3, "Administration of Radiation Work Permits."

3.6.4 RWP Compliance Each worker shall be responsible for awareness and compliance with the radiation protection requirements of an RWP and for meeting the prerequisites for RWP entry.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 35 of 46 3.7 Radioactive Material Control*

3.7.1 Packaging, Handling, and Transfer A. Workers shall not remove any radioactive materials from an RCA without Radiation Protection authorization.

B. When radioactive material is in transit from one RCA to another, steps shall be taken to ensure that no material is lost in transit, left unattended, or placed in an unauthorized area.

C. Movement and storage of radioactive material within an RCA shall be done in such a way as to prevent the spread of contamination and the excessive exposure of people and to prevent loss of radioactive material during transport.

D. Radioactive material, i.e., that which is subject to labeling in accordance with 10 CFR 20.1904 and 10 CFR 20.1905, shall be stored only in designated storage areas.

Exempt quantities of radioactive material covered by a general license issued by the NRC are not subject to this requirement.

E. In addition to the above requirements, personnel involved in the packaging, handling, and transfer of radioactive material shall comply with the requirements of the Radioactive Material Shipment Manual (RMSM).

3.7.2 Labeling*

All containers of radioactive material shall be labeled in accordance with 10 CFR 20.1904, unless exempted by 10 CFR 20.1905.

3.7.3 Receipt*

Receipt surveys of packages or radioactive material shall be conducted in accordance with the requirements of SPP-5.6, "Controlling Byproduct and Source MateriaL" Proper precautions shall be established for opening and handling the packages.

3.7.4 General Requirements*

A. The RPM/RSO determines the appropriate radioactivity analysis configuration for bulk material (e.g., sample size, counting times, geometry, and equipment) such that no detectable radioactive material will be released from the site or facility.

B. Liquids, sludge, sand, gravel, and other such materials contaminated as the result of plant operations shall not be released off site unless approval is granted by the NRC.

This does not preclude releases covered by plant technical specifications. Soil in clean areas may contain trace quantities of radionuclides but need only be controlled for release to unrestricted areas.

3.8 Incident Reporting and Investigation Radiological Incident Investigation and reporting shall be in accordance with the NPG Corrective Action Program.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 36 of 46 3.9 RP Instruments Environmental Radiological Monitoring and Instrumentation Laboratory (ERMI) shall provide program definition and control for the Radiation Protection instrument program.

ERMI is responsible for the coordinated purchase of Radiation Protection portable instruments within the scope of this procedure. ERMI is also normally responsible for primary calibration and maintenance of these instruments.

NOTE Radiation Protection material/equipment shall be procured as QA Level 0 unless the material/equipment directly interfaces with a safety-related or another augmented QA system. If not procured as QA Level 0 then materials and equipment will be procured per the applicable procedures.

Each site is responsible for appropriate usage of, inventory control and accountability of, and timely response and source checking of the Radiation Protection instruments in its possession.

3.9.1 Calibration*

A. Calibration Frequency All Radiation Protection instruments aSSigned for operational in-field use shall be calibrated at a frequency commensurate with in-field performance.

B. Calibration Facilities Calibration facilities and equipment, including radiation sources and calibration procedures, shall be established using ANSI N323-1978 as guidance.

3.9.2 Procurement*

A. Routine Radiation Protection instrument purchases shall be made by ERMI based upon requests from each site and the list described in SubSection 3.9.3D.

B. Special purchases may be made by each site to fulfill unanticipated needs.

Such purchases shall be coordinated between the site and ERMI, so that proper inventories and calibration schedules can be maintained.

C. All new portable RP instruments shall be delivered to the calibration facility for acceptance testing and calibration before field use.

D. Exceptions to these procurement requirements may be granted by the Manager, ERMI on a case-by-case basis.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 37 of 46 3.9.3 Control and Use*

A. Response Checks

1. Instruments shall be response and source checked at a frequency which is sufficient to demonstrate proper instrument operation.

2. If an instrument fails a response or source check, it shall be tagged and segregated from operational instruments and an assessment made and documented of the validity of measurements performed with that instrument since its previous successful response check.

B. Instrument User Training A method shall be established to ensure that the user is trained in the proper use and care of RP instruments.

C. Instrument Accountability

1. Each site shall employ an accounting system for Radiation Protection instruments. This system shall contain information necessary for internal inventory control.

2. The primary calibration facility shall also employ an accounting system for RP instruments. This system shall contain information necessary for establishing the location and calibration status of each instrument.

D. Instrument Inventory Each site should develop a list of instruments and corresponding quantities necessary for maintaining the capability of performing adequate surveys. Instrument levels should be maintained at these values.

3.10 Radiation Protection Quality Assurance Requirements Radiation Protection Quality Assurance requirements are found in SPP-5.9, "Radiological Control and Radioactive Material Shipment Augmented Quality Assurance Program."

4.0 RECORDS There are no specific records generated by this procedure.

S.O DEFINITIONS*

Administrative Dose Levels (ADLs) - A numerical dose constraint established at a level below the regulatory limits set forth in 10 CFR Part 20, which are established as a guideline to administratively control and help optimize individual and collective radiation exposure.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 38 of 46 5.0 DEFINITIONS* (continued)

Airborne Radioactivity Area - A room, enclosure, or area in which airborne radioactive materials, composed wholly or partly of licensed material, exist in concentrations - (1) in excess of the derived air concentrations specified in Appendix 8 to 10 CFR 20, or (2) to such a degree that an individual present in the area without respiratory protective equipment could exceed, during the hours an individual is present in a week, an intake of 0.6 percent of the annual limit on intake or 12 DAC-hours.

Annual Limit on Intake (AU) - The derived limit for the amount of radioactive materials taken into the body of an adult worker by inhalation or ingestion in a year. One ALI of a given radionuclide would result in a CEDE of 5 rems or a CDE of 50 rems to any individual organ or tissue.

Bioassay - Is the determination of the kind, quantity, concentration, and location of radioactive material in the human body by direct measurement (in vivo) or indirect analysis of excreta (in vitro). RWP entry and corresponding air sample records are acceptable alternatives to vivo and in vitro bioassay.

Bioassay Area - Means any airborne radioactivity area and any other area where unencapsulated radioactive material is present in a form and quantity such that the area has significant potential for becoming an airborne radioactivity area. Entry into bioassay areas is governed by RWPs.

Clean Area - An area below radioactive contamination levels as in Section 3.5.3A where radioactive materials are not normally permitted.

Committed Dose Equivalent (CD E) - The dose equivalent to organs or tissues that will be received from an intake of radioactive material by an individual during the 50-year period following the intake.

Committed Effective Dose Equivalent (CEDE) - The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues.

Contaminated Area - A radiologically controlled area in which uncontained, removable radioactive material (contamination) is present in excess of the levels of SubSection 3.5.3A of this document.

Controlled Area - An area, outside of a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason.

Critical level of Detection (Lc) - The critical level of (Lc) is the level at which there is a statistical probability (with a predetermined confidence) of incorrectly identifying a background value as "greater than background". Any response above this level is considered to be greater than background and will be reported and recorded as detected.

Declared Pregnant Woman - Means a woman who has voluntarily informed the licensee, in writing, of her pregnancy and the estimated date of conception. The declaration remains in effect until the declared pregnant woman withdraws the declaration in writing or is no longer pregnant.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006

) Processes Page 39 of 46 S.O DEFINITIONS* (continued)

Deep Dose Equivalent (DDE) - Applies to external whole-body exposure. The dose equivalent at a tissue depth of 1 cm (1000 mg/cm 2 ).

Derived Air Concentration (DAC) - The concentration of a given radionuclide in air which, if breathed by the reference man for a working year of 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months under conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI. For modes of intake other than inhalation, an equivalent DAC-hr shall be determined and included as DAC-hrs in the individual's dose tracking record. The equivalent DAC-hr is equal to the number of hours of exposure at the DAC (i.e., DAC-hrs exposure which would result in an equivalent intake of radioactive material as has been observed in a particular exposure incident). This permits the comparison of intake by inhalation with other modes of intake (ingestion, injection, absorption, etc.). 2000 DAC-hr is equal to one ALI.

Dose - A generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent, as defined in applicable sections of 10 CFR 20.

NOTE For purposes of this document and implementing procedures, radiation exposure as expressed in units of R/hr and subunits, thereof, is equivalent to dose (rad) and dose equivalent (rem). Based on ANSI N13.11 development and terminology, any

) acute dose greater than 10 rem is generally denoted in units of rad, since that level is considered as the accident range of personnel exposure. Any dose less than that level is considered the protective range of personnel exposure.

Evaluation Level (EL) - The level above which a bioassay or air monitoring result shall be investigated, to the extent reasonable, to determine actual conditions and parameters for dose evaluation. An investigation may involve special measurements, work history review, determination of material form, and modification of biokinetic parameters. Investigate incident, as appropriate, to determine probable cause and the extent of condition. As appropriate, initiate any possible corrective actions or additional measures to prevent recurrence. The EL is 50-mrem committed effective dose equivalent.

Experience - As used in this document and ANSI N18.1-1971and ANSI/ANS-3.1 (1981),

actual applicable working experience performing duties commensurate with the position.

Observation of others is not considered experience. Up to 12 months of OJT may be credited toward experience on a one-for-one basis.

Fixed Contamination - Contamination which is not transferred through casual contact and is not detected by smear survey. It may become removable through operations such as grinding, welding, etc.

High Radiation Area (HRA) - An area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 100 mrem in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 30 centimeters from the radiation source or 30 centimeters from any surface that the radiation penetrates.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 40 of 46 S.O DEFINITIONS* (continued)

Hot Particle - A single discrete object (particle) generally difficult to see (usually

<100 micron) with the naked eye, and at least 0.1 microcuries of radioactivity. It is either an activated corrosion/wear product or fuel fragment with high specific activity. For the purpose of an approximate field calculation, any ORP surveyed with a standard frisker probe (HP-260, HP-210, etc.) and found to have levels of greater than or equal to 20,000 cpm, shall be considered a hot particle.

Individual Monitoring Devices (individual monitoring equipment) - Means devices designed to be worn by a single individual for the assessment of dose equivalent such as film badges, thermoluminescence dosimeters (TLOs), pocket ionization chambers, electronic dosimeters, and personal ("lapel") air sampling devices.

Lens Dose Equivalent (LDE) - Applies to the external exposure of the lens of the eye and is taken as the dose equivalent at a tissue depth of 0.3 centimeter (300 mg/cm 2 ).

Licensee - Means the holder of a license.

Member of the Public - Any individual except when that individual is receiving an occupational dose.

Occupational Dose - The dose received by an individual in the course of employment in which the individual's assigned duties involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation, whether in the possession of the licensee or other person. Occupational dose does not include dose received from background radiation, as a patient from medical practices, or from voluntary participation in medical research, or as a member of the public.

On-Site Experience - Applicable work performed at the plant for which the individual seeks qualification. Work shall involve that plants systems and procedures. Observation of others performing work is not experience. In those cases where the collective experience does not exceed the sum of the minimum for individual positions, support shall be provided by additional personnel so that the collective experience exceeds the sum of the minimum.

On-The-Job Training (OJT) - Performance of duties, commensurate with the level to which the training will be credited, under the direction of appropriately experienced personnel.

Planned Special Exposure (PSE) - An infrequent exposure to radiation, separate from and in addition to the annual dose limits.

RP Instrument - Any Radiation Protection instrument used (not including installed facility radiation monitoring system) to measure radiation exposure, exposure rate, dose, dose rate, dose equivalent, or dose equivalent rate or to assess airborne or surface contamination.

Instruments utilized in the external and internal dosimetry programs are excluded from this definition.

RP Technician - A technician qualified in radiation protection and serving in a responsible position per ANSI N18.1-1971.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 41 of 46 5.0 DEFINITIONS* (continued)

Radiation Area - An area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 5 mrem in one hour at 30 cm from the radiation source or from any surface that the radiation penetrates.

Radiation Work Permit (RWP) - A document for controlling the radiological aspects of work.

Radiologically Controlled Area (RCA) - An area within (or that may coincide with) the Restricted Area (defined in 10 CFR 20.1003) boundaries that may have increasing radiological hazards.

Removable Contamination - Contamination which may be easily transferred to personnel or surfaces through casual contact.

Response Check - Exposure of the instrument to radiation in a reproducible geometry such that a reading is obtained for each scale or decade normally used in order to verify that the instrument response is acceptable for performing surveys.

Restricted Area - Any area access to which is limited by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials (10 CFR 20.1003).

Screening Level (SL) - The level below which a follow-up bioassay measurement need not be considered for investigation of intake and assignment of dose. The SL is based on a committed effective dose equivalent of 10 mrem. Calculated committed effective dose equivalent equal to or greater than the SL are recorded and included in an individual's Total Effective Dose Equivalent. Bioassay measurements indicating above the SL should have sufficient follow-up bioassays to accurately define the amount of radioactive material intake and the appropriate biokineticpathway.

Shallow Dose Equivalent (SDE) - Applies to the external exposure of the skin or extremity.

The dose equivalent at a tissue depth of 0.007 cm (7 mg/cm 2 ) averaged over a contiguous area of ten square centimeters receiving the highest dose.

Source Check - Similar to response check except that only one scale must be checked.

Total Effective Dose Equivalent (TED E) - The sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

Total Organ Dose Equivalent (TODE) - The sum of the deep dose equivalent and the committed dose equivalent to the organ of interest.

Unplanned Intakes - Intake of radioactive material for which no documented TEDE ALARA evaluation was performed and result in an internal dose at 10 mrem.

Very High Radiation Area - An area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving an absorbed dose in excess of 500 rads in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 1 meter from a radiation source or 1 meter from any surface that the radiation penetrates.

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 42 of 46 5.1 Acronyms/Abbreviations ADL - Administrative Dose Level ALARA - As Low As Reasonably Achievable ALI - Annual Limit on Intake CDE - Committed Dose Equivalent CEDE - Committed Effective Dose Equivalent CPM - Counts Per Minute DAC - Derived Air Concentration DDE - Deep Dose Equivalent DPM - Disintegrations Per Minute DRP - Discrete Radioactive Particle EL - Evaluation Level ERMI - Environmental Radiological Monitoring and Instrumentation HRA - High Radiation Area LDE - Eye Dose Equivalent NPG - Nuclear Power Group NRC - Nuclear Regulatory Commission NVLAP - National Voluntary Laboratory Accreditation Program OJT - On-The-Job Training PCM - Personnel Contamination Monitor PSE - Planned Special Exposure REAG - Radiation Effects Advisory Group RP - Radiation Protection RPM - Radiation Protection Manager RSO - Radiation Safety Officer RWP - Radiation Work Permit SAM - Small Article Monitor

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 43 of 46 5.1 Acronyms/Abbreviations (continued)

SDE-WB - Shallow Dose Equivalent - Whole Body SED - Site Emergency Director SL - Screening Level TEDE - Total Effective Dose Equivalent TLD - Thermoluminescent Dosimeter TODE - Total Organ Dose Equivalent

6.0 REFERENCES

6.1 Source Documents Title 10 Code of Federal Regulations Part 19 "Notices, Instructions and Reports to Workers; Inspection and Investigations."

Title 10 Code of Federal Regulations Part 20, "Standards for Protection Against Radiation" Title 10 Code of Federal Regulations Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material" Title 10 Code of Federal Regulations Part 31, "General Domestic Licenses for Byproduct Material" Title 10 Code of Federal Regulations Part 32, "Specific Domestic Licenses to Manufacture or Transfer Certain Items Containing Byproduct Material" Title 10 Code of Federal Regulations Part 33, "Specific Domestic Licenses of Broad Scope For Byproduct Material" Title 10 Code of Federal Regulations Part 34, "Licenses for Radiography and Radiation Safety Requirements for Radiographic Operations" BFN, SQN, and WBN Technical Specifications Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A u.S. NRC Regulatory Guide 1.8, "Personnel Selection and Training," Revisions 1 and 2 ANSI N18.1-1971, "Selection and Training of Nuclear Plant Personnel" ANSIIANS-3.1-1981 u.S. NRC Regulatory Guide 4.15, "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment," 1977

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 44 of 46 6.1 Source Documents (continued)

U.S. NRC Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Plants," Revision 1, June 1974 6.2 Developmental References U.S. NRC Regulatory Guide 1.16, "Reporting of Operating Information - Appendix A Technical Specifications," August 1975 U.S. NRC Regulatory Guide 8.8. "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Stations will be As Low As is Reasonably Achievable," June 1978 U.S. NRC Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational Radiation Exposure As Low As Reasonably Achievable," Revision 1, September 1975 U.S. NRC Regulatory Guide 8.13, Revision 2, "Instructions Concerning Prenatal Radiation Exposure," December 1987 U.S. NRC Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection,"

October 1976 U.S. NRC Regulatory Guide 8.27, "Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Plants," April 1981 U.S. NRC Regulatory Guide 8.29, "Instructions Concerning Risks from Occupational Radiation Exposure," July 1981 ANSI N13.11-1983, "Personnel Dosimetry Performance - Criteria for Testing" ANSI N45.2-1971, "Quality Assurance Program Requirements for Nuclear Plants" ANSI N323-1978, "Radiation Protection Instrumentation Test and Calibration" SPP-2.1 - "Administration of Standards Programs and Procedures and Standard Department Procedures" International Electrotechnical Commission (IEC) 395-1972, "Portable X or Gamma Radiation Exposure Rate Meters and Monitors for Use in Radiological Protection" "Guidelines for Radiological Protection at Nuclear Stations," INPO 05-008, December 2005 National Council on Radiation Protection and Measurements Report No. 91, "Recommendations on Limits for Exposure to Ionizing Radiation," June 1, 1987 SPP-2.3, "Document Control" SPP-2.4, "Records Management" SPP-4.1, "Procurement of Material and Services"

NPG Standard Radiological Controls SPP-5.1 Programs and Rev. 0006 Processes Page 45 of 46 6.2 Developmental References (continued)

SPP-4.2, "Material Receipt and Inspection" TVA Nuclear Training Manual SPP-2.6, "Control and Use of Computer Software" SPP-9.3, "10CFR50.59 Evaluations of Changes, Tests, and Experiments" ANSI N413-1984, "Guidelines for the Documentation of Digital Computer Programs" Institute of Electrical and Electronics Engineers Standard 829-1983, "IEEE STANDARD FOR SOFTWARE TEST DOCUMENTATION" SPP-5.2, "ALARA Program" SPP-5.6, "Controlling Byproduct and Source Material" INPO 90-015 "Performance Objectives and Criteria for Operating and Near-Term Operating License Plants" ANI/MAELU Information Bulletin 80-1A, Revision 4, "Nuclear Liability Insurance Records Retention" ANSI N13.30-1990, "Performance Criteria for Radiobioassay" Radioactive Material Shipment Manual U.S. NRC Regulatory Guide 8.7, "Instructions for Recording and Reporting Occupational Radiation Exposure Data," June 1992 U.S. NRC Regulatory Guide 8.9, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program," July 1993 U.S. NRC Regulatory Guide 8.34, "Monitoring Criteria and Methods to Calculate Occupational Radiation Doses," July 1992 U.S. Nuclear Regulatory Guide 8.35, "Planned Special Exposures," June 1992 U.S. Nuclear Regulatory Guide 8.36, "Radiation Dose to the Embryo/Fetus," July 1992 U.S. Nuclear Regulatory Guide 8.38, "Control of Access to High and Very High Radiation Areas in Nuclear Plants," June 1993 ANSIIASQC Z1.4-1981, Sampling Procedures and Tables for Inspection by Attributes ANIIMAELU, Inspection Criteria 8.1 through 8.10.

U.S. Department of Commerce, National Institute of Standards and Technology, HANDBOOK 150 NVLAP PROCEDURES AND GENERAL REQUIREMENTS.

NPG Standard Radiological Controls SPP-S.1 Programs and Rev. 0006 Processes Page 46 of 46 6.2 Developmental References (continued) u.s. Department of Commerce, National Institute of Standards and Technology, HANDBOOK 150-4 NVLAP IONIZING RADIATION DOSIMETRY USNRC Technical Communication, Radiation Records Data Collection and Analysis, January 4,1994, NRC Contract NRC-04-91-054, SAIC No. 1-264-03-789-01 ANI Inspection Criteria "Radiation Protection" NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR 20 NRC Health Physics Position (HPPOS) 246 NRC Information Notice No. 90-48 Enforcement Policy for Hot Particle Exposures (NRC, 1990).

JPM NO. 480TC REV. NO. 0 PAGE 1 of20 JPM NUMBER: 480TC TITLE: CLASSIFY THE EVENT PER THE REP (GENERAL - LOSS OF ALL PWR TO ALL UNIT SPECIFIC 4KV SID BDS >3 HOURS)

ADMIN: Emergency Plan (SRO ONLY)

SUBMITTED BY: DATE: _ _ __

VALIDATED BY: DATE: _ _ __

APPROVED BY: DATE: _ _ __

TRAINING PLANTCONCURRENCE: _ _ _ _ _ _ _ _ ___ DATE: _ _ __

OPERATIONS

Examination JPMs Require Operations Training Manager Approval or Designee Approval and Plant Concurrence

)

JPM NO. 480TC REV. NO. 0 PAGE 2 of20 REVISION LOG Revision Effective Pages Description Number Date Affected Of Revision 0 08/08/08 All Initial issue

)

JPM NO. 480TC REV. NO. 0 PAGE 3 of 20 OPERATOR: _____________________________________________

RO SRO DATE: __----,--_______

JPM NUMBER: 480TC TASK NUMBER: S-000-EM-21 (SRO ONLY)

TASK TITLE: CLASSIFY THE EVENT PER THE REP (GENERAL - LOSS OF ALL PWR TO ALL UNIT SPECIFIC 4KV SID BDS >3 HOURS)

KIA NUMBER: 2.4.38 KIA RATING: RO 2.2 SRO 4.0 TASK STANDARD: CLASSIFY THE EVENT AS A GENERAL EMERGENCY (S.1-G)

BASED ON LOSS OF ALL POWER TO UNIT SPECIFIC 4KV SHUTDOWN BDS ON ANY UNIT FOR >3 HOURS. MAKE NOTIFICATIONS SUCH THAT; (TIME ODS NOTIFIED) - (TIME DECLARED) S S MINUTES (TIME NRC NOTIFIED) - (TIME DECLARED) S 60 MINUTES PERFORMANCE LOCATION: SIMULATOR X PLANT - CONTROL ROOM_

REFERENCES/PROCEDURES NEEDED: EPIP-1, Rev 43, EPIP-S, Rev 37 VALIDATION TIME: SIMULATOR: 20 min LOCAL: _____

MAX. TIME ALLOWED: S/60 (FOR TIME CRITICAL JPMs ONLY)

PERFORMANCE TIME:

COMMENTS:

ADDITIONAL COMMENT SHEETS ATTACHED? YES _ _ NO RESULTS: SATISFACTORY UNSATISFACTORY EXAMINER SIGNATURE: DATE: _ _ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 40f20 IN-SIMULATOR: I will explain the initial conditions and state the task to be performed.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied~ When your task is given, you will repeat the task and I will acknowledge "That's Correct". (OR "That's Incorrect", if applicable). When you have completed your assigned task, you will say, "my task is complete" and I will acknowledge that your task is complete.

INITIAL CONDITIONS: You are the SHIFT MANAGER. Unit 3 was in MODE 2 at 2% power when a severe storm caused damage to the switchyard with loss of ALL OFFSITE POWER at 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months . Also, an unisolable rupture occurred in the EECW system for Unit 3 Diesels and cannot be repaired for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . All Unit 3 Diesels are lost due to the loss of cooling water. Unit 3 shutdown boards cannot be crosstied to Unit 1/2. EOI-1 has been entered and all rods inserted on the scram; SBGT A & Bare operating and no elevated radiological stack release is predicted.

INITIATING CUES: The UNIT SUPERVISOR has informed you of the EECW line rupture causing loss of all Unit 3 Diesels with an estimated time of repair being 1700. It is now 1314. Using the following parameters provided to you by the Control Room operating crew, CLASSIFY THE EVENT according to the EPIPs and perform any required actions. The TSC and CECC are not staffed.

Reactor Level -40 inches on Emergency Range, controlled by RCIC Reactor Pressure 950 controlled by SRV's (MSIV's isolated)

OW Pressure 1.38 psig OW Temperature 145 of Torus Temperature 91°F PSC Pressure 1.0 psig Torus Level -2 inches Wind Speed 5 mph, Wind Direction/North (SOME portions of this JPM are TIME CRITICAL)

)

JPM NO. 480TC REV. NO. 0 PAGE 50f20 START TIME _ _ _ _ __

PERFORMANCE STEP: CRITICAL x NOT CRITICAL Refers to EPIP-1 to classify the emergency event STANDARD:

SRO/SED refers to EPIP 1, Section 5, Loss of AC Power and declares a GENERAL EMERGENCY (5.1-G) based on Loss of voltage to ALL unit specific 4KV Shutdown Boards from Table 5.1 AND restoration of at least one 4KV Shutdown Board is NOT likely within three hours.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

PERFORMANCE STEP: CRITICAL x NOT CRITICAL Implements EPIP-5, GENERAL EMERGENCY STANDARD:

SRO/SED recognizes/implements a GENERAL EMERGENCY per EPIP-5.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 6 of20

[ BROWNS FERRY GENERAL EMERGENCY I EPIP-4 3.0 EMERGENCY CLASSIFICATION ACTIONS This section of the procedure is utilized for actions to be taken when the initial Site Area Emergency classification is originating from the Control Room. If the Technical Support Center is operational, utilize the instructions found in Appendix E of this procedure for actions to be taken upon the Site Area Emergency classification being declared.

3.1 Activation of the Emergency Response Organization (ERO)

CAUTION Ongoing or anticipated security events may present a danger to normal staffing of the Emergency Response Organization. Select the "Staging Area" option when events are ongoing or anticipated that may present a danger to normal ERO staffing as determined by the SED and/or Nuclear Security.

NOTE Normally Appendix B, "Unit Operator Notifications", is conducted by a Unit 1, Unit Operator, Depending upon the affected unit, this action may be delegated toa Unit Operator on an unaffected unit.

)

JPM NO. 480TC REV. NO. 0 PAGE 7 of 20 TIME EVENT DECLARED _ _ __

PERFORMANCE STEP: CRITICAL x NOT CRITICAL 3.1.1 NOTIFY ... a Unit Operator of the Site Area Emergency Classification, AND 3.1.2 DIRECT. .. the Unit Operator to implement Appendix 8, activating the paging system using option;

DRILL

EMERGENCY

STAGING AREA (See caution note above)

STANDARD:

DIRECTS Unit Operator to make notifications per Appendix 8.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 8 of 20

')

PERFORMANCE STEP: CRITICAL x NOT CRITICAL 3.2 Operations Duty Specialist (ODS) Notification / State of Alabama Notification Note

1. The ODS should be notified within 5 minutes after the emergency has been declared.

2. Completion of Appendix A for the General Emergency Classification includes the development of a Protective Action Recommendation (PAR) when the CECC is not operational. Utilize Appendix G, PAR flowchart when determining PAR, if required. PAR must be made by SM/SED 3.2.1 Complete Appendix A (Initial Notification Form) Utilize Appendix G, "Protective Action Recommendation", flowchart as appropriate.

STANDARD:

APPENDIX A is complete with EAL Designator 5.1-G GENERAL EMERGENCY status and a PAR Recommendation 2. EOI-1 has been entered and all rods inserted on the scram. Loss of ALL OFFSITE power and EECW piping rupture causing loss to Unit 3 Diesel Generators--estimated time of repair for leak is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Unable to crosstie 4KV Shutdown Boards with Unit 1/2. Reactor level -40 inches on Emergency Range controlled by RCIC (MSIV's are isolated). Reactor pressure 950 controlled by SRV's, OW pressure 1.38 psig, OW temperature 145 OF, Torus temperature 91°F, Torus level-2 inches, Torus pressure 1.0 psig. Wind speed is 5 MPH and direction is North. Unit 3 conditions are fairly stable with no abnormal radiological releases offsite.

(Classification of event and PAR Recommendation are CRITICAL, description is NOT.)

SAT UNSAT N/A _ _ COMMENTS:---'-_ _ _ __

)

JPM NO. 480TC REV. NO. 0 PAGE 9 of 20 PERFORMANCE STEP: CRITICAL x NOT CRITICAL 3.2.2 NOTIFY ... the ODS, utilizing the "Direct Ring-Down" telephone or at extension 5-751-1700 or 5-751-2495.

AND REPORT ... to the ODS the information recorded on Appendix A.

AND FAX ... a copy of Appendix A to the ODS for confirmation of information at 5-751-8620.

STANDARD:

Contacts the ODS within 5 minutes of declaring the event and simulates sending fax.

(Only contacting the ODS within 5 minutes is Critical)

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

RECORD TIME ODS NOTIFIED

JPM NO. 480TC REV. NO. 0 PAGE 10 of 20 PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.2.3 IF ... the ODS was contacted, THEN ... the State of Alabama notification action is complete.

AND RE-ENTER at Step 3.3. Otherwise continue.

STANDARD:

Continues to step 3.3, since ODS was notified.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

NOTE

The State of Alabama should be contacted within 15 minutes of the emergency classification.

Completion of Appendix A for the General Emergency Classification includes the development of a Protective Action Recommendation (PAR) when the CECC is not operational. Utilize Appendix G, PAR flowchart when determining PAR, when required. PAR must be made by SM/SED.

JPM NO. 480TC REV. NO. 0 PAGE 11 of 20 PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.2.4 IF ... the ODS cannot be contacted within 10 minutes, THEN ... NOTIFY and REPORT the information recorded on Appendix A to the following:

Limestone County 9-232-2631 (after hours) 9-232-0111 Morgan County 9-1-256-351-4620 (after hours) 9-1-256-353-2515 Option 0 Lawrence County 9-1-256-974-7641 (after hours) 9-1-256-974-7911 Lauderdale County 9-1-256-7664201 (after hours) 9-1-256-760-9117 State of Alabama at:

24 Hours Primary: 9-1-205-280-2310 Backup: 9-1-800-843-0699 Backup: 9-1-334-324-0076 AND FAX ... a copy of Appendix A to the State of Alabama for confirmation of information at 9-1-205-280-2495.

STANDARD:

N/A - The ODS was notified.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 12 of 20 PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.3 ODS State of Alabama Notification Confirmation Receive a confirmation call from the ODS verifying that the notification of the State of Alabama was completed. Do this concurrently with the implementation of this procedure.

STANDARD:

Continues in procedure until conformation call is received and acknowledges receipt.

SAT UNSAT N/A _ _ COMMENTS: _ _ _---,-_ _

)

JPM NO. 480TC REV. NO. 0 PAGE 13 of20 CAUTION Ongoing or anticipated security events may present a danger to site personnel. Do not conduct the notification of site personnel PA message during an ongoing or anticipated security event. All pertinent site personnel PA messages will be conducted per AOI-100-8 for security events.

PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.4 Notification of Site Personnel CONDUCT a Plant PA announcement similar to the following:

(Dial 687 to obtain the Plant PA)

Let me have your attention please.

This is (name) _ _ _ _ _ _ .

A General Emergency Classification has been declared.

We are currently implementing EPIP-5.

If you have not already done so, please report to your assigned emergency center at this time.

STANDARD:

P. A. Announcement was made giving name, General Emergency status on Unit 3, and informs plant personnel that EPIP-5 is being implemented and directs plant personnel to report to their assigned Emergency Response Facility, if they haven't already done so.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 14 of 20 CAUTION Do not initiate Assembly / Accountability when:

1. A severe weather condition exists or is projected on-site, such as a Tornado.

2. An on-site security risk condition exists that may present a danger to site personnel during the Assembly / Accountability process as determined by SED/Nuclear Security.

PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.5 Assembly / Accountability 3.5.1 IF ... Assembly / Accountability has not been conducted, THEN ... IMPLEMENT EPIP-8, Appendix C concurrently with this procedure. This action may be delegated.

3.5.2 IF ... an order to evacuate non-emergency responders has not been issued, THEN ... upon completion of Assembly / Accountability, INITIATE the order to "Evacuate Non-Emergency Responders," through implementation of EPIP-8, Appendix F, concurrently with this procedure.

3.5.3 IF... conditions exist that do not allow for an Assembly /

Accountability or Evacuation at this time, THEN ... CONTINUE to assess the situation, implementing EPIP-8 as applicable.

STANDARD:

Acknowledges that STA is performing EPIP-8 and continues to step 3.6 SAT UNSAT N/A _ _ COM ME NTS:___________

JPM NO. 480TC REV. NO. 0 PAGE 15 of 20

')

PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.6 Dose Assessment EVALUATE. .. the need for dose assessment.

IF ... dose assessment is needed, THEN ... CONTACT, if operational, the Central Emergency Control Center (CECC) at 5-751-1614.

OR IF ... the CECC is not operational, THEN ... CONTACT, the Radiological Protection Shift Supervisor or designee at 7865 and request the implementation of EPIP-13 for dose assessment.

STANDARD:

SRO/SED acknowledges that the CECC is not staffed and contacts the Radiological Protection Shift Supervisor and request the implementation of EPIP-13, if deemed necessary.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 16 of 20 PERFORMANCE STEP: CRITICAL x NOT CRITICAL 3.7 Notification of the Nuclear Regulatory Commission (NRC)

NOTE If possible, when making notifications to the NRC, utilize the Emergency Notification System (ENS). Dial the first number listed on the sticker affixed to the ENS telephone by dialing 9 "The Ten Digit Number Listed on the ENS Telephones". If the number is busy, then select in order, the alternate numbers until a connection is achieved. No access codes should be required.

NOTIFY ... the NRC immediately but no later than one hour after the emergency has been declared.

IF ... REQUESTED by the NRC to maintain an open and continuous line of communications, THEN ... MAINTAIN an open and continuous line of communications as directed by NRC.

STANDARD:

SRO/SED notified NRC within 60 minutes on the Simulator by calling the console operator and requesting NRC.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

RECORD TIME NRC NOTIFIED

JPM NO. 480TC REV. NO. 0 PAGE 17 of 20 PERFORMANCE STEP: CRITICAL NOT CRITICAL x 3.8 Review of Procedure Review this procedure to ensure that all steps and actions have been completed and all place keeping blocks have been checked or denoted as instructed. This action may be delegated.

STANDARD:

SRO/SED reviews procedure to ensure all steps and actions have been completed, placekeeping blocks checked as instructed.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

)

JPM NO. 480TC REV. NO. 0 PAGE 18 of 20 3.9 Monitor / Re-evaluate the Event Monitoring and reevaluation of plant events along with communicating significant changes should be performed continuously as a function of the emergency response. Methods used to communicate significant changes are not formalized and may vary depending upon staffing levels as well as availability of personnel or equipment. Appendix C provides a systematic approach to monitor/reevaluate and communicate significant changes in plant conditions.

Utilize Appendix C to monitor/re-evaluate and communicate plant conditions and significant changes. Significant changes in plant conditions are at a minimum when other EAL conditions exist indicating the current emergency classification.

END OF TASK STOP TIME,_ _ _ __

JPM NO. 480TC REV. NO. 0 PAGE 19 of 20 PERFORMANCE STEP: CRITICAL NOT CRITICAL x PERFORMER demonstrated the use of SELF CHECKING during this JPM STANDARD:

PERFORMER verified applicable components by utilizing SELF CHECKING in accordance with plant standards.

SAT UNSAT N/A _ _ COMMENTS:, _ _ _ _ __

PERFORMANCE STEP: CRITICAL NOT CRITICAL x PERFORMER demonstrated the use of 3-WAY COMMUNICATION during this JPM STANDARD:

PERFORMER utilized 3-WAY COMMUNICATION in accordance with plant standards.

SAT UNSAT N/A _ _ COMMENTS: _ _ _ _ __

PAGE 20 of 20 IN-SIMULATOR: I will explain the initial conditions and state the task to be performed.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. When your task is given, you will repeat the task and I will acknowledge "That's Correct". (OR "That's Incorrect", if applicable). When you have completed your assigned task, you will say, "my task is complete" and I will acknowledge that your task is complete.

INITIAL CONDITIONS: You are the SHIFT MANAGER. Unit 3 was in MODE 2 at 2% power when a severe storm caused damage to the switchyard with loss of ALL OFFSITE POWER at 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months . Also, an unisolable rupture occurred in the EECW system for Unit 3 Diesels and cannot be repaired for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . All Unit 3 Diesels are lost due to the loss of cooling water. Unit 3 shutdown boards cannot be crosstied to Unit 1/2. EOI-1 has been entered and all rods inserted on the scram; SBGT A & Bare operating and no elevated radiological stack release is predicted.

INITIATING CUES: The UNIT SUPERVISOR has informed you of the EECW line rupture causing loss of all Unit 3 Diesels with an estimated time of repair being 1700. It is now 1314. Using the following parameters provided to you by the Control Room operating crew, CLASSIFY THE EVENT according to the EPIPs and perform any required actions. The TSC and CECC are not staffed.

Reactor Level -40 inches on Emergency Range, controlled by RCIC Reactor Pressure 950 controlled by SRV's (MSIV's isolated)

OW Pressure 1.38 psig OW Temperature 145 of Torus Temperature 91°F PSC Pressure 1.0 psig Torus Level -2 inches Wind Speed 5 mph, Wind Direction/North (SOME portions of this JPM are TIME CRITICAL)

(

BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 Loss of normal and alternate supply voltage to ALL c:

unit specific 4KV shutdown boards from Table 5.1 z for greater than 15 minutes c:

(I)

AND c:

At least two Diesel Generators supplying power to >>

r unit specific 4KV shutdown boards listing in Table 5.1. m OPERATING CONDITION: m ALL z Loss of voltage to ANY THREE unit specific 4KV Loss of voltage to ALL unit specific 4KV shutdown shutdown boards from Table 5.1 for greater than boards from Table 5.1 for greater than 15 minutes.

15 minutes AND >>

r Only ONE source of power available to the m remaining board. ~

OPERATING CONDITION:

Mode 1 or 2 or 3 I OPERATING CONDITION:

Mode 4 or 5 or Defueled Loss of voltage to ALL unit specific 4KV shutdown (I) boards from Table 5.1 for greater than 15 minutes. =i m

m S

m

0 G) m z

o OPERATING CONDITION: -<

Mode 1 or 2 or 3 Loss of voltage to ALL unit specific 4KV shutdown G) boards from Table 5.1 m z

AND m Either of the following conditions exists;

Restoration of at least one 4KV shutdown board r

~

is NOT likely within three hours. m

Adequate core cooling can NOT be assured. s:

m

~

m z

(")

OPERATING CONDITION: -<

Mode 1 or 2 or 3 PAGE 47 OF 206 REVISION 43

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE EPIP- 5 GENERAL EMERGENCY REVISION 37 PREPARED BY: RANDY WALDREP PHONE: 2038 RESPONSIBLE ORGANIZATION: EMERGENCY PREPAREDNESS APPROVED BY: Tony Elms DATE: 03/30/2008 EFFECTIVE DATE: 04/10/2008 LEVEL OF USE: REFERENCE USE QUALITY-RELATED

HISTORY OF REVISION I REVIEW REV. REVISED REASON FOR CURRENT REVISION NO. PAGES 34 11 IC-42 EPIP-5, rev. 34 - The Protective Action Recommendation Logic Diagram is being revised to include a specific sheltering recommendation for a short term release. This change was made to incorporate criteria contained in NRC Regulatory Issue Summary 2004-13, Consideration of Sheltering in Licensee's Range of Protective Action Recommendations. Additionally the direct contact telephone number for Lawrence County is being changed at Lawrence County's request. The current number is a Emergency Management Agency number, but the new revised number is more accessible.

35 ALL IC-43 EPIP-5, rev. 35 reflects formatting changes to increase ease of use. The guidance for monitoring/re-evaluating the event was moved to Appendix C. The follow-up information form became Appendix 0 (previously Attachment D). The instructions for TSC implementation of EPIP-5 was moved to Appendix E. A flow illustration was added as Appendix F. The Protective Action Recommendation chart became Appendix G (previously Attachment C).

Additionally, the revision incorporates identified changes resulting from annual review, standardization issues, areas for improvements identified by users, cautions regarding onsite protective actions (RIS 2004-15) as well as other editorial changes.

36 4,11,18 IC-44 EPIP-5, revision 36 converted the document from W95 to XP and added a new phone number for Lauderdale Country EMA. Added caution statement to Appendix B for Unit Operator actions prior to steps 3-6.

37 4,6,7,14,18, IC-45 Note supporting step 3.2.4 and Appendix Estep 1.3 revised to change "should" 20,21 regarding state notifications to "shall". Caution note supporting step 3.5.1, Appendix C step 2.0 and Appendix Estep 4.1 revised to add example 3 of when assembly/accountability should not be initiated. Section 3.7 and Appendix Estep 6.0 revised to add a caution to ensure that all previous emergency classifications have been communicated to the NRC (PER 138293, Corrective Action 1).

TABLE OF CONTENTS

1.0 INTRODUCTION

..................................................................................................................................................1 1.1 Purpose ................................................................................................................................................................ 1

2.0 REFERENCES

.....................................................................................................................................................1 2.1 Industry Documents .............................................................................................................................................. 1 2.2 Plant Instructions ..................................................................................................................................................1 3.0 EMERGENCY CLASSIFICATION ACTIONS ......................................................................................................2 3.1 Activation of the Emergency Response Organization (ERO) ..................................... ,......................................... 2 3.2 Operations Duty Specialist (ODS) Notification / State of Alabama Notification ................................................... 3 3.3 ODS State of Alabama Notification Confirmation ................................................................................................. 5 3.4 Notification of Site Personnel ...............................................................................................................................5 3.5 Assembly / Accountability .....................................................................................................................................6 3.6 Dose Assessment .................................................................................................................................................6 3.7 Notification of the Nuclear Regulatory Commission (NRC) .................................................................................. 7 3.8 Review of Procedure ............................................................................................................................................7 3.9 Monitor / Evaluate the Event ................................................................................................................................7 4.0 DOCUMENTATION .............................................................................................................................................8 4.1 Emergency Records .............................................................................................................................................8 4.2 Drill and Exercise Records ..................................................................................................................................8 5.0 ILLUSTRATIONS/APPENDICES ........................................................................................................................8 Appendix A - General Emergency Initial Notification Form ...................................................................... ,................. 9 Appendix B - Unit Operator Notifications ..................................................................................................................10 Appendix C - Monitor / Re-Evaluate the Event. ........................................................................................................ 13 Appendix 0 - General Emergency Follow-up Information Form .............................................................................. 15 Appendix E - Technical Support Center General Emergency Classification Instruction .......................................... 16 Appendix F - EPIP-5 Procedure Flow Illustration .................................................................................................... 22 Appendix G- Protective Action Recommendation Flowchart .................................................................................. 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5

1.0 INTRODUCTION

1.1 Purpose The purpose of this procedure is to provide for the timely notification of appropriate individuals or organizations when the Shift Manager or the Site Emergency Director (SED) has determined through the use of EPIP-1 that an event has occurred which is classified as an General Emergency. Additionally, this procedure provides for periodic evaluation of the current situation by the Shift Manager/SED to determine whether the General Emergency should be terminated or continued.

This procedure is initiated by implementation of EPIP-1, "Emergency Classification Procedure."

Initial classifications are conducted from the body of this instruction. Classifications that are made following the Technical Support Center becoming operational is accomplished from an appendix of this procedure.

2.0 REFERENCES

2.1 Industry Documents A. NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" B. 10 CFR 50.47, Code of Federal Regulations C. 10 CFR 72.75, Code of Federal Regulations 2.2 Plant Instructions

, A. TVA Radiological Emergency Plan

/

B. EPIP - 1, "Emergency Classification Procedure" C. EPIP - 2, "Notification of Unusual Event" D. EPIP - 3, "Alert" E. EPIP - 4, "Site Area Emergency" PAGE 1 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 3.0 EMERGENCY CLASSIFICATION ACTIONS This section of the procedure is utilized for actions to be taken when the initial General Emergency, emergency classification is originating from the Control Room. If the Technical Support Center is operational, utilize the instructions found in Appendix E of this procedure for actions to be taken upon the General Emergency, emergency classification being declared.

3.1 Activation of the Emergency Response Organization (ERO)

CAUTION Ongoing or anticipated security events may present a danger to normal staffing of the Emergency Response Organization. Select the "Staging Area" option when events are ongoing or anticipated that may present a danger to normal ERO staffing as determined by the SED and/or Nuclear Security.

NOTE Normally Appendix B, "Unit Operator Notifications", is conducted by a Unit 1, Unit Operator, Depending upon the affected, unit this action may be delegated to a Unit Operator on an unaffected unit.

3.1.1 NOTIFY ... a Unit Operator of the General Emergency D Emergency Classification, AND 3.1.2 DIRECT ... the Unit Operator to implement Appendix B, activating the paging system using option D DRILL D EMERGENCY D STAGING AREA (See caution note above)

)

PAGE 2 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 3.2 Operations Duty Specialist (ODS) Notification I State of Alabama Notification NOTE

1. The ODS should be notified within 5 minutes after the emergency has been declared.

2. Completion of Appendix A for the General Emergency Classification includes the development of a Protective Action Recommendation (PAR) when the CECC is not operational. Utilize Appendix G, PAR flowchart when determining PAR, if required. PAR must be made by SM/SED.

3.2.1 COMPLETE Appendix A (Initial Notification Form). Utilize 0 Appendix G, "Protective Action Recommendation", flowchart as appropriate.

3.2.2 NOTIFY ... the ODS, utilizing the "Direct Ring-Down" telephone or at extension 5-751-1700 or 5-751-2495 AND REPORT ... to the ODS the information recorded on /

Appendix A. Initials Time AND FAX ... a copy of Appendix A to the ODS for confirmation of 0 information at 5-751-8620.

3.2.3 IF ... the ODS was contacted, THEN ... the State of Alabama notification action is complete.

AND RE-ENTER at Step 3.3. Otherwise continue.

PAGE 3 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 NOTE

The State of Alabama shall be contacted within 15 minutes of the emergency classification.

Completion of Appendix A for the General Emergency Classification includes the development of a Protective Action Recommendation (PAR) when the CECC is not operational. Utilize Appendix G, PAR flowchart when determining PAR, when required.

PAR must be made by SM/SED.

3.2.4 IF ... the ODS cannot be contacted within 10 minutes, THEN ... NOTIFY and REPORT the information recorded on Appendix A to the following:

Limestone County 9-232-2631 /

(after hours) 9-232-0111 Initials Time

Morgan County 9-1-256-351-4620 /

(after hours) 9-1-256-353-2515 Option 0 Initials Time

Lawrence County 9-1-256-974-7641 /

(after hours) 9-1-256-974-7911 Initials Time

Lauderdale County 9-1-256-760-6363 /

(after hours) 9-1-256-760-9117 Initials Time

State of Alabama at: /

Initials Time 24 Hours Primary: 9-1-205-280-2310 Backup: 9-1-800-843-0699 Backup: 9-1-334-324-0076 AND FAX ... a copy of Appendix A to the State of Alabama for D confirmation of information at 9-1-205-280-2495.

PAGE 4 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 3.3 ODS State of Alabama Notification Confirmation Receive a confirmation call from the ODS verifying that the /

notification of the State of Alabama was completed. Do this Initials Time concurrently with the implementation of this procedure. (N/A this step if State was contacted directly) 3.4 Notification of Site Personnel CAUTION Ongoing or anticipated security events may present a danger to site personnel. Do not conduct the notification of site personnel PA message during an ongoing or anticipated security event. All pertinent site personnel PA messages will be conducted per AOI-1 00-8 for security events.

CONDUCT a Plant PA announcement similar to the following: D (Dial 687 to obtain the Plant PA)

Let me have your attention please.

This is (name) _ _ __

A General Emergency, Emergency Classification has been declared.

We are currently implementing EPIP-5.

If you have not already done so, please report to your assigned emergency center at this time.

PAGE 5 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 3.5 Assembly I Accountability CAUTION Do not initiate Assembly / Accountability when:

1. A severe weather condition exists or is projected on-site, such as a Tornado.

2. An on-site security risk condition exists that may present a danger to site personnel during the Assembly / Accountability process as determined by SED/Nuclear Security.

3. Rapid Evacuation of the Protected Area (REPA) has been conducted.

3.5.1 IF ... Assembly / Accountability has not been conducted, D THEN ... IMPLEMENT EPIP-8, Appendix C concurrently with this procedure. This action may be delegated.

3.5.2 IF ... an order to evacuate non-emergency responders has D not been issued, THEN ... upon completion of Assembly / Accountability, INITIATE the order to "Evacuate Non-Emergency Responders," through implementation of EPIP-8, Appendix F, concurrently with this procedure.

3.5.3 IF ... conditions exist that do not allow for an Assembly / D Accountability or Evacuation at this time, THEN ... CONTINUE to assess the situation, implementing EPIP-8 as applicable.

3.6 Dose Assessment EVALUATE ... the need for dose assessment. D IF ... dose assessment is needed, THEN ... CONTACT, if operational, the Central Emergency Control Center (CECC) at 5-751-1614.

OR IF ... the CECC is not operational, THEN ... CONTACT, the Radiological Protection Shift Supervisor or designee at 7865 and request the implementation of EPIP-13 for dose assessment.

PAGE 6 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 3.7 Notification of the Nuclear Regulatory Commission (NRC)

CAUTION Ensure that previous declared emergency classifications have been communicated to NRC, specifically in fast breaking events, where emergency classifications are rapidly changing.

NOTE If possible, when making notifications to the NRC, utilize the Emergency Notification System (ENS). Dial the first number listed on the sticker affixed to the ENS telephone by dialing 9 "The Ten Digit Number Listed on the ENS Telephones". If the number is busy, then select in order, the alternate numbers until a connection is achieved. No access codes should be required.

NOTIFY ... the NRC immediately but no later than one hour after D the emergency has been declared.

IF ... REQUESTED by the NRC to maintain an open and continuous line of communications, THEN ... MAINTAIN an open and continuous line of communications as directed by NRC.

3.8 Review of Procedure Review this procedure to ensure that all steps and actions have D been completed and all place keeping blocks have been checked or denoted as instructed. This action may be delegated.

3.9 Monitor I Re-evaluate the Event Monitoring and reevaluation of plant events along with communicating significant changes should be performed continuously as a function of the emergency response. Methods used to communicate significant changes are not formalized and may vary depending upon staffing levels as well as availability of personnel or equipment. Appendix C provides a systematic approach to monitor/re-evaluate and communicate significant changes in plant conditions.

Utilize Appendix C to monitor/re-evaluate and communicate plant conditions and significant changes. Significant changes in plant conditions are at a minimum when other EAL conditions exist indicating the current emergency classification.

PAGE 7 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5

)

4.0 DOCUMENTATION 4.1 Emergency Records The records generated due to declaration of an emergency classification are considered Lifetime Retention Non-QA records. These records shall be forwarded to the BFN EP Manager. The records necessary to demonstrate performance are then submitted to the Corporate EP Manager for storage.

4.2 Drill and Exercise Records The records deemed necessary to demonstrate performance of key actions during drills are considered Non-QA records. These records shall be forwarded to the BFN EP Manager. The BFN EP Manager shall retain records necessary to demonstrate six-year plan requirements for six years. The BFN EP Manager shall retain other records in this category for three years.

5.0 ILLUSTRATIONS IAPPENDICES Appendix A - General Emergency Initial Notification Form Appendix B - Unit Operator Notifications

') Appendix C - Monitor I Re-Evaluate the Event

/ Appendix D - General Emergency Follow-up Information Form Appendix E - Technical Support Center General Emergency Classification Instruction Appendix F - EPIP-5 Procedure Flow Illustration .

Appendix G - Protective Action Recommendation Flowchart LAST TEXT

)

PAGES OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX A Page 1 of 1 GENERAL EMERGENCY INITIAL NOTIFICATION FORM

1. D This is a Drill D This is an Actual Event - Repeat - This is an Actual Event

2. This is , Browns Ferry has declared a GENERAL EMERGENCY affecting: D Unit 1 D Unit2 D Unit 3 D Common

3. EAL Designator(s):

4. Brief Description of the Event:

5. Radiological Conditions: (Check one under both Airborne and Liquid column.)

Airborne Releases Offsite Liguid Releases Offsite D Minor releases within federally approved limits 1 D Minor releases within federally approved limits 1 D Releases above federally approved limits 1 D Releases above federally approved limits 1 D Release information not known D Release information not known

( 1Tech Specs) ( 1Tech Specs)

6. Event Declared: Time: Date:

7. The Meteorological Conditions are: (Use 91 meter data from the Met Tower)

Wind Direction is FROM: degrees Wind Speed: m.p.h

8. Provide Protective Action Recommendation: Check either 1 or 2 or 3.

D Recommendation 1 R R D Recommendation 2

EVACUATE LISTED SECTORS (2 mile Radius & 10 E WIND FROM E

EVACUATE LISTED SECTORS miles downwind) C DEGREES C (2 mile radius & 5 mile downwind) I I

Shelter all other non-listed sectors.

SHELTER all other non-listed

Consider issuance of POTASSIUM IODINE in 1 (Mark wind 2 sectors.

accordance with the State Plan. direction from

Consider issuance of POTASSIUM Step 7) IODIDE in accordance with the State Plan.

A-2, B-2, F-2, G-2, E-5, -10, F-5, -10, G-5, -10 4 - 40 A-2, B-2, F-2, G-2, E-5, F-5, G-5 A-2, B-2, F-2, G-2, F-5, -10, G-5, -10, H-10 41- 73 A-2, B-2, F-2, G-2, F-5, G-5 A-2, B-2, F-2, G-2, G-5, -10, H-10, 1-10 74 - 92 A-2, B-2, F-2, G-2, G-5 A-2, B-2, F-2, G-2, A-5, G-5, H-10, 1-10, J-10,K-10 93 - 137 A-2, B-2, F-2, G-2, A-5, G-5 A-2, B-2, F-2, G-2, A-5, -10, 1-10, J-10, K-10 138 - 203 A-2, B-2, F-2, G-2, A-5 A-2, B-2, F-2, G-2, A-5, -10, B-5, -1 ° 204 - 282 A-2, B-2, F-2, G-2, A-5, B-5 A-2, B-2, F-2, G-2, B-5, -10, C-10, 0-10, E-5, -10 283 - 326 A-2, B-2, F-2, G-2, B-5, E-5 A-2, B-2, F-2, G-2, C-10, 0-10, E-5,-10, F-5,-10 327 - 3 A-2, B-2, F-2, G-2, E-5, F-5 D Recommendation 3

SHELTER all sectors

CONSIDER issuance of Potassium Iodide in accordance with the State Plan.

9. Please repeat the information you have received to ensure accuracy.

Action: When completed, fax this appendix as prescribed by procedure.

PAGE 9 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX B Page 1 of 3 UNIT OPERATOR NOTIFICATIONS NOTE

The Emergency Paging System (EPS) consists of a dedicated touch screen CRT.

Activation of any screen feature requires the user place their fingertip within the boundary of the select button and leave it there for at least 1 second. The CRT Screen will normally display a large rectangle that indicates that the paging system is available but currently inactive.

If the EPS fails to operate, contact the SM/SED immediately. Request that the ODS be contacted to initiate the system from his location. If the system fails to operate from the ODS area, then utilize the Weekly Duty List and Call-Out List to manually staff each emergency responder position, implementing this attachment at step E.

1. Activate of the Emergency Paging System (EPS)

A. PRESS the EPS CRT screen once to activate the paging options. D B. PRESS the appropriate option as instructed by the SED . D

PAGER TEST

DRILL

EMERGENCY

STAGING AREA

ABORT C. PRESS the START button to initiate the option or ABORT to D deny the option request.

PAGE 10 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX B Page 2 of 3 UNIT OPERATOR NOTIFICATIONS D. MONITOR the Paging System Terminal Display D NOTE Monitor ERO positions through OSC Document Control. Positions below OSC Document Control are courtesy pages and are not subject to call-out.

1. IF ... A "NO" response is observed, OR The position being paged has not responded within approximately 20 minutes, THEN ... Utilize the Weekly Duty List and attempt to contact the position representative with available information. (No Fitness for Duty question is required.)

2. IF ... The individual cannot be reached utilizing the Weekly Duty List, THEN ... Utilize the Call-Out List and attempt to contact an alternate position representative. (Fitness for Duty question is required when utilizing the Call-Out List. )

E. Manual Call-Out D

1. Utilize the current Weekly Duty List and contact positions as listed. (No Fitness for Duty question is required.)

2. If a position can not be reached from the current Weekly Duty list, then refer to the Call-out List as applicable to fill all vacant positions. (Fitness for Duty question is required when utilizing the Call-Out List.)

F. CONTINUE until all pOSitions have been filled. D

2. Notify Unit Supervisors on shift of the emergency. D PAGE 11 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5

) APPENDIX B Page 3 of 3 UNIT OPERATOR NOTIFICATIONS CAUTION Ongoing or anticipated security events may present a danger to site personnel. If the GENERAL EMERGENCY has been declared due to security related events, DELAY making the following notifications in steps 3-6 until verification has been received from the Shift Manager that there is no danger to site personnel.

3. Notify Nuclear Security Shift Supervisor and state "A GENERAL D EMERGENCY HAS BEEN DECLARED" and direct to activate EPIP-11, "Security and Access Control".

Plant Extension 3238 or 2219

4. Notify the Chemistry Lab and state "A GENERAL EMERGENCY HAS D BEEN DECLARED" and direct to implement the applicable TI-331, "Post

\

) Accident Sampling Procedure" and CI-900 series, "Analysis Procedures".

Plant Extension 2367 or 2368

5. Notify the RP Lab and state "A GENERAL EMERGENCY HAS BEEN D DECLARED" and direct to activate EPIP-14, "Radiological Control Procedure".

Plant Extension 7865 or 3104

6. Notify the "On-Call" NRC Resident and state "A GENERAL D EMERGENCY HAS BEEN DECLARED".

Plant Extension 2572 (Secretary) or from Weekly Duty List

)

PAGE 12 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIXC Page 1 of 2 MONITOR I RE-EVALUATE THE EVENT 1.0 IF ... significant changes in plant conditions such as other EAL conditions supporting the General Emergency or significant changes in radiological conditions, THEN ... COMPLETE Appendix 0 I Initials Time AND COMMUNICATE the "Follow-Up" information to:

On-Site Emergency Centers D Plant Personnel through PA announcements (if D applicable)

CECC (5-751-1614) D ODS (5-751-1700 or 5-751-2495) D State of Alabama D 24 Hours Primary: 9-1-205-280-2310 Backup: 9-1-800-843-0699 Backup: 9-1-334-324-0076 Nuclear Regulatory Commission (refer to Note in Step D 3.7 in body of procedure)

PAGE 13 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX C Page 2 of 2 MONITOR I RE-EVALUATE THE EVENT CAUTION Do not initiate Assembly / Accountability when:

1. A severe weather condition exists or is projected on-site, such as a Tornado.

2. An on-site security risk condition exists that may present a danger to site personnel during the Assembly / Accountability process as determined by SED/Nuclear Security.

3. Rapid Evacuation of the Protected Area (REPA) has been conducted.

2.0 IF ... conditions warrant the activation of Assembly / Accountability or /

Evacuation, Initials Time THEN ... ENTER, EPIP-8, and implement accordingly. Otherwise N/A this step.

/

3.0 IF ... conditions warrant termination of the emergency classification, Initials Time THEN ... ENTER, EPIP-16, "Termination and Recovery Procedure" and exit this procedure. Otherwise N/A this step.

PAGE 14 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX D Page 1 of 1 GENERAL EMERGENCY FOLLOW-UP INFORMATION FORM

1. 0 THIS IS A DRILL o THIS IS AN ACTUAL EVENT

2. There has been a General Emergency declared at Browns Ferry affecting:

o Unit 1 0 Unit 2 0 Unit 3 0 Common

3. Reactor Status: Unit 1 o Shutdown OAt Power 0 At Power ON/A Unit 2 o Shutdown OAt Power 0 At Power ON/A Unit 3 o Shutdown OAt Power 0 At Power ON/A

4. Additional EAL Designator(s):

5. Significant Changes in Plant Conditions:

6. Significant Changes in Radiological Conditions:

7. Off-site Protective Action Recommendations:

o Recommendation 1 0 Recommendation 2 o Recommendation 3 (CECC to provide detailed PAR Sector Recommendations)

8. On-site Protective Actions: Assembly / Accountability ONo Olnitiated OCompleted Site Evacuation ONo Olnitiated OCompleted

9. Meteorological conditions are: Wind Speed mph (Use 91 Meter Data on the Met Tower) Wind Direction from degrees

10. Please repeat the information you have received to ensure accuracy.

11. Fax to applicable contact after reporting follow-up information: CECC(5-751-1682), ODS (5-751-8620, or State of Alabama (9-1-205-280-2495).

Completed by: _ _ _ _ _ _ _ _ _ _ _ _ __ Date/Time / ----

PAGE 15 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX E Page 1 of6 TECHNICAL SUPPORT CENTER GENERAL EMERGENCY CLASSIFICATION INSTRUCTION 1.0 Notification of the CECC and/or State of Alabama of General Emergency Classification 1.1 CECC Notification D 1.1.1 COMPLETE in the following information:

GE Classification EAL Designator: _ _ _ __

GE Classification declared at time: - - - - -

Site Emergency Director: (name) _ _ _ _ _ __

AND CONTACT the CECC Director and communicate the - - /- -

information recorded in step 1.1 utilizing the J Initials Time CECC "Direct Ring-Down" telephone or at extension 5-751-1614.

1.1.2 IF ... the CECC Director was contacted Then ... the State of Alabama notification action is complete.

AND RE-ENTER this appendix at Step 2.0. Otherwise continue in this appendix.

PAGE 16 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX E Page 2 of6 TECHNICAL SUPPORT CENTER GENERAL EMERGENCY CLASSIFICATION INSTRUCTION 1.2 ODS Notification NOTE

The ODS should be contacted within 5 minutes of the emergency classification.

Completion of Appendix A for the General Emergency Classification includes the development of a Protective Action Recommendation (PAR) when the CECC is not operational. Utilize Appendix G, PAR flowchart when determining PAR, when required.

PAR must be made by SM/SED.

1.2.1 IF ... the CECC Director was not contacted, THEN ... COMPLETE Appendix A (Initial Notification Form). Utilize Appendix G, "Protective Action 0 Recommendation", flowchart as appropriate.

AND NOTIFY ... the ODS, at extension 5-751-1700 or 0 5-751-2495 AND REPORT ... to the ODS the information recorded - - /- -

on Appendix A. Initials Time AND FAX ... a copy of Appendix A to the ODS for 0 confirmation of information at 5-751-8620.

1.2.2 IF ... the ODS was contacted, 0 Then ... the State of Alabama notification action is complete.

AND RE-ENTER this appendix at Step 2.0. Otherwise continue in this appendix.

PAGE 17 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX E Page 3 of 6 TECHNICAL SUPPORT CENTER GENERAL EMERGENCY CLASSIFICATION INSTRUCTION 1.3 State of Alabama Notification NOTE

The State of Alabama shall be contacted within 15 minutes of the emergency classification.

Completion of Appendix A for the General Emergency Classification includes the development of a Protective Action Recommendation (PAR) when the CECC is not operational. Utilize Appendix G, PAR flowchart when determining PAR, when required.

PAR must be made by SM/SED.

1.3.1 IF ... the ODS cannot be contacted within 10 minutes THEN ... NOTIFY and REPORT the information recorded on Appendix A to the following:

Limestone County 9-232-2631 - - /- -

(after hours) 9-232-0111 Initials Time

Morgan County 9-1-256-351-4620 /

(after hours) 9-1-256-353-2515 Option 0 Initials Time

Lawrence County 9-1-256-974-7641 - - /- -

(after hours) 9-1-256-974-7911 Initials Time

Lauderdale County 9-1-256-760-6363 /

(after hours) 9-1-256-760-9117 Initials Time

State of Alabama at: - - /- -

Initials Time 24 Hours Primary: 9-1-205-280-2310 Backup: 9-1-800-843-0699 Backup: 9-1-334-324-0076 AND FAX ... a copy of Appendix A to the State of Alabama, D Office of Radiation Control for confirmation of information at 9-1-205-280-2495.

PAGE 18 OF 23 REVISION 0037

I BROWNS FERRY GENERAL EMERGENCY EPIP-5

)

APPENDIX E Page 4 af6 TECHNICAL SUPPORT CENTER GENERAL EMERGENCY CLASSIFICATION INSTRUCTION 2.0 CECC or ODS State of Alabama Notification Confirmation Receive a confirmation call from the CECC or the ODS verifying - - /- -

that the notification of the State of Alabama was completed. Do Initials Time this concurrently with the implementation of this procedure. (N/A this step if State was contacted directly) 3.0 Notification of Site Personnel CAUTION Ongoing or anticipated security events may present a danger to site personnel. Do not conduct the notification of site personnel PA message during an ongoing or anticipated security event. All pertinent site personnel PA messages will be conducted per AOI-100-8 for security events. .

) CONDUCT a Plant PA announcement similar to the following: D (Dial 687 to obtain the Plant PA)

Let me have your attention please.

This is (name) _ _ _--.

An General Emergency Classification has been declared.

We are currently implementing EPIP-5.

If you have not already done so, please report to your assigned emergency center at this time.

{

)

PAGE 19 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5

')

APPENDIX E Page 5 af6 TECHNICAL SUPPORT CENTER GENERAL EMERGENCY CLASSIFICATION INSTRUCTION 4.0 Assembly I Accountability CAUTION Do not initiate Assembly / Accountability when:

1. A severe weather condition exists or is projected on-site, such as a tornado.

2. An on-site security risk condition exists that may present a danger to site personnel during the Assembly / Accountability process as determined by SED/Nuclear Security.

3. Rapid Evacuation of the Protected Area (REPA) has been conducted.

4.1 IF ... Assembly / Accountability has not been conducted, D THEN ... IMPLEMENT EPIP-8, Appendix C concurrently with this procedure. This action may be delegated.

4.2 IF ... an order to evacuate non-emergency responders has not D been issued,

) THEN ... upon com'pletion of Assembly / Accountability, INITIATE the order to "Evacuate Non-Emergency Responders,"

through implementation of EPIP-8, Appendix F, concurrently with this procedure.

4.3 IF ... conditions exist that do not allow for an Assembly / D Accountability or Evacuation at this time, THEN ... CONTINUE to assess the situation, implementing EPIP-8 as applicable.

5.0 Dose Assessment EVALUATE ... the need for dose assessment. D IF ... dose assessment is needed, THEN ... CONTACT, if operational the Central Emergency Control Center (CECC) at 5-751-1614.

OR IF ... the CECC is not operational THEN ... REQUEST, the Radiological Protection Manager

) conduct a dose assessment utilizing EPIP-13.

PAGE 20 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIXE Page 6 of 6 TECHNICAL SUPPORT CENTER GENERAL EMERGENCY CLASSIFICATION INSTRUCTION 6.0 Notification of the Nuclear Regulatory Commission (NRC)

CAUTION Ensure that previous declared emergency classifications have been communicated to NRC, specifically in fast breaking events, where emergency classifications are rapidly changing.

NOTE

If possible, when making notifications to the NRC utilize the Emergency Notification System (ENS). Dial the first number listed on the sticker affixed to the ENS telephone, by dialing 9-1- "The Ten Digit Number Listed on the ENS Telephones". If the number is busy, then select in order, the alternate numbers until a connection is achieved. No access codes should be required.

This action may be delegated to the TSC NRC Coordinator.

NOTIFY ... the NRC immediately but no later than one hour after the D emergency has been declared.

IF ... REQUESTED by the NRC to maintain an open and continuous line of communications, THEN ... MAINTAIN an open and continuous line of communications as directed by NRC.

7.0 Review of Procedure D Review this procedure to ensure that all steps and actions have been completed and all place keeping blocks have been checked or denoted as instructed.

8.0 Monitor I Re-evaluate the Event Monitoring and reevaluation of plant events along with communicating significant changes should be performed continuously as a function of the emergency response.

Methods used to communicate significant changes are not formalized and may vary depending upon staffing levels as well as availability of personnel or equipment.

Appendix C provides a systematic approach to monitor/re-evaluate and communicate significant changes in plant conditions.

Utilize Appendix C to monitor/re-evaluate and communicate plant conditions and significant changes. Significant changes in plant conditions are at a minimum when other EAL conditions exist indicating the current emergency classification.

PAGE 21 OF 23 REVISION 0037

BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX F EPIP-5 Flow Illustration E for classification when the TSC is T

Caution: Appendix E Step 3.1 Appendix B H

Security Issues may Step 1.0 Activation of the Unit Operator present a danger to staffing ERa H ERa Notifications IF cbcc CECC 1 State of Alabama Notification Note:

Not Operational Appendix E Step 3.2 Notify ODS within 5 min. Appendix A Step 2.0 ODS 1 Slate of Utilize Ring-down phone UseApp G PAR Flowchart Alabama Notification -- Initial Notification Form CECC laDS State Notification Con fi rmation Appendix G Caution:

Step 3.3 PAR Appendix E Security Issues may ODSI Slale Flowchart Step 3.0 Notification Confirmation Notification of Site Personnel T

1 present a danger to site personnel notification

~

Caution: Appendix E Step 4.0

. ( / Do notC,.",

Initiate when:

Security Issues may Step3*R Notification of Site Assembly 1 Severe Weather present a danger to site personnel notification Personnel I Accountability Security Event Caution:

EPIP-B Do not Initiate when: Appendix E Assembly and Severe Weather Step 5.0 Accountability Security Event Dose Assessment N Note:

I F possible use the ENS when making this notification Note: Step 3.7 IF possible use the ENS Notification of when making this NRC Appendix E notification 10 CFR 50.72 Step 7.0 Review of Procedure Step 3.B Review of Procedure Appendix E Step 8.0 Monitor 1 Re-evaluate Step 3.9 Plant Events Monitor I Re-evaluate Plant Events Appendix 0 EPIP-16 Follow-Up Termination and Recovery Notification Form Procedure PAGE 22 OF 23 REVISION 0037

[ BROWNS FERRY GENERAL EMERGENCY EPIP-5 APPENDIX G Protective Action Recommendation Flowchart PROTECTIVE ACTION RECOMMENDATIONS Note 1: If conditions are unknown utilizing the flowchart, then answer NO.

Note 2: A short term release is defined as "a release that does not exceed a 15 minute duration".

GENERAL EMERGENCY DECLARED CONTI NUE ASSESSMENT

\ Modify protective actions based.onavailable plant and field monitoring information.

\ Locate and evaluate localized hot spots.

NO RECOMMENDATION 3 SHELTER 10 MILE EPZ RECOMMENDATION 2 EVACUATE 2 MILES RADIUS EVACUATE 2 MILESRADIUS AND5 MILES DOWNVVlND AND 10 MILES DOWNWIND AND AND SHELTER*REMAINDER OF SHELTER REMAINDER OF 10 MILE EPZ 10 MILE EPZ TABLE 1 I Protective Action Guides TYPE LIMIT Measured 3.9E-6 micro Ci/cc ofIodine l31 or 1 REM/hr External Dose Projected 1 REM TEDE or 5 REM Thyroid CDE LAST PAGE PAGE 23 OF 23 REVISION 0037

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2.0 REFERENCES

5.0 Leave

6.0 REFERENCES

6.0 REFERENCES

1.0 INTRODUCTION

2.0 REFERENCES

1.0 INTRODUCTION

2.0 REFERENCES

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