ML13330A922

Enclosure 2 - Volume 16 - Improved Technical Specifications Conversion, ITS Chapter 5.0 Administrative Controls, Revision 0

ML13330A922
Person / Time
Site:	Sequoyah
Issue date:	11/22/2013
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
Shared Package
ML13329A881	List: ML13329A717 ML13329A718 ML13329A790 ML13329A791 ML13329A792 ML13329A854 ML13329A855 ML13329A858 ML13329A859 ML13329A916 ML13330A922 ML13330A924 ML13330A925 ML13330A927 ML13330A928 ML13330A929 ML13330A930 ML13330A931 ML13330A953
References
NUREG-1431, Rev 4
Download: ML13330A922 (270)

Text

Enclosure 2, Volume 16, Rev. 0, Page 1 of 270 ENCLOSURE 2 VOLUME 16 SEQUOYAH NUCLEAR PLANT UNIT 1 AND UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Revision 0 Enclosure 2, Volume 16, Rev. 0, Page 1 of 270

Enclosure 2, Volume 16, Rev. 0, Page 2 of 270 LIST OF ATTACHMENTS

1. ITS Chapter 5.1 - Responsibility

2. ITS Chapter 5.2 - Organization

3. ITS Chapter 5.3 - Unit Staff Qualifications

4. ITS Chapter 5.4 - Procedures

5. ITS Chapter 5.5 - Programs and Manuals

6. ITS Chapter 5.6 - Reporting Requirements

7. ITS Chapter 5.7 - High Radiation Area Enclosure 2, Volume 16, Rev. 0, Page 2 of 270

, Volume 16, Rev. 0, Page 3 of 270 ATTACHMENT 1 ITS 5.1, RESPONSIBILITY , Volume 16, Rev. 0, Page 3 of 270

Enclosure 2, Volume 16, Rev. 0, Page 4 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Enclosure 2, Volume 16, Rev. 0, Page 4 of 270

Enclosure 2, Volume 16, Rev. 0, Page 5 of 270 ITS ITS 5.1 A01 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY LA01 5.1.1 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

INSERT 1 M01 5.1.2 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) shall be M02 responsible for the Control Room command function.

LA01 LA02 6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants.

6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS An onsite and an offsite organization shall be established for unit operation and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A).

b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety.

This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear See ITS safety is assured. 5.2

c. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant.

d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 FACILITY STAFF

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units.

b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

February 16, 2001 SEQUOYAH - UNIT 1 6-1 Amendment No. 32, 58, 74, 152, 178, 212, 233, 266 Page 1 of 4 Enclosure 2, Volume 16, Rev. 0, Page 5 of 270

Enclosure 2, Volume 16, Rev. 0, Page 6 of 270 5.1 M01 INSERT 1 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

Insert Page 6-1 Page 2 of 4 Enclosure 2, Volume 16, Rev. 0, Page 6 of 270

Enclosure 2, Volume 16, Rev. 0, Page 7 of 270 ITS A01 ITS 5.1 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY LA01 5.1.1 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

INSERT 1 M01 5.1.2 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) M02 shall be responsible for the Control Room command function. LA01 LA02 6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants.

6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS An onsite and an offsite organization shall be established for unit operation and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A).

b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

See ITS

c. The Plant Manager shall be responsible for overall unit safe operation, and shall have 5.2 control over those onsite resources necessary for safe operation and maintenance of the plant.

d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 FACILITY STAFF

a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units.

b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

February 16, 2001 SEQUOYAH - UNIT 2 6-1 Amendment No. 24, 50, 66, 142, 169, 202, 223, 257 Page 3 of 4 Enclosure 2, Volume 16, Rev. 0, Page 7 of 270

Enclosure 2, Volume 16, Rev. 0, Page 8 of 270 5.1 M01 INSERT 1 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

Insert Page 6-1 Page 4 of 4 Enclosure 2, Volume 16, Rev. 0, Page 8 of 270

Enclosure 2, Volume 16, Rev. 0, Page 9 of 270 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITIES ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.1.1 states that the Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. ITS 5.1.1 states that the plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. Additionally, it requires that the plant manager or his designee approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. This changes the CTS by adding an approval requirement for the plant manager or his designee.

The purpose of the ITS 5.1.1 requirement is to provide additional assurance that the plant manager has direct responsibility for overall unit operation. This change is acceptable because having the plant manager or his designee approve actions affecting nuclear safety is consistent with CTS 6.2.1.c (ITS 5.2.1.b) requirement that the plant manager shall be responsible for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant. This change is designated as more restrictive because it adds a requirement for the plant manager or his designee to the CTS.

M02 CTS 6.1.2 allows a designated individual to assume the responsibility for the control room command function when the Shift Manager is absent from the Control Room. ITS 5.1.2 provides the allowance for the designated individual to assume the responsibility for the control room command function, but provides additional requirements for the designated individual. In MODE 1, 2, 3, or 4, ITS 5.1.2 requires the designated individual to hold an active Senior Reactor Operator license. In MODE 5 or 6, ITS 5.1.2 requires the designated individual to hold an active Senior Reactor Operator license or Reactor Operator license.

This changes the CTS by adding qualification requirements for the designated individual that assumes the control room command function.

The purpose of the ITS 5.1.2 requirement is to ensure that the control room command function is maintained. This change is acceptable because the additional requirements ensure that the designated individual assuming the control room command functions meets the appropriate qualification Sequoyah Unit 1 and Unit 2 Page 1 of 3 Enclosure 2, Volume 16, Rev. 0, Page 9 of 270

Enclosure 2, Volume 16, Rev. 0, Page 10 of 270 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITIES requirements. This change is designated as more restrictive because it adds qualification requirements for the designated individual that assumes the control room command function to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.1.1 uses the title "Plant Manager" and CTS 6.1.2 uses the title "Shift Manager." ITS 5.1.1 uses the generic title "plant manager" and ITS 5.1.2 uses the generic title "shift manager." This changes the CTS by moving the specific organizational titles to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) and replacing them with generic titles.

The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific SQN organizational titles out of the Technical Specifications is consistent with the NRC letter from C Grimes to the Owners Group Technical Specification Committee Chairman, dated November 10, 1994.

The various requirements of the plant manager and shift manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the Nuclear Power Organization Topical Report (TVA-NPOD89-A) as described in ITS 5.2.1.a. Any changes to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) are made under 10 CFR 50.54(a)(3), which ensures that changes are properly evaluated.

This change is a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications.

LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.1.3 states that the Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants. ITS 5.1 does not contain this requirement. This changes the CTS by moving the requirements of the Chief Nuclear Officer to the Nuclear Power Organization Topical Report (TVA-NPOD89-A).

The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the UFSAR. Changes to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) are made under 10 CFR 50.54(a)(3), which ensures Sequoyah Unit 1 and Unit 2 Page 2 of 3 Enclosure 2, Volume 16, Rev. 0, Page 10 of 270

Enclosure 2, Volume 16, Rev. 0, Page 11 of 270 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITIES that changes are properly evaluated. This change is a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications LESS RESTRICTIVE CHANGES None Sequoyah Unit 1 and Unit 2 Page 3 of 3 Enclosure 2, Volume 16, Rev. 0, Page 11 of 270

Enclosure 2, Volume 16, Rev. 0, Page 12 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 12 of 270

Enclosure 2, Volume 16, Rev. 0, Page 13 of 270 CTS Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

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REVIEWER'S NOTES---------------------------------------

1. Titles for members of the unit staff shall be specified by use of an overall statement referencing an ANSI Standard acceptable to the NRC staff from which the titles were obtained, or an alternative title may be designated for this position. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special titles because of unique organizational structures.

2. The ANSI Standard shall be the same ANSI Standard referenced in Section 1 5.3, Unit Staff Qualifications. If alternative titles are used, all requirements of these Technical Specifications apply to the position with the alternative title as apply with the specified title. Unit staff titles shall be specified in the Final Safety Analysis Report or Quality Assurance Plan. Unit staff titles shall be maintained and revised using those procedures approved for modifying/revising the Final Safety Analysis Report or Quality Assurance Plan.

6.1.1 5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

DOC M01 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

shift manager 6.1.2 5.1.2 The [Shift Supervisor (SS)] shall be responsible for the control room command 2 shift manager function. During any absence of the [SS] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command shift manager function. During any absence of the [SS] from the control room while the unit is 2 in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.1-1 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 13 of 270

Enclosure 2, Volume 16, Rev. 0, Page 14 of 270 CTS Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

---

REVIEWER'S NOTES---------------------------------------

1. Titles for members of the unit staff shall be specified by use of an overall statement referencing an ANSI Standard acceptable to the NRC staff from which the titles were obtained, or an alternative title may be designated for this position. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special titles because of unique organizational structures.

2. The ANSI Standard shall be the same ANSI Standard referenced in Section 1 5.3, Unit Staff Qualifications. If alternative titles are used, all requirements of these Technical Specifications apply to the position with the alternative title as apply with the specified title. Unit staff titles shall be specified in the Final Safety Analysis Report or Quality Assurance Plan. Unit staff titles shall be maintained and revised using those procedures approved for modifying/revising the Final Safety Analysis Report or Quality Assurance Plan.

6.1.1 5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

DOC M01 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

shift manager 6.1.2 5.1.2 The [Shift Supervisor (SS)] shall be responsible for the control room command 2 shift manager function. During any absence of the [SS] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command shift manager function. During any absence of the [SS] from the control room while the unit is 2 in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.1-1 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 14 of 270

Enclosure 2, Volume 16, Rev. 0, Page 15 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.1, RESPONSIBILITY

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 15 of 270

Enclosure 2, Volume 16, Rev. 0, Page 16 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 16 of 270

Enclosure 2, Volume 16, Rev. 0, Page 17 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 17 of 270

, Volume 16, Rev. 0, Page 18 of 270 ATTACHMENT 2 ITS 5.2, ORGANIZATION , Volume 16, Rev. 0, Page 18 of 270

, Volume 16, Rev. 0, Page 19 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 19 of 270

Enclosure 2, Volume 16, Rev. 0, Page 20 of 270 A01 ITS ITS 5.2 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

See ITS 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) shall be 5.1 responsible for the Control Room command function.

6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants.

6.2 ORGANIZATION 5.2.1 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS

, respectively A01 An onsite and an offsite organization shall be established for unit operation and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

and established throughout A01 5.2.1.a a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate,

, A01 in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). , including the plant-specific titles of those personnel fulfilling the M01 responsibilities of the positions delineated in these Technical Specifications, A specified corporate officer 5.2.1.c b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. LA01 This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear LA01 safety is assured.

of the plant A01 5.2.1.b c. The Plant Manager shall be responsible for overall unit safe operation and shall have control activities over those onsite resources necessary for safe operation and maintenance of the plant.

A01 A01 or perform 5.2.1.d d. The individuals who train the operating staff and those who carry out health physics and quality these assurance functions may report to the appropriate onsite manager; however, they shall have individuals sufficient organizational freedom to ensure their independence from operating pressures.

Unit 5.2.2 6.2.2 FACILITY STAFF The unit staff organization shall include the following: A01 5.2.2.a a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units.

5.2.2.b b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

February 16, 2001 SEQUOYAH - UNIT 1 6-1 Amendment No. 32, 58, 74, 152, 178, 212, 233, 266 Page 1 of 10 Enclosure 2, Volume 16, Rev. 0, Page 20 of 270

Enclosure 2, Volume 16, Rev. 0, Page 21 of 270 A01 ITS ITS 5.2 ADMINISTRATIVE CONTROLS 5.2.2,c c. A Radiological Control technician# shall be onsite when fuel is in the reactor.

d. DELETED

e. DELETED 5.2.2.d f. The Operations Superintendent shall hold a Senior Reactor Operator license.

g. DELETED 5.2.2.e h. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

A01 position vacant not more than 5.2.2.c #The Radiological Control technician may be offsite for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

provide for A01 February 2, 2010 SEQUOYAH - UNIT 1 6-2 Amendment No. 32, 58, 74, 152, 156, 178, 227, 233, 240, 266, 281, 327 Page 2 of 10 Enclosure 2, Volume 16, Rev. 0, Page 21 of 270

Enclosure 2, Volume 16, Rev. 0, Page 22 of 270 A01 ITS ITS 5.2 Table 6.2-1 MINIMUM SHIFT CREW COMPOSITION WITH UNIT 2 IN MODE 5 OR 6 OR DE-FUELED THIS PAGE INTENTIONALLY DELETED February 16, 2001 SEQUOYAH - UNIT 1 6-3 Amendment No. 32, 58, 74, 178, 266 Page 3 of 10 Enclosure 2, Volume 16, Rev. 0, Page 22 of 270

Enclosure 2, Volume 16, Rev. 0, Page 23 of 270 A01 ITS ITS 5.2 TABLE 6.2-1 (Continued)

TABLE NOTATION THIS PAGE INTENTIONALLY DELETED February 16, 2001 SEQUOYAH - UNIT 1 6-4 Amendment No. 58, 74, 178, 266 Page 4 of 10 Enclosure 2, Volume 16, Rev. 0, Page 23 of 270

Enclosure 2, Volume 16, Rev. 0, Page 24 of 270 A01 ITS ITS 5.2 ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED) 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R See ITS 5.3 (May 1977).

6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m).

February 11, 2003 SEQUOYAH - UNIT 1 6-5 Amendment No. 12, 58, 74, 119, 152, 163, 178, 212, 233, 266, 281 Page 5 of 10 Enclosure 2, Volume 16, Rev. 0, Page 24 of 270

Enclosure 2, Volume 16, Rev. 0, Page 25 of 270 A01 ITS ITS 5.2 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

See ITS 6.1.2 The Shift Manager (or during his absence from the Control Room, a designated individual) 5.1 shall be responsible for the Control Room command function.

6.1.3 The Chief Nuclear Officer is responsible for the safe operation of all TVA Nuclear Power Plants.

6.2 ORGANIZATION 5.2.1 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS

, respectively A01 An onsite and an offsite organization shall be established for unit operation and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

and established throughout 5.2.1.a a. Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated,

, A01

, including the plant-specific titles of those as appropriate, in the form of organizational charts, functional descriptions of personnel fulfilling the departmental responsibilities and relationships, and job descriptions for key personnel M01 responsibilities of the positions delineated in positions, or in equivalent forms of documentation. These requirements shall be these Technical documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A).

Specifications, A specified corporate officer LA01 5.2.1.c b. The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable LA01 performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

of the plant A01 5.2.1.b c. The Plant Manager shall be responsible for overall unit safe operation, and shall have activities control over those onsite resources necessary for safe operation and maintenance of the plant.

A01 or perform A01 5.2.1.d d. The individuals who train the operating staff and those who carry out health physics and these individuals quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

Unit 5.2.2 6.2.2 FACILITY STAFF The unit staff organization shall include the following: A01 5.2.2.a a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each unit for which a reactor is operating in MODES 1, 2, 3, or 4. With both units shutdown or defueled, a total of three non-licensed operators are required for the two units.

5.2.2.b b. Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Sections 6.2.2.a and 6.2.2.h for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

February 16, 2001 SEQUOYAH - UNIT 2 6-1 Amendment No. 24, 50, 66, 142, 169, 202, 223, 257 Page 6 of 10 Enclosure 2, Volume 16, Rev. 0, Page 25 of 270

Enclosure 2, Volume 16, Rev. 0, Page 26 of 270 A01 ITS ITS 5.2 ADMINISTRATIVE CONTROLS 5.2.2,c c. A Radiological Control technician# shall be onsite when fuel is in the reactor.

d. DELETED

e. DELETED 5.2.2.d f. The Operations Superintendent shall hold a Senior Reactor Operator license.

g. DELETED 5.2.2.e

h. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

A01 position vacant not more than 5.2.2.c # The Radiological Control technician may be offsite for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in 01order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

provide for A01 February 2, 2010 SEQUOYAH - UNIT 2 6-2 Amendment No. 50, 66, 142, 145 169, 218, 223, 230, 257, 272, 320 Page 7 of 10 Enclosure 2, Volume 16, Rev. 0, Page 26 of 270

Enclosure 2, Volume 16, Rev. 0, Page 27 of 270 A01 ITS ITS 5.2 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION WITH UNIT 1 IN MODE 5 OR 6 OR DE-FUELED THIS PAGE INTENTIONALLY DELETED February 16, 2001 SEQUOYAH - UNIT 2 6-3 Amendment No. 50, 66, 169, 257 Page 8 of 10 Enclosure 2, Volume 16, Rev. 0, Page 27 of 270

Enclosure 2, Volume 16, Rev. 0, Page 28 of 270 A01 ITS ITS 5.2 TABLE 6.2-1 (Continued)

TABLE NOTATION THIS PAGE INTENTIONALLY DELETED February 16, 2001 SEQUOYAH - UNIT 2 6-4 Amendment Nos. 50, 66, 169, 257 Page 9 of 10 Enclosure 2, Volume 16, Rev. 0, Page 28 of 270

Enclosure 2, Volume 16, Rev. 0, Page 29 of 270 A01 ITS ITS 5.2 ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED) 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R See ITS 5.3 (May 1977).

6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m).

6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 RADIOLOGICAL ASSESSMENT REVIEW COMMITTEE (RARC) (DELETED February 11, 2003 SEQUOYAH - UNIT 2 6-5 Amendment No. 34, 50, 66, 108, 142, 153, 169, 189, 202, 223, 257, 272 Page 10 of 10 Enclosure 2, Volume 16, Rev. 0, Page 29 of 270

Enclosure 2, Volume 16, Rev. 0, Page 30 of 270 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.2.1.a regarding documentation and updating of the relationships between operating organization position, requires the organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions to be documented in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). ITS 5.2.1.a states "These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the Nuclear Power Organization Topical Report (TVA NPOD89-A). This changes the CTS by requiring that the specific SQN organizational titles be specified in the Nuclear Power Organization Topical Report (TVA-NPOD89-A).

This change is acceptable because specifying the relationship of the specific SQN organizational titles to the generic titles used in the Technical Specifications and organizational positions, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions used in the Technical Specifications and industry standards in the Nuclear Power Organization Topical Report (TVA-NPOD89-A) continues to ensure that organizational positions and associated responsibilities will be maintained. This change adds the requirements to the Technical Specifications.

This change is designated as more restrictive because it requires additional information be maintained in the Nuclear Power Organization Topical Report (TVA-NPOD89-A).

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.2.1.b uses the title "Chief Nuclear Officer," and CTS 6.2.1.c uses the title "Plant Manager." ITS 5.2.1.b uses the generic title Sequoyah Unit 1 and Unit 2 Page 1 of 2 Enclosure 2, Volume 16, Rev. 0, Page 30 of 270

Enclosure 2, Volume 16, Rev. 0, Page 31 of 270 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION "plant manager," and ITS 5.2.1.c uses the generic title "A specified corporate officer." This changes the CTS by moving the specific SQN organizational titles to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) and replacing them with generic titles.

The removal of these details, which are related to meeting Technical Specification requirements from the Technical Specifications, is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific SQN organizational titles out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairman, dated November 10, 1994.

The various requirements of the plant manager and the specified corporate officer are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the Nuclear Power Organization Topical Report (TVA-NPOD89-A). Any changes to the Nuclear Power Organization Topical Report (TVA-NPOD89-A) will be made under 10 CFR 50.54(a)(3) which will ensure the changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to meeting Technical Specification requirements is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Sequoyah Unit 1 and Unit 2 Page 2 of 2 Enclosure 2, Volume 16, Rev. 0, Page 31 of 270

Enclosure 2, Volume 16, Rev. 0, Page 32 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 32 of 270

Enclosure 2, Volume 16, Rev. 0, Page 33 of 270 CTS Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 6.2.1 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

6.2.1.a a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the [FSAR/QA Plan], 1 2 Nuclear Power Organization Topical Report (TVA-NPOD89-A);

6.2.1.c b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant,  ; 2 6.2.1.b c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety, and

2 6.2.1.d d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 5.2.2 Unit Staff The unit staff organization shall include the following:

6.2.2.a a. A non-licensed operator shall be assigned to each reactor containing fuel unit and an additional non-licensed operator shall be assigned for each control 4 room from which a reactor is operating in MODES 1, 2, 3, or 4. 2

---

REVIEWER'S NOTE----------------------------------------

2 Two unit sites with both units shutdown or defueled require a total of three non- 3 licensed operators for the two units. ,

are required SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.2-1 Rev. 4.0 4 Enclosure 2, Volume 16, Rev. 0, Page 33 of 270

Enclosure 2, Volume 16, Rev. 0, Page 34 of 270 CTS Organization 5.2 5.2 Organization 6.2.2 5.2.2 Unit Staff (continued) 6.2.2.b b. Shift crew composition may be less than the minimum requirement of 5

10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e for a period of time not to Specifications exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

6.2.2.c c. A radiation protection technician shall be on site when fuel is in the reactor.

The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

superintendent 6.2.2.f d. The operations manager or assistant operations manager shall hold an 4 SRO license.

6.2.2.h e. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.2-2 Rev. 4.0 4 Enclosure 2, Volume 16, Rev. 0, Page 34 of 270

Enclosure 2, Volume 16, Rev. 0, Page 35 of 270 CTS Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 6.2.1 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

6.2.1.a a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the [FSAR/QA Plan], 1 2 Nuclear Power Organization Topical Report (TVA-NPOD89-A);

6.2.1.c b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant,  ; 2 6.2.1.b c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety, and

2 6.2.1.d d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 5.2.2 Unit Staff The unit staff organization shall include the following:

6.2.2.a a. A non-licensed operator shall be assigned to each reactor containing fuel unit and an additional non-licensed operator shall be assigned for each control 4 room from which a reactor is operating in MODES 1, 2, 3, or 4. 2

---

REVIEWER'S NOTE----------------------------------------

2 Two unit sites with both units shutdown or defueled require a total of three non- 3 licensed operators for the two units. ,

are required SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.2-1 Rev. 4.0 4 Enclosure 2, Volume 16, Rev. 0, Page 35 of 270

Enclosure 2, Volume 16, Rev. 0, Page 36 of 270 CTS Organization 5.2 5.2 Organization 6.2.2 5.2.2 Unit Staff (continued) 6.2.2.b b. Shift crew composition may be less than the minimum requirement of 5

10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.e for a period of time not to Specifications exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

6.2.2.c c. A radiation protection technician shall be on site when fuel is in the reactor.

The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

superintendent 6.2.2.f d. The operations manager or assistant operations manager shall hold an 4 SRO license.

6.2.2.h e. An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.2-2 Rev. 4.0 4 Enclosure 2, Volume 16, Rev. 0, Page 36 of 270

Enclosure 2, Volume 16, Rev. 0, Page 37 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION

1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

3. ISTS 5.2.1.a contains a Reviewer's Note that allows two units with both units shutdown or defueled to have a total of three non-licensed operators for the two units. This Note applies to Sequoyah Nuclear Plant (SQN) since it is a two unit plant.

Additionally, CTS 6.2.2.a contains this same statement. Therefore, the Reviewer's Note has been deleted and the information contained in the note has been added to ITS 5.2.1.a.

4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

5. Grammatical/editorial change made for enhanced clarity.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 37 of 270

Enclosure 2, Volume 16, Rev. 0, Page 38 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 38 of 270

Enclosure 2, Volume 16, Rev. 0, Page 39 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 39 of 270

, Volume 16, Rev. 0, Page 40 of 270 ATTACHMENT 3 ITS 5.3, UNIT STAFF QUALIFICATIONS , Volume 16, Rev. 0, Page 40 of 270

, Volume 16, Rev. 0, Page 41 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 41 of 270

Enclosure 2, Volume 16, Rev. 0, Page 42 of 270 A01 ITS ITS 5.3 ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED)

See ITS 5.2 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 5.3 6.3 FACILITY STAFF QUALIFICATIONS 5.3.1 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977).

5.3.2 6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m).

February 11, 2003 SEQUOYAH - UNIT 1 6-5 Amendment No. 12, 58, 74, 119, 152, 163, 178, 212, 233, 266, 281 Page 1 of 4 Enclosure 2, Volume 16, Rev. 0, Page 42 of 270

Enclosure 2, Volume 16, Rev. 0, Page 43 of 270 A01 ITS ITS 5.3 ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 THIS SPECIFICATION IS DELETED 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 6.8 PROCEDURES & PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities See ITS referenced below: 5.4

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

May 24, 2002 SEQUOYAH - UNIT 1 6-6 Amendment No. 36, 42, 58, 74, 152, 163, 178, 198, 212, 233, 276 Page 2 of 4 Enclosure 2, Volume 16, Rev. 0, Page 43 of 270

Enclosure 2, Volume 16, Rev. 0, Page 44 of 270 A01 ITS ITS 5.3 ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING (ISE) (DELETED)

See ITS 5.2 6.2.4 SHIFT TECHNICAL ADVISOR (STA) (DELETED) 5.3 6.3 FACILITY STAFF QUALIFICATIONS 5.3.1 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications referenced for comparable positions in Regulatory Guide 1.8, Revision 2 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977).

5.3.2 6.3.2 For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of TS 6.3.1, perform the functions described in 10 CFR 50.54(m).

6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED See ITS 5.2 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED) 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 RADIOLOGICAL ASSESSMENT REVIEW COMMITTEE (RARC) (DELETED February 11, 2003 SEQUOYAH - UNIT 2 6-5 Amendment No. 34, 50, 66, 108, 142, 153, 169, 189, 202, 223, 257, 272 Page 3 of 4 Enclosure 2, Volume 16, Rev. 0, Page 44 of 270

Enclosure 2, Volume 16, Rev. 0, Page 45 of 270 A01 ITS ITS 5.3 ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

See ITS 5.4

c. Surveillance and test activities of safety related equipment.

d. DELETED

e. DELETED

f. Fire Protection Program implementation.

g. DELETED May 24, 2002 SEQUOYAH - UNIT 2 6-6 Amendment No. 28, 50, 66, 142, 223, 267 Page 4 of 4 Enclosure 2, Volume 16, Rev. 0, Page 45 of 270

Enclosure 2, Volume 16, Rev. 0, Page 46 of 270 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 46 of 270

Enclosure 2, Volume 16, Rev. 0, Page 47 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 47 of 270

Enclosure 2, Volume 16, Rev. 0, Page 48 of 270 CTS Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 6.3 5.3 Unit Staff Qualifications

---

REVIEWER'S NOTE-------------------------------------------------

Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position qualifications. Generally, the first method is preferable; however, 1 the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.

6.3.1 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of

[Regulatory Guide 1.8, Revision 2, 1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff]. [The staff not covered by Regulatory 2

Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff]. INSERT 1 6.3.2 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m).

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.3-1 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 48 of 270

Enclosure 2, Volume 16, Rev. 0, Page 49 of 270 ITS 5.3 2

INSERT 1 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977).

Insert Page 5.3-1 Enclosure 2, Volume 16, Rev. 0, Page 49 of 270

Enclosure 2, Volume 16, Rev. 0, Page 50 of 270 CTS Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 6.3 5.3 Unit Staff Qualifications

---

REVIEWER'S NOTE-------------------------------------------------

Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by specifying individual position qualifications. Generally, the first method is preferable; however, 1 the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.

6.3.1 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of

[Regulatory Guide 1.8, Revision 2, 1987, or more recent revisions, or ANSI Standard acceptable to the NRC staff]. [The staff not covered by Regulatory 2

Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff]. INSERT 1 6.3.2 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m).

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.3-1 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 50 of 270

Enclosure 2, Volume 16, Rev. 0, Page 51 of 270 ITS 5.3 2

INSERT 1 (April 1987) for all new personnel qualifying on positions identified in Regulatory Position C.1 after January 1, 1990. Personnel qualified on these positions prior to this date will still meet the requirements of Regulatory Guide 1.8, Revision 1-R (May 1977).

Insert Page 5.3-1 Enclosure 2, Volume 16, Rev. 0, Page 51 of 270

Enclosure 2, Volume 16, Rev. 0, Page 52 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 52 of 270

Enclosure 2, Volume 16, Rev. 0, Page 53 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 53 of 270

Enclosure 2, Volume 16, Rev. 0, Page 54 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 54 of 270

, Volume 16, Rev. 0, Page 55 of 270 ATTACHMENT 4 ITS 5.4, PROCEDURES , Volume 16, Rev. 0, Page 55 of 270

, Volume 16, Rev. 0, Page 56 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 56 of 270

Enclosure 2, Volume 16, Rev. 0, Page 57 of 270 A01 ITS ITS 5.4 ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 DELETED 6.5 REVIEW AND AUDIT 6.5.0 DELETED 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC) (DELETED)

See ITS 5.3 6.5.1A TECHNICAL REVIEW AND CONTROL (DELETED) 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) (DELETED) 6.5.3 THIS SPECIFICATION IS DELETED 6.6 REPORTABLE EVENT ACTION (DELETED) 6.7 SAFETY LIMIT VIOLATION (DELETED) 5.4 6.8 PROCEDURES & PROGRAMS 5.4.1 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

5.4.1.a a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

Add proposed Specification 5.4.1.b M01 May 24, 2002 SEQUOYAH - UNIT 1 6-6 Amendment No. 36, 42, 58, 74, 152, 163, 178, 198, 212, 233, 276 Page 1 of 4 Enclosure 2, Volume 16, Rev. 0, Page 57 of 270

Enclosure 2, Volume 16, Rev. 0, Page 58 of 270 A01 ITS ITS 5.4 ADMINISTRATIVE CONTROLS A02

b. Refueling operations.

A02

c. Surveillance and test activities of safety-related equipment.

d. DELETED

e. DELETED 5.4.1.d f. Fire Protection Program implementation.

g. DELETED 5.4.1.c h. Quality Assurance Program for effluent and environmental monitoring, using the guidance LA01 contained in Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975.

A03

i. OFFSITE DOSE CALCULATION MANUAL implementation.

Add proposed Specification 5.4.1.e M02 6.8.2 DELETED 6.8.3 DELETED 6.8.4 The following programs shall be established, implemented, and maintained.

a. Primary Coolant Sources Outside Containment See ITS 5.5 A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The February 11, 2003 SEQUOYAH - UNIT 1 6-7 Amendment No. 42, 58, 74, 148, 178, 233, 281 Page 2 of 4 Enclosure 2, Volume 16, Rev. 0, Page 58 of 270

Enclosure 2, Volume 16, Rev. 0, Page 59 of 270 A01 ITS ITS 5.4 ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION (DELETED)

See ITS 5.3 6.7 SAFETY LIMIT VIOLATION (DELETED) 5.4 6.8 PROCEDURES AND PROGRAMS 5.4.1 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

5.4.1.a a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

Add proposed Specification 5.4.1.b M01

b. Refueling operations.

A02

c. Surveillance and test activities of safety related equipment.

A02

d. DELETED

e. DELETED 5.4.1.d f. Fire Protection Program implementation.

g. DELETED May 24, 2002 SEQUOYAH - UNIT 2 6-6 Amendment No. 28, 50, 66, 142, 223, 267 Page 3 of 4 Enclosure 2, Volume 16, Rev. 0, Page 59 of 270

Enclosure 2, Volume 16, Rev. 0, Page 60 of 270 A01 ITS ITS 5.4 ADMINISTRATIVE CONTROLS 5.4.1.c h. Quality Assurance Program for effluent and environmental monitoring, using the guidance LA01 contained in Regulatory Guide 4.15, December 1977 or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975.

A03

i. OFFSITE DOSE CALCULATION MANUAL implementation.

Add proposed Specification 5.4.1.e M02 6.8.2 DELETED 6.8.3 DELETED 6.8.4 The following programs shall be established, implemented, and maintained.

a. Primary Coolant Sources Outside Containment A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the safety injection system, residual heat removal system, chemical and volume control system, containment spray system, and RCS sampling system. The program shall include the following:

(i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at lease once per 18 months.

The provisions of SR 4.0.2 are applicable

b. In-Plant Radiation Monitoring (DELETED)

c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube See ITS degradation. This program shall include: 5.5 (i) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for off-control point chemistry conditions, (vi) Procedures identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Deleted February 11, 2003 SEQUOYAH - UNIT 2 6-7 Amendment No. 34, 50, 66, 134, 149, 169, 223, 272 Page 4 of 4 Enclosure 2, Volume 16, Rev. 0, Page 60 of 270

Enclosure 2, Volume 16, Rev. 0, Page 61 of 270 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.8.1.b requires written procedures be established, implemented and maintained covering refueling operations. CTS 6.8.1.c requires written procedures be established, implemented and maintained covering surveillance and test activities of safety-related equipment. ITS 5.4.1 requires written procedures shall be established, implemented, and maintained to the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. This changes the CTS by removing the specific wording of CTS 6.8.1.b and CTS 6.8.1.c.

This change is acceptable because the recommendations of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 already require procedures for refueling operations and surveillance tests for safety related activities. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 6.8.1.i requires written procedures be established, implemented and maintained for the OFFSITE DOSE CALCULATION MANUAL (ODCM) implementation. ITS 5.4.1 requires procedures for various activities, but does not specifically list the ODCM. This changes the CTS by removing the specific requirement for written procedures to implement the ODCM.

This change is acceptable because implementing procedures for the ODCM are required by ITS 5.4.1.e. ITS 5.4.1.e (as described in DOC M02) requires that written procedures be established, implemented and maintained for all programs and manuals listed in ITS 5.5. ITS 5.5 includes the ODCM. Therefore, it is not necessary to specifically identify each program in ITS 5.4.1. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 ITS 5.4.1.b requires that written procedures shall be established, implemented, and maintained for the emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. The CTS does not include this requirement. This changes the CTS by adopting a new requirement for emergency operating procedures.

Sequoyah Unit 1 and Unit 2 Page 1 of 3 Enclosure 2, Volume 16, Rev. 0, Page 61 of 270

Enclosure 2, Volume 16, Rev. 0, Page 62 of 270 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES The purpose of ITS 5.4.1.b is to ensure that written procedures are established, implemented, and maintained covering the emergency operating procedures to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. This change is acceptable because it is consistent with an existing requirement to comply with NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33, for emergency operating procedures. This change is designated as more restrictive because it imposes a new requirement for procedures within the Technical Specifications.

M02 ITS 5.4.1.e requires that written procedures shall be established, implemented, and maintained for all programs specified in Specification 5.5. The CTS does not include this requirement for any program except the OFFSITE DOSE CALCULATION MANUAL. This changes the CTS by adopting a new requirement for procedures to address all programs described in ITS 5.5.

The purpose of ITS 5.4.1.e is to ensure that written procedures are established, implemented, and maintained covering all programs specified in ITS 5.5. This change is acceptable because it requires written procedures, including proper procedure control to address programs required by ITS 5.5. This change is designated as more restrictive because it imposes new requirements for procedures within the Technical Specifications.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, or ISI Program) CTS 6.8.1.h requires written procedures be established, implemented and maintained covering the Quality Assurance Program for effluent and environmental monitoring, "using the guidance in Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Revision 1, 1974, and Regulatory Guide 4.1, Revision 1, April 1975." ITS 5.4.1.c does not include the Regulatory Guide references. This changes the CTS by moving the references to the Regulatory Guides to the Nuclear Quality Assurance Program (NQAP).

The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for written procedures covering quality assurance for effluent and environmental monitoring. Also, this change is acceptable because these types of procedural details will be adequately controlled in the NQAP. Any changes to the NQAP are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because references for meeting Technical Specification requirements are being removed from the Technical Specifications.

Sequoyah Unit 1 and Unit 2 Page 2 of 3 Enclosure 2, Volume 16, Rev. 0, Page 62 of 270

Enclosure 2, Volume 16, Rev. 0, Page 63 of 270 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES LESS RESTRICTIVE CHANGES None Sequoyah Unit 1 and Unit 2 Page 3 of 3 Enclosure 2, Volume 16, Rev. 0, Page 63 of 270

Enclosure 2, Volume 16, Rev. 0, Page 64 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 64 of 270

Enclosure 2, Volume 16, Rev. 0, Page 65 of 270 CTS Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 6.8 5.4 Procedures 6.8.1 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

6.8.1.a a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, 1 DOC M01 b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as 2 stated in [Generic Letter 82-33], 3 1 program 4 6.8.1.h c. Quality assurance for effluent and environmental monitoring, 1 6.8.1.f d. Fire Protection Program implementation, and 1 6.8.1.i e. All programs specified in Specification 5.5.

DOC M02 SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.4-1 Rev. 4.0 4 Enclosure 2, Volume 16, Rev. 0, Page 65 of 270

Enclosure 2, Volume 16, Rev. 0, Page 66 of 270 CTS Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 6.8 5.4 Procedures 6.8.1 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

6.8.1.a a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, 1 DOC M01 b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as 2 stated in [Generic Letter 82-33], 3 1 program 4 6.8.1.h c. Quality assurance for effluent and environmental monitoring, 1 6.8.1.f d. Fire Protection Program implementation, and 1 6.8.1.i e. All programs specified in Specification 5.5.

DOC M02 SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.4-1 Rev. 4.0 4 Enclosure 2, Volume 16, Rev. 0, Page 66 of 270

Enclosure 2, Volume 16, Rev. 0, Page 67 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES

1. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

2. Typographical/grammatical error corrected.

3. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

4. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 67 of 270

Enclosure 2, Volume 16, Rev. 0, Page 68 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 68 of 270

Enclosure 2, Volume 16, Rev. 0, Page 69 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 69 of 270

, Volume 16, Rev. 0, Page 70 of 270 ATTACHMENT 5 ITS 5.5, PROGRAMS AND MANUALS , Volume 16, Rev. 0, Page 70 of 270

, Volume 16, Rev. 0, Page 71 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 71 of 270

Enclosure 2, Volume 16, Rev. 0, Page 72 of 270 A01 ITS ITS 5.5

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or See ITS

c. Reactor coolant system leakage through a steam generator to the secondary system 1.0 (primary to secondary leakage).

MEMBER(S) OF THE PUBLIC 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and 5.5.1.a parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring 5.5.1.b Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE See ITS 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of 1.0 core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

February 23, 2006 SEQUOYAH - UNIT 1 1-4 Amendment No. 12, 71, 148, 155, 169, 174, 178, 281, 306 Page 1 of 64 Enclosure 2, Volume 16, Rev. 0, Page 72 of 270

Enclosure 2, Volume 16, Rev. 0, Page 73 of 270 A01 ITS ITS 5.5 SURVEILLANCE REQUIREMENTS (Continued) 4.0.3 (Continued)

If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered.

See ITS 4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made 3.0 when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

5.5.6 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be as follows:

pumps and valves LA01 Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

LA01

a. Provisions that inservice testing of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

b. The provisions of SR 4.0.2 are applicable to the frequencies for performing inservice inspection activities; 5.5.5 c. Inspection of each reactor coolant pump flywheel per the recommendation of Regulation Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975 or in lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the removed flywheels may be conducted at 20-year intervals (the provisions of SR 4.0.2 are not applicable); and A02

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirement of any TS. LA01 5.5.6 Inservice Testing Program pumps and valves LA01 5.5.6 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

5.5.6.a a. Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a; October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-3 Amendment No. 78, 162, 202, 208, 274, 280, 293, 301, 308 Page 2 of 64 Enclosure 2, Volume 16, Rev. 0, Page 73 of 270

Enclosure 2, Volume 16, Rev. 0, Page 74 of 270 A01 ITS ITS 5.5 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 5.5.6 4.0.5 (Continued) 5.5.6.a b. Testing Frequencies applicable to the ASME OM Code and applicable Addenda as follows:

ASME OM Code and applicable Addenda Required frequencies for terminology for inservice performing inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days 5.5.6.b c. The provisions of SR 4.0.2 are applicable to the above required Frequencies and other normal and accelerated frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities; 5.5.6.c d. The provisions of SR 4.0.3 are applicable to inservice testing and activities; and 5.5.6.d e. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

October 4, 2006 SEQUOYAH - UNIT 1 3/4 0-4 Amendment No. 78, 162, 202, 208, 274, 280, 293, 308 Page 3 of 64 Enclosure 2, Volume 16, Rev. 0, Page 74 of 270

Enclosure 2, Volume 16, Rev. 0, Page 75 of 270 A01 ITS ITS 5.5 CONTAINMENT SYSTEMS EMERGENCY GAS TREATMENT SYSTEM - EGTS - CLEANUP SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.8 Two independent emergency gas treatment system cleanup subsystems (EGTS) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. See ITS 3.6.10 ACTION:

With one EGTS cleanup subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.8 Each EGTS cleanup subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control See ITS room, flow through the HEPA filters and charcoal adsorbers and verifying that the system 3.6.10 operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

Add proposed ITS 5.5.9 generic program statement A03

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or 5.5.9 charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria 5.5.9.b and uses the test procedures of Regulatory Position C.5.a., C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm + 10%.

LA02

2. Verifying within 31 days after removal that a laboratory analysis of a 5.5.9.c representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.

5.5.9.a 3. Verifying a system flow rate of 4000 cfm + 10% during system operation when 5.5.9.b tested in accordance with ANSI N510-1975.

5.5.9.c November 2, 2000 SEQUOYAH - UNIT 1 3/4 6-13 Amendment No. 263 Page 4 of 64 Enclosure 2, Volume 16, Rev. 0, Page 75 of 270

Enclosure 2, Volume 16, Rev. 0, Page 76 of 270 A01 ITS ITS 5.5 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

LA02 5.5.9 c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in 5.5.9.c accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.

5.5.9.d d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm +/- 10%.

2. Verifying that the filter train starts on a Phase A containment isolation Test Signal. See ITS 3.6.10

3. Verify the operation of the filter cooling bypass valves.

4. Verifying that each system produces a negative pressure of greater than or equal See ITS to 0.5 inches W. G. in the annulus within 1 minute after a start signal. 3.6.7 5.5.9

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are 5.5.9.a tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

5.5.9

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated 5.5.9.b hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 November 2, 2000 SEQUOYAH - UNIT 1 3/4 6-14 Amendment No. 21, 88, 103, 263 Page 5 of 64 Enclosure 2, Volume 16, Rev. 0, Page 76 of 270

Enclosure 2, Volume 16, Rev. 0, Page 77 of 270 A01 ITS ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Add proposed ITS 5.5.9 generic program statement A03 5.5.9 c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria 5.5.9.b and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm +/- 10%.

LA02 5.5.9.c 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9.a 5.5.9.b 3. Verifying a system flow rate of 4000 cfm +/- 10% during system operation when tested 5.5.9.d in accordance with ANSI N510-1975.

LA02 5.5.9 d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance 5.5.9.c with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9 e. At least once per 18 months by:

5.5.9.d 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the system at a flow rate of 4000 cfm +/- 10%.

2. Verifying that on a safety injection signal or a high radiation signal from the air intake stream, the system automatically diverts its inlet flow through the HEPA filters and See ITS 3.7.10 charcoal adsorber banks.

5.5.9

f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-5.5.9.a place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

5.5.9

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

h. Perform required CRE unfiltered air inleakage testing in accordance with the Control Room See ITS 3.7.10 Envelope Habitability Program.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 October 28, 2008 SEQUOYAH - UNIT 1 3/4 7-18 Amendment No. 12, 68, 88, 263, 321 Page 6 of 64 Enclosure 2, Volume 16, Rev. 0, Page 77 of 270

Enclosure 2, Volume 16, Rev. 0, Page 78 of 270 A01 ITS ITS 5.5 PLANT SYSTEMS 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent auxiliary building gas treatment filter trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. See ITS 3.7.12 SURVEILLANCE REQUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the system operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

Add proposed ITS 5.5.9 generic program statement A03 5.5.9 b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria 5.5.9.b and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm +/- 10%.

LA02

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory 5.5.9.c Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9.a 3. Verifying a system flow rate of 9000 cfm +/- 10% during system operation when tested 5.5.9.b in accordance with ANSI N510-1975.

5.5.9.d November 2, 2000 SEQUOYAH - UNIT 1 3/4 7-19 Amendment No. 12, 263 Page 7 of 64 Enclosure 2, Volume 16, Rev. 0, Page 78 of 270

Enclosure 2, Volume 16, Rev. 0, Page 79 of 270 A01 ITS ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

LA02 5.5.9

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance 5.5.9.c with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9 d. At least once per 18 months by:

5.5.9.d 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter trains start on a Containment Phase A Isolation test signal.

3. Verifying that the system maintains the spent fuel storage area and the ESF pump See ITS 3.7.12 rooms at a pressure equal to or more negative than minus 1/4 inch water gage relative the outside atmosphere while maintaining a total system flow of 9000 cfm

+/- 10%.

5.5.9.e 4. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N510-1975.

5.5.9

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-5.5.9.a place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

5.5.9

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 August 18, 2005 SEQUOYAH - UNIT 1 3/4 7-20 Amendment Nos. 12, 88, 103, 122, 263, 303 Page 8 of 64 Enclosure 2, Volume 16, Rev. 0, Page 79 of 270

Enclosure 2, Volume 16, Rev. 0, Page 80 of 270 A01 ITS ITS 5.5 TABLE 4.8.1a DIESEL GENERATOR BATTERY SURVEILLANCE REQUIREMENTS (1) (2)

CATEGORY A CATEGORY B (3)

Parameter Allowable Limits for each Limits for each value for each designated pilot cell connected cell connected cell See ITS 3.8.6

>Minimum level >Minimum level Above top of Electrolyte Level indication mark, indication mark, and plates, and not and 1/4 above 1/4 above overflowing maximum level maximum level indication mark indication mark 5.5.15.b.2 Float Voltage 2.13 volts 2.13 volts(C) > 2.07 volts Not more than

.020 below the average of all 1.190 connected cells See ITS (a) (b) Average of all Average of all 3.8.6 Specific Gravity 1.195 connected cells connected cells

> 1.200 > 1.190(b)

(a) Corrected for electrolyte temperature and level.

(b) Or battery charging current is less than 2 amps.

(c) Corrected for average electrolyte temperature. LA05 (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. See ITS (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered 3.8.6 OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.

(3) Any Category B parameter not within its allowable value indicates an inoperable battery.

March 25, 1982 SEQUOYAH - UNIT 1 3/4 8-7a Amendment No. 12 Page 9 of 64 Enclosure 2, Volume 16, Rev. 0, Page 80 of 270

Enclosure 2, Volume 16, Rev. 0, Page 81 of 270 A01 ITS ITS 5.5 TABLE 4.8.2 BATTERY SURVEILLANCE REQUIREMENTS (1) (2)

CATEGORY A CATEGORY B Limits for each Limits for each (3)

Parameter Allowable value for designated pilot cell connected cell each connected cell See ITS

>Minimum level >Minimum level Above top of plates, and 3.8.6 Electrolyte Level indication mark, indication mark, and not overflowing and 1/4 above 1/4 above maximum maximum level level indication mark indication mark 5.5.15.b.2 2.13 volts 2.13 volts (c) > 2.07 volts Float Voltage Not more than .020 below the average of all 1.195 connected cells See ITS 3.8.6 (a) (b) Average of all connected Average of all connected Specific Gravity 1.200 cells > 1.205 cells > 1.195(b)

(a) Corrected for electrolyte temperature and level.

(b) Or battery charging current is less than 2 amps.

(c) Corrected for average electrolyte temperature. LA05 (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. See ITS 3.8.6 (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.

(3) Any Category B parameter not within its allowable value indicates an inoperable battery.

March 25, 1982 SEQUOYAH - UNIT 1 3/4 8-13a Amendment No. 12 Page 10 of 64 Enclosure 2, Volume 16, Rev. 0, Page 81 of 270

Enclosure 2, Volume 16, Rev. 0, Page 82 of 270 A01 ITS ITS 5.5 REFUELING OPERATIONS 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 One auxiliary building gas treatment filter train shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the storage pool.

ACTION:

a. With no auxiliary building gas treatment filter train OPERABLE, suspend all operations involving movement of fuel within the spent fuel pit or crane operation with loads over the spent fuel pit until at least one auxiliary building gas treatment filter train is restored to OPERABLE status.

See ITS

b. The provisions of Specification 3.0.3 are not applicable. 3.7.12 SURVEILLANCE REQUIREMENTS 4.9.12 The above required auxiliary buildings gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

Add proposed ITS 5.5.9 generic program statement A03 5.5.9 b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and 5.5.9.b uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm +/- 10%.

LA02 5.5.9.c 2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5%

when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.

5.5.9.a 3. Verifying a system flow rate of 9000 cfm +/- 10% during system operations when tested in 5.5.9.b accordance with ANSI N510-1975.

5.5.9.d April 11, 2005 SEQUOYAH - UNIT 1 3/4 9-12 Amendment No. 263, 301 Page 11 of 64 Enclosure 2, Volume 16, Rev. 0, Page 82 of 270

Enclosure 2, Volume 16, Rev. 0, Page 83 of 270 A01 ITS ITS 5.5 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

LA02 5.5.9

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with 5.5.9.c Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.

5.5.9 d. At least once per 18 months by:

5.5.9.d 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter train starts on a high radiation signal from the fuel pool radiation See ITS monitoring system. 3.7.12 5.5.9.e 3. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N510-1975.

5.5.9

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in 5.5.9.a accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

5.5.9

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 November 2, 2000 SEQUOYAH - UNIT 1 3/4 9-13 Amendment No. 88, 122, 263 Page 12 of 64 Enclosure 2, Volume 16, Rev. 0, Page 83 of 270

Enclosure 2, Volume 16, Rev. 0, Page 84 of 270 A01 ITS ITS 5.5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.1.1 This specification is deleted.

3.11.1.2 This specification is deleted.

3.11.1.3 This specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-1 Amendment No. 42, 148 Page 13 of 64 Enclosure 2, Volume 16, Rev. 0, Page 84 of 270

Enclosure 2, Volume 16, Rev. 0, Page 85 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement A04 5.5.10 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited by 5.5.10.c the following expression: LA03 concentration of isotope i i

(effluent concentration 6,700 limit of isotope i) less than the amount that would result in excluding tritium and dissolved or entrained noble gases. concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest A05 potable water supply and the nearest surface water

a. Condensate Storage Tank supply in an unrestricted area, in the event of an

b. Steam Generator Layup Tank uncontrolled release of the tanks' contents.

c. Outside temporary tanks for radioactive liquid APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the LA03 above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5.5.10.c 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. LA03 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies. A04 April 11, 2005 SEQUOYAH - UNIT 1 3/4 11-2 Amendment No. 42, 148, 174, 301 Page 14 of 64 Enclosure 2, Volume 16, Rev. 0, Page 85 of 270

Enclosure 2, Volume 16, Rev. 0, Page 86 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.2.1 This specification is deleted.

3.11.2.2 This specification is deleted.

3.11.2.3 This specification is deleted.

3.11.2.4 This specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-3 Amendment No. 42, 65, 109, 114, 148 Page 15 of 64 Enclosure 2, Volume 16, Rev. 0, Page 86 of 270

Enclosure 2, Volume 16, Rev. 0, Page 87 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement A04 5.5.10 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or 5.5.10.a equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in a waste gas holdup tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits LA03 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With the concentration of oxygen in a waste gas holdup tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, without delay suspend all additions of waste gases to the affected waste gas holdup tank and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5.5.10.a 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the above limits by monitoring the waste gas additions to the waste gas holdup system with LA03 the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04 April 11, 2005 SEQUOYAH - UNIT 1 3/4 11-4 Amendment No. 42, 148, 301 Page 16 of 64 Enclosure 2, Volume 16, Rev. 0, Page 87 of 270

Enclosure 2, Volume 16, Rev. 0, Page 88 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement A04 5.5.10 3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or 5.5.10.b equal to 50,000 curies of noble gases (considered as Xe-133).

to less than the amount that would result in a whole APPLICABILITY: At all times. body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

ACTION:

LA03

a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5.5.10.b 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank. LA03 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04 April 11, 2005 SEQUOYAH - UNIT 1 3/4 11-5 Amendment No. 42, 148, 301 Page 17 of 64 Enclosure 2, Volume 16, Rev. 0, Page 88 of 270

Enclosure 2, Volume 16, Rev. 0, Page 89 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.3 DELETED LIMITING CONDITION FOR OPERATION 3.11.3 This specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-6 Amendment No. 42, 148 Page 18 of 64 Enclosure 2, Volume 16, Rev. 0, Page 89 of 270

Enclosure 2, Volume 16, Rev. 0, Page 90 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.4 DELETED LIMITING CONDITION FOR OPERATION 3.11.4 This specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 11-7 Amendment No. 42, 148 Page 19 of 64 Enclosure 2, Volume 16, Rev. 0, Page 90 of 270

Enclosure 2, Volume 16, Rev. 0, Page 91 of 270 A01 ITS ITS 5.5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.12.1 This Specification is deleted.

3.12.2 This Specification is deleted.

3.12.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 1 3/4 12-1 Amendment No. 42, 114, 148 Page 20 of 64 Enclosure 2, Volume 16, Rev. 0, Page 91 of 270

Enclosure 2, Volume 16, Rev. 0, Page 92 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS

b. Refueling operations.

c. Surveillance and test activities of safety-related equipment.

d. DELETED

e. DELETED

f. Fire Protection Program implementation.

g. DELETED See ITS 5.4

h. Quality Assurance Program for effluent and environmental monitoring, using the guidance contained in Regulatory Guide 4.15, December 1977, or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975.

i. OFFSITE DOSE CALCULATION MANUAL implementation.

6.8.2 DELETED 6.8.3 DELETED 5.5 6.8.4 The following programs shall be established, implemented, and maintained.

5.5.2 a. Primary Coolant Sources Outside Containment 5.5.2 A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The February 11, 2003 SEQUOYAH - UNIT 1 6-7 Amendment No. 42, 58, 74, 148, 178, 233, 281 Page 21 of 64 Enclosure 2, Volume 16, Rev. 0, Page 92 of 270

Enclosure 2, Volume 16, Rev. 0, Page 93 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.2 systems include the safety injection system, residual heat removal system, chemical and volume control system, containment spray system, and RCS sampling system. The program shall include the following:

5.5.2.a (i) Preventive maintenance and periodic visual inspection requirements, and 5.5.2.b (ii) Integrated leak test requirements for each system at lease once per 18 months.

5.5.2 The provisions of SR 4.0.2 are applicable.

b. In-Plant Radiation Monitoring (DELETED) 5.5.8 c. Secondary Water Chemistry 5.5.8 A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

5.5.8.a (i) Identification of a sampling schedule for the critical variables and control points for these variables, 5.5.8.b (ii) Identification of the procedures used to measure the values of the critical variables, 5.5.8.c (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, 5.5.8.d (iv) Procedures for the recording and management of data, 5.5.8.e (v) Procedures defining corrective actions for off-control point chemistry conditions, 5.5.8.f (vi) Procedures identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action.

February 11, 2003 SEQUOYAH - UNIT 1 6-8 Amendment No. 58, 74, 178, 233, 281 Page 22 of 64 Enclosure 2, Volume 16, Rev. 0, Page 93 of 270

Enclosure 2, Volume 16, Rev. 0, Page 94 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS

d. DELETED

e. DELETED 5.5.3

f. Radioactive Effluent Controls Program 5.5.3 A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

5.5.3.a 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the ODCM, 5.5.3.b 2) Limitations on the concentrations of radioactive material released in liquid effluents to, UNRESTRICTED AREAS conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, 5.5.3.c 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, 5.5.3.d 4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5.5.3.e 5) Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

February 11, 2003 SEQUOYAH - UNIT 1 6-9 Amendment Nos. 12, 32, 58, 74, 148, 159, 174, 272, 281 Page 23 of 64 Enclosure 2, Volume 16, Rev. 0, Page 94 of 270

Enclosure 2, Volume 16, Rev. 0, Page 95 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.3.f 6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 5.5.3.g 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:

5.5.3.g.1 1. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and 5.5.3.g.2 2. For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/year to any organ.

5.5.3.h 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 5.5.3.i 9) Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and 5.5.3.j 10) Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

5.5.3 The provisions of SR 4.0.2 and 4.0.3 are applicable to the radioactive effluent controls program surveillance frequency.

g. Radiological Environmental Monitoring Program (DELETED)

February 11, 2003 SEQUOYAH - UNIT 1 6-10 Amendment No. 12, 32, 58, 74, 148, 174, 233, 281 Page 24 of 64 Enclosure 2, Volume 16, Rev. 0, Page 95 of 270

Enclosure 2, Volume 16, Rev. 0, Page 96 of 270 A01 ITS ITS 5.5 5.5.14 h. Containment Leakage Rate Testing Program establish 5.5.14.a A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved This program exemptions. Visual examination and testing, including test intervals and extensions, shall be in accordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 with exceptions provided in the site implementing instructions and the following:

5.5.14.a.1 BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

12.46 5.5.14.b The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig.

11.33 L03 5.5.14.c The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

5.5.14.d Leakage rate acceptance criteria are:

5.5.14.d.1 a. Containment overall leakage rate acceptance criteria is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests; 5.5.14.d.2 b. Air lock testing acceptance criteria are:

5.5.14.d.2 1) 1. Overall air lock leakage rate is 0.05 La when tested at Pa.

5.5.14.d.2 2)

2. For each door, leakage rate is 0.01 La when pressurized to 6 psig for at least two minutes.

5.5.14.d.3 c. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate less than or equal to 0.05 La.

5.5.14.d.4 d. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:

5.5.14.d.4 1) 1. The combined bypass leakage rate to the auxiliary building shall be less than or equal to 0.25 La by applicable Type B and C tests.

5.5.14.d.4 2) 2. Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test. 11.33 5.5.14.f The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

5.5.14.e The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

i. Configuration Risk Management Program (DELETED)

April 13, 2009 SEQUOYAH - UNIT 1 6-10a Amendment No. 217, 241, 281, 287, 323 Page 25 of 64 Enclosure 2, Volume 16, Rev. 0, Page 96 of 270

Enclosure 2, Volume 16, Rev. 0, Page 97 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.12 j. Technical Specification (TS) Bases Control Program 5.5.12 This program provides a means for processing changes to the Bases of TSs.

5.5.12.a a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

5.5.12.b b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

5.5.12.b.1 1. A change in the TS incorporated in the license or 5.5.12.b.2 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

5.5.12.c c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

5.5.12.d d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.7 k. Steam Generator (SG) Program 5.5.7 A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

5.5.7.a a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met.

5.5.7.b b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

5.5.7.b.1 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full February 23, 2006 SEQUOYAH - UNIT 1 6-11 Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306 Page 26 of 64 Enclosure 2, Volume 16, Rev. 0, Page 97 of 270

Enclosure 2, Volume 16, Rev. 0, Page 98 of 270 A01 ITS ITS 5.5 6.0 ADMINISTRATIVE CONTROLS 5.5.7.b.1 power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

5.5.7.b.2

2. Accident induced leakage performance criterion: The accident-induced leakage is not to exceed 1.0 gpm for the faulted SG. The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the maximum leakage rate established in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

5.5.7.b.3

3. The operational leakage performance criterion is specified in LCO 3.4.6.2, Reactor Coolant System, Operational Leakage.

plugging A06 5.5.7.c

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

5.5.7.d d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be plugging performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

5.5.7.d.1

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. installation A06 INSERT 1 5.5.7.d.2

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 L01 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

February 23, 2006 SEQUOYAH - UNIT 1 6-11a Amendment No. 306 Page 27 of 64 Enclosure 2, Volume 16, Rev. 0, Page 98 of 270

Enclosure 2, Volume 16, Rev. 0, Page 99 of 270 ITS 5.5 L01 INSERT 1 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

Insert Page 6-11a Page 28 of 64 Enclosure 2, Volume 16, Rev. 0, Page 99 of 270

Enclosure 2, Volume 16, Rev. 0, Page 100 of 270 A01 ITS ITS 5.5 6.0 ADMINISTRATIVE CONTROLS affected and potentially affected A07 5.5.7.d.3

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

results in more frequent inspections A07 5.5.7.e

e. Provisions for monitoring operational primary-to-secondary leakage.

5.5.4 l. Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4.

STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED ANNUAL REPORTS 1/

See ITS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar 5.6 year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 DELETED 1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

August 2, 2006 SEQUOYAH - UNIT 1 6-11b Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306, 309 Page 29 of 64 Enclosure 2, Volume 16, Rev. 0, Page 100 of 270

Enclosure 2, Volume 16, Rev. 0, Page 101 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) (DELETED) 5.5.1 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.5.1 6.14.1 Changes to the ODCM:

5.5.1.a 1. Shall be documented and records of reviews performed shall be retained in a manner convenient for review. This documentation shall contain:

5.5.1.a.1 a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and 5.5.1.a.2 b. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

plant manager M01 5.5.1.b 2. Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-A.

LA04 5.5.1.c 3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

July 1, 1998 SEQUOYAH - UNIT 1 6-16 Amendment No. 42, 58, 74, 148, 169, 174, 178, 233 Page 30 of 64 Enclosure 2, Volume 16, Rev. 0, Page 101 of 270

Enclosure 2, Volume 16, Rev. 0, Page 102 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid)** (DELETED)

February 11, 2003 SEQUOYAH - UNIT 1 6-17 Amendment No. 42, 58, 74, 148, 169, 174, 233, 281 Page 31 of 64 Enclosure 2, Volume 16, Rev. 0, Page 102 of 270

Enclosure 2, Volume 16, Rev. 0, Page 103 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.11 6.16 DIESEL FUEL OIL TESTING PROGRAM A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

5.5.11.a a. Acceptability of new fuel oil prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. A clear and bright appearance with proper color; or a water and sediment content within limits L02 5.5.11.b b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and 5.5.11.c c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days in accordance LA06 with ASTM D-2276, Method A. A12 ASTM D6217-11 The provisions of SR 3.0.2 and SR 3.0.3 are A08 5.5.16 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM applicable to the Diesel Fuel Oil Testing A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:

5.5.16.a a. The definition of the CRE and the CRE boundary.

5.5.16.b b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

5.5.16.c c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

required by the VFTP A09 5.5.16.d d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate of 4000 cubic feet per minute plus or minus 10 percent, at a Frequency of 36 months A10 on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

October 28, 2008 SEQUOYAH - UNIT 1 6-18 Amendment No. 261, 321 Page 32 of 64 Enclosure 2, Volume 16, Rev. 0, Page 103 of 270

Enclosure 2, Volume 16, Rev. 0, Page 104 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.16 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (continued) 5.5.16.e e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

5.5.16.f f. The provisions of SR 4.0.2 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.5.13 Add proposed program 5.5.13 M02 5.5.15 Add proposed program 5.5.15 M03 5.5.17 Add proposed program 5.5.17 M04 October 28, 2008 SEQUOYAH - UNIT 1 6-19 Amendment No. 321 Page 33 of 64 Enclosure 2, Volume 16, Rev. 0, Page 104 of 270

Enclosure 2, Volume 16, Rev. 0, Page 105 of 270 A01 ITS ITS 5.5 DEFINITIONS IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

a. Leakage, such as that from pump seals or valve packing (except reactor coolant pump seal injection or leakoff) that is captured and conducted to collection systems or a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located See ITS 1.0 and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor coolant system leakage through a steam generator to the secondary system (primary to secondary leakage).

MEMBER(S) OF THE PUBLIC 1.17 DELETED OFFSITE DOSE CALCULATION MANUAL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and 5.5.1.a parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) 5.5.1.b descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant See ITS 1.0 instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

May 22, 2007 SEQUOYAH - UNIT 2 1-4 Amendment Nos. 63, 134, 146, 159, 165, 169, 250, 272, 305 Page 34 of 64 Enclosure 2, Volume 16, Rev. 0, Page 105 of 270

Enclosure 2, Volume 16, Rev. 0, Page 106 of 270 A01 ITS ITS 5.5 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.3 (Continued) up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified surveillance interval, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation See ITS must immediately be declared not met, and the applicable ACTION(s) must be entered. 3.0 4.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 4.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

5.5.6 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be as follows:

pumps and valves LA01 Inservice Inspection Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

LA01

a. Provisions that inservice testing of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a;

b. The provisions of SR 4.0.2 are applicable to the frequencies for performing inservice inspection activities; 5.5.5 c. Inspection of each reactor coolant pump flywheel per the recommendation of Regulation Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975 or in lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place ultrasonic examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (magnetic particle and/or liquid penetrant) of exposed surfaces of the removed flywheels may be conducted at 20-year intervals (the provisions of SR 4.0.2 are not applicable); and A02

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirement of any TS. LA01 5.5.6 Inservice Testing Program pumps and valves LA01 5.5.6 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

5.5.6.a October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-3 Amendment No. 69, 152, 198, 263, 271, 283, 290 Page 35 of 64 Enclosure 2, Volume 16, Rev. 0, Page 106 of 270

Enclosure 2, Volume 16, Rev. 0, Page 107 of 270 A01 ITS ITS 5.5 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 5.5.6 4.0.5 (Continued) 5.5.6.a a. Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a;

b. Testing frequencies applicable to the ASME OM Code and applicable Addenda as follows:

ASME OM Code and applicable Addenda Required frequencies for terminology for inservice performing inservice testing activities testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days 5.5.6.b c. The provisions of SR 4.0.2 are applicable to the above required Frequencies and other normal and accelerated frequencies specified as 2 years or less in the Inservice Test Program for performing inservice testing activities; 5.5.6.c

d. The provisions of SR 4.0.3 are applicable to inservice testing and activities; and 5.5.6.d e. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

October 4, 2006 SEQUOYAH - UNIT 2 3/4 0-4 Amendment No. 69, 152, 198, 263, 271, 283, 297 Page 36 of 64 Enclosure 2, Volume 16, Rev. 0, Page 107 of 270

Enclosure 2, Volume 16, Rev. 0, Page 108 of 270 A01 ITS ITS 5.5 CONTAINMENT SYSTEMS EMERGENCY GAS TREATMENT SYSTEM - EGTS - CLEANUP SUBSYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.8 Two independent emergency gas treatment system cleanup subsystems (EGTS) shall be OPERABLE.

See ITS 3.6.10 APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one EGTS cleanup subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.1.8 Each EGTS cleanup subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, See ITS flow through the HEPA filters and charcoal adsorbers and verifying that the system operates 3.6.10 for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

Add proposed ITS 5.5.9 generic program statement A03

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or 5.5.9 charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and 5.5.9.b uses the test procedures of Regulatory Position C.5.a., C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm +/- 10%.

LA02

2. Verifying within 31 days after removal that a laboratory analysis of a representative 5.5.9.c carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5%

when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9.a 3. Verifying a system flow rate of 4000 cfm +/- 10% during system operation when tested in 5.5.9.b 5.5.9.c accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 2 3/4 6-13 Amendment No. 254 Page 37 of 64 Enclosure 2, Volume 16, Rev. 0, Page 108 of 270

Enclosure 2, Volume 16, Rev. 0, Page 109 of 270 A01 ITS ITS 5.5 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

LA02 5.5.9 c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with 5.5.9.c Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9.d d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 5 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm + 10%.

2. Verifying that the filter train starts on a Phase A containment isolation Test Signal.

See ITS 3.6.10

3. Verify the operation of the filter cooling bypass valves.

4. Verifying that each system produces a negative pressure of greater than or equal to 0.5 See ITS inches W.G. in the annulus within 1 minute after a start signal. 3.6.7 5.5.9

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter 5.5.9.a banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm + 10%.

5.5.9

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm + 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 November 2, 2000 SEQUOYAH - UNIT 2 3/4 6-14 Amendment No. 11, 77, 92, 254 Page 38 of 64 Enclosure 2, Volume 16, Rev. 0, Page 109 of 270

Enclosure 2, Volume 16, Rev. 0, Page 110 of 270 A01 ITS ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Add proposed ITS 5.5.9 generic program statement A03 5.5.9 c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses 5.5.9.b the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 4000 cfm +/- 10%.

LA02 5.5.9.c 2. Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9.a 5.5.9.b 3. Verifying a system flow rate of 4000 cfm + 10% during system operation when tested in 5.5.9.d accordance with ANSI N510-1975.

LA02 5.5.9 d. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with 5.5.9.c Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ATSM D3803-1989 at a temperature of 30°C (86°F) and a relative humidity of 70%.

5.5.9 e. At least once per 18 months by:

5.5.9.d

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the system at a flow rate of 4000 cfm +/- 10%.

2. Verifying that on a safety injection signal or high radiation signal from the air intake stream, See ITS the system automatically diverts its inlet flow through the HEPA filters and charcoal adsorber 3.7.10 banks.

5.5.9

f. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in 5.5.9.a accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

5.5.9

g. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm +/- 10%.

h. Perform required CRE unfiltered air inleakage testing in accordance with the Control Room See ITS 3.7.10 Envelope Habitability Program.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 October 28, 2008 SEQUOYAH - UNIT 2 3/4 7-18 Amendment No. 60, 77, 254, 313 Page 39 of 64 Enclosure 2, Volume 16, Rev. 0, Page 110 of 270

Enclosure 2, Volume 16, Rev. 0, Page 111 of 270 A01 ITS ITS 5.5 PLANT SYSTEMS 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 Two independent auxiliary building gas treatment filter trains shall be OPERABLE.

APPLICABILITY: Modes 1, 2, 3 and 4.

ACTION:

With one auxiliary building gas treatment filter train inoperable, restore the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. See ITS 3.7.12 SURVEILLANCE REQUIREMENTS 4.7.8 Each auxiliary building gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filter and charcoal adsorber train and verifying that the system operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

Add proposed ITS 5.5.9 generic program statement A03 5.5.9 b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

5.5.9.a 1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria 5.5.9.b and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 and 9), and the system flow rate is 9000 cfm + 10%.

LA02

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory 5.5.9.c Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9.a 3. Verifying a system flow rate of 9000 cfm + 10% during system operation when 5.5.9.b 5.5.9.d tested in accordance with ANSI N510-1975.

November 2, 2000 SEQUOYAH - UNIT 2 3/4 7-19 Amendment No. 254 Page 40 of 64 Enclosure 2, Volume 16, Rev. 0, Page 111 of 270

Enclosure 2, Volume 16, Rev. 0, Page 112 of 270 A01 ITS ITS 5.5 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

LA02 5.5.9

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with 5.5.9.c Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9 d. At least once per 18 months by:

5.5.9.d

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter trains start on a Containment Phase A Isolation test signal.

3. Verifying that the system maintains the spent fuel storage area and the ESF pump See ITS 3.7.12 rooms at a pressure equal to or more negative than minus 1/4 inch water gauge relative the outside atmosphere while maintaining a total system flow of 9000 cfm +/-

10%.

5.5.9.e 4. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N510-1975.

5.5.9

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place 5.5.9.a in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/-

10%.

5.5.9

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 August 18, 2005 SEQUOYAH - UNIT 2 3/4 7-20 Amendment No. 77, 111, 254, 293 Page 41 of 64 Enclosure 2, Volume 16, Rev. 0, Page 112 of 270

Enclosure 2, Volume 16, Rev. 0, Page 113 of 270 A01 ITS ITS 5.5 TABLE 4.8-1a DIESEL GENERATOR BATTERY SURVEILLANCE REQUIREMENTS CATEROGY A(1) CATEGORY B(2)

Parameter Limit for each designated Limits for each connected Allowable(3) pilot cell cell value for each connected cell See ITS 3.8.6 Electrolyte >Minimum level indication >Minimum level indication Above top of Level mark, and 1/4 above mark, and 1/4 above plates, maximum level indication maximum level indication and not mark mark overflowing 5.5.15.b.2 Float Voltage 2.13 volts 2.13 volts(c) > 2.07 volts Not more than

.020 below the average of all 1.190 connected cells Specific 1.195(b) Average of all Average of all Gravity(a) connected cells connected cells

> 1.200 1.190(b)

See ITS 3.8.6 (a) Corrected for electrolyte temperature and level.

(b) Or battery charging current is less than 2 amps.

(c) Corrected for average electrolyte temperature. LA05 (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. See ITS (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered 3.8.6 OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.

(3) Any Category B parameter not within its allowable value indicates an inoperable battery.

SEQUOYAH - UNIT 2 3/4 8-8a Page 42 of 64 Enclosure 2, Volume 16, Rev. 0, Page 113 of 270

Enclosure 2, Volume 16, Rev. 0, Page 114 of 270 A01 ITS ITS 5.5 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEROGY A(1) CATEGORY B(2)

Parameter Limit for each designated Limits for each connected Allowable(3) pilot cell cell value for each connected cell See ITS Electrolyte >Minimum level indication >Minimum level indication Above top of 3.8.6 Level mark, and 1/4 above mark, and 1/4 above plates, maximum level indication maximum level indication and not mark mark overflowing 5.5.15.b.2 Float Voltage 2.13 volts 2.13 volts(c) > 2.07 volts Not more than

.020 below the average of all 1.195 connected cells See ITS (b) 3.8.6 Specific 1.200 Average of all Average of all Gravity(a) connected cells connected cells

> 1.205 1.195(b)

(a) Corrected for electrolyte temperature and level.

(b) Or battery charging current is less than 2 amps.

(c) Corrected for average electrolyte temperature. LA05 (1) For any Category A parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the Category B measurements are taken and found to be within their allowable values, and provided all parameter(s) are restored to within limits within the next 6 days. See ITS 3.8.6 (2) For any Category B parameter(s) outside the limit(s) shown, the battery may be considered OPERABLE provided that they are within their allowable values and provided the parameter(s) are restored to within limits within 7 days.

(3) Any Category B parameter not within its allowable value indicates an inoperable battery.

SEQUOYAH - UNIT 2 3/4 8-14 Page Page 4339 of of 64 58 Enclosure 2, Volume 16, Rev. 0, Page 114 of 270

Enclosure 2, Volume 16, Rev. 0, Page 115 of 270 A01 ITS ITS 5.5 REFUELING OPERATIONS 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 One auxiliary building gas treatment filter train shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in the storage pool.

ACTION:

a. With no auxiliary building gas treatment filter train OPERABLE, suspend all operations involving movement of fuel within the spent fuel pit or crane operation with loads over the spent fuel pit until at least one auxiliary building gas treatment filter train is restored to See ITS OPERABLE status. 3.7.12

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required auxiliary building gas treatment filter train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on.

Add proposed ITS 5.5.9 generic program statement A03

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or 5.5.9 charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the system by:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Regulatory 5.5.9.a Guide 1.52, Revision 2, March 1978 (except for the provisions of ANSI N510 Sections 8 5.5.9.b and 9), and the system flow rate is 9000 cfm + 10%.

LA02

2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory 5.5.9.c Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5%

when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

3. Verifying a system flow rate of 9000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

5.5.9.a 5.5.9.b 5.5.9.d April 11, 2005 SEQUOYAH - UNIT 2 3/4 9-14 Amendment No. 254, 290 Page 44 of 64 Enclosure 2, Volume 16, Rev. 0, Page 115 of 270

Enclosure 2, Volume 16, Rev. 0, Page 116 of 270 A01 ITS ITS 5.5 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

LA02 5.5.9

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of representative carbon sample obtained in accordance with 5.5.9.c Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86° F) and a relative humidity of 70%.

5.5.9 d. At least once per 18 months by:

5.5.9.d 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3 inches Water Gauge while operating the filter train at a flow rate of 9000 cfm +/- 10%.

2. Verifying that the filter train starts on a high radiation signal from the fuel pool radiation See ITS 3.7.12 monitoring system.

5.5.9.e 3. Verifying that the heaters dissipate 32 +/- 3.2 kw when tested in accordance with ANSI N510-1975.

5.5.9

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in-place in 5.5.9.a accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

5.5.9

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 9000 cfm +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. A03 November 2, 2000 SEQUOYAH - UNIT 2 3/4 9-15 Amendment No. 77, 111, 254 Page 45 of 64 Enclosure 2, Volume 16, Rev. 0, Page 116 of 270

Enclosure 2, Volume 16, Rev. 0, Page 117 of 270 A01 ITS ITS 5.5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.1.1 This Specification is deleted.

3.11.1.2 This Specification is deleted.

3.11.1.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-1 Amendment No. 34, 134 Page 46 of 64 Enclosure 2, Volume 16, Rev. 0, Page 117 of 270

Enclosure 2, Volume 16, Rev. 0, Page 118 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement A04 5.5.10 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited by 5.5.10.c the following expression: LA03 I

concentration of isotope i i

(effluent concentration 6,700 limit of isotope i) less than the amount that would result in excluding tritium and dissolved or entrained noble gases. concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water A05

a. Condensate Storage Tank supply in an unrestricted area, in the event of an

b. Steam Generator Layup Tank uncontrolled release of the tanks' contents.

c. Outside temporary tanks for radioactive liquid APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit. LA03

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5.5.10.c 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. LA03 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies. A04 April 11, 2005 SEQUOYAH - UNIT 2 3/4 11-2 Amendment No. 134, 165, 290 Page 47 of 64 Enclosure 2, Volume 16, Rev. 0, Page 118 of 270

Enclosure 2, Volume 16, Rev. 0, Page 119 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS LIMITING CONDITION FOR OPERATION 3.11.2.1 This Specification is deleted.

3.11.2.2 This Specification is deleted.

3.11.2.3 This Specification is deleted.

3.11.2.4 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-3 Amendment No. 34, 57, 99, 104, 134 Page 48 of 64 Enclosure 2, Volume 16, Rev. 0, Page 119 of 270

Enclosure 2, Volume 16, Rev. 0, Page 120 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement A04 5.5.10 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or 5.5.10.a equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in a waste gas holdup tank greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits LA03 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With the concentration of oxygen in a waste gas holdup tank greater than 4% by volume and the hydrogen concentration greater than 2% by volume, without delay suspend all additions of waste gases to the affected waste gas holdup tank and reduce the concentration of oxygen to less than or equal to 2% by volume without delay.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5.5.10.a 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the above limits by monitoring the waste gas additions to the waste gas holdup system with LA03 the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04 April 11, 2005 SEQUOYAH - UNIT 2 3/4 11-4 Amendment No. 34, 134, 290 Page 49 of 64 Enclosure 2, Volume 16, Rev. 0, Page 120 of 270

Enclosure 2, Volume 16, Rev. 0, Page 121 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS GAS DECAY TANKS LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement A04 5.5.10 3.11.2.6 The quantity of radioactivity contained in each gas decay tank shall be limited to less than or 5.5.10.b equal to 50,000 curies of noble gases (considered as Xe-133).

to less than the amount that would result in a whole APPLICABILITY: At all times. body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

ACTION:

LA03

a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5.5.10.b 4.11.2.6 The quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank. LA03 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas Tank Monitoring Program Surveillance Frequencies. A04 April 11, 2005 SEQUOYAH - UNIT 2 3/4 11-5 Amendment No. 34, 134, 290 Page 50 of 64 Enclosure 2, Volume 16, Rev. 0, Page 121 of 270

Enclosure 2, Volume 16, Rev. 0, Page 122 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.3 DELETED LIMITING CONDITION FOR OPERATION 3.11.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-6 Amendment No. 34, 134 Page 51 of 64 Enclosure 2, Volume 16, Rev. 0, Page 122 of 270

Enclosure 2, Volume 16, Rev. 0, Page 123 of 270 A01 ITS ITS 5.5 RADIOACTIVE EFFLUENTS 3/4.11.4 DELETED LIMITING CONDITION FOR OPERATION 3.11.4 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 11-7 Amendment No. 34, 134 Page 52 of 64 Enclosure 2, Volume 16, Rev. 0, Page 123 of 270

Enclosure 2, Volume 16, Rev. 0, Page 124 of 270 A01 ITS ITS 5.5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION 3.12.1 This Specification is deleted.

3.12.2 This Specification is deleted.

3.12.3 This Specification is deleted.

November 16, 1990 SEQUOYAH - UNIT 2 3/4 12-1 Amendment No. 34, 104, 134 Page 53 of 64 Enclosure 2, Volume 16, Rev. 0, Page 124 of 270

Enclosure 2, Volume 16, Rev. 0, Page 125 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS

h. Quality Assurance Program for effluent and environmental monitoring, using the guidance contained in Regulatory Guide 4.15, December 1977 or Regulatory Guide 1.21, Rev. 1, 1974 and Regulatory Guide 4.1, Rev. 1, 1975.

See ITS

i. OFFSITE DOSE CALCULATION MANUAL implementation. 5.4 6.8.2 DELETED 6.8.3 DELETED 5.5 6.8.4 The following programs shall be established, implemented, and maintained.

5.5.2 a. Primary Coolant Sources Outside Containment 5.5.2 A program to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the safety injection system, residual heat removal system, chemical and volume control system, containment spray system, and RCS sampling system. The program shall include the following:

5.5.2.a (i) Preventive maintenance and periodic visual inspection requirements, and 5.5.2.b (ii) Integrated leak test requirements for each system at lease once per 18 months.

5.5.2 The provisions of SR 4.0.2 are applicable

b. In-Plant Radiation Monitoring (DELETED) 5.5.8 c. Secondary Water Chemistry 5.5.8 A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include:

5.5.8.a (i) Identification of a sampling schedule for the critical variables and control points for these variables, 5.5.8.b (ii) Identification of the procedures used to measure the values of the critical variables, 5.5.8.c (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage 5.5.8.d (iv) Procedures for the recording and management of data, 5.5.8.e (v) Procedures defining corrective actions for off-control point chemistry conditions, 5.5.8.f (vi) Procedures identifying (a) the authority responsible for the interpretation of the data; and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Deleted February 11, 2003 SEQUOYAH - UNIT 2 6-7 Amendment No. 34, 50, 66, 134, 149, 169, 223, 272 Page 54 of 64 Enclosure 2, Volume 16, Rev. 0, Page 125 of 270

Enclosure 2, Volume 16, Rev. 0, Page 126 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS

e. DELETED 5.5.3 f. Radioactive Effluent Controls Program 5.5.3 A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

5.5.3.a 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the ODCM, 5.5.3.b 2) Limitations on the concentrations of radioactive material released in liquid effluents to, UNRESTRICTED AREAS conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, 5.5.3.c 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, 5.5.3.d 4) Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5.5.3.e 5) Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

5.5.3.f

6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases February 11, 2003 SEQUOYAH - UNIT 2 6-8 Amendment No. 28, 50, 66, 134, 165, 261, 272 Page 55 of 64 Enclosure 2, Volume 16, Rev. 0, Page 126 of 270

Enclosure 2, Volume 16, Rev. 0, Page 127 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 6.8.4 f. Radioactive Effluent Controls Program (Cont.)

5.5.3.f of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 5.5.3.g 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be in accordance with the following:

5.5.3.g.1 1. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and 5.5.3.g.2 2. For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/year to any organ.

5.5.3.h 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 5.5.3.i 9) Limitations on the annual and quarterly doses to a member of the public from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and 5.5.3.j 10) Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

5.5.3 The provisions of SR 4.0.2 and 4.0.3 are applicable to the radioactive effluent controls program surveillance frequency.

g. Radiological Environmental Monitoring Program (DELETED) 5.5.14 h. Containment Leakage Rate Testing Program establish 5.5.14.a A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved This program exemptions. Visual examination and testing, including test intervals and extensions, shall be in accordance with Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995 with exceptions provided in the site implementing instructions and the following:

5.5.14.a.1 BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (13.2 psig) and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

12.46 5.5.14.b The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig.

11.33 L03 5.5.14.c The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.

April 13, 2009 SEQUOYAH - UNIT 2 6-9 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 265, 272, 276, 315 Page 56 of 64 Enclosure 2, Volume 16, Rev. 0, Page 127 of 270

Enclosure 2, Volume 16, Rev. 0, Page 128 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.14.d Leakage rate acceptance criteria are:

5.5.14.d.1 a. Containment overall leakage rate acceptance criteria is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests; 5.5.14.d.2 b. Air lock testing acceptance criteria are:

5.5.14.d.2 1) 1)Overall air lock leakage rate is 0.05 La when tested at Pa.

5.5.14.d.2 2)

2) For each door, leakage rate is 0.01 La when pressurized to 6 psig for at least two minutes.

5.5.14.d.3 c. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate less than or equal to 0.05 La.

5.5.14.d.4 d. BYPASS LEAKAGE PATHS TO THE AUXILIARY BUILDING acceptance criteria are:

5.5.14.d.4 1) 1. The combined bypass leakage rate to the auxiliary building shall be less than or equal to 0.25 La by applicable Type B and C tests.

5.5.14.d.4 2)

2. Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test. 11.33 L03 5.5.14.f The provisions of SR 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

5.5.14.e The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

i. Configuration Risk Management Program (DELETED) 5.5.12

j. Technical Specification (TS) Bases Control Program 5.5.12 This program provides a means for processing changes to the Bases of these TSs.

5.5.12.a a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

5.5.12.b

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

5.5.12.b.1

1. A change in the TS incorporated in the license or 5.5.12.b.2

2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

5.5.12.c

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

April 13, 2009 SEQUOYAH - UNIT 2 6-10 Amendment No. 28, 50, 64, 66, 134, 165, 202, 207, 223, 231, 265, 271, 272, 276, 298, 305, 315 Page 57 of 64 Enclosure 2, Volume 16, Rev. 0, Page 128 of 270

Enclosure 2, Volume 16, Rev. 0, Page 129 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.12.d d. Proposed changes that meet the criteria of Specification 6.8.4.j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.7 k. Steam Generator (SG) Program 5.5.7 A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

5.5.7.a a. Provisions for Condition Monitoring Assessments.

Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

5.5.7.b b. Provisions for Performance Criteria for SG Tube Integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

5.5.7.b.1 1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents (DBAs). This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the DBA primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the DBAs, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

5.5.7.b.2 2. Accident induced leakage performance criterion: The accident-induced leakage is not to exceed 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs. The A11 primary-to-secondary accident induced leakage rate for any DBA, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

5.5.7.b.3 3. The operational leakage performance criterion is specified in Limiting Condition for Operation (LCO) 3.4.6.2, Reactor Coolant System, Operational Leakage.

July 10, 2012 SEQUOYAH - UNIT 2 6-10a Amendment No. 28, 50, 64, 66, 134,165, 202, 207, 223, 231, 265, 271, 272, 276, 298, 305, 323 Page 58 of 64 Enclosure 2, Volume 16, Rev. 0, Page 129 of 270

Enclosure 2, Volume 16, Rev. 0, Page 130 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS plugging A06 5.5.7.c c. Provisions for SG Tube Repair Criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

5.5.7.d d. Provisions for SG Tube Inspections.

plugging Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

5.5.7.d.1

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. installation A06 INSERT 1 5.5.7.d.2 2. Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective L01 full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SGs shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

affected and potentially affected A07 5.5.7.d.3 3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack. results in more frequent inspections A07 5.5.7.e e. Provisions for Monitoring Operational Primary-to-Secondary Leakage.

5.5.4 l. Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

July 10, 2012 SEQUOYAH - UNIT 2 6-10b Amendment No. 28, 34, 50, 64, 66, 107, 134, 165, 207, 223, 231, 271, 272, 289, 293, 305, 315, 318, 323 Page 59 of 64 Enclosure 2, Volume 16, Rev. 0, Page 130 of 270

Enclosure 2, Volume 16, Rev. 0, Page 131 of 270 ITS 5.5 L01 INSERT 1 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

Insert Page 6-10b Page 60 of 62 Enclosure 2, Volume 16, Rev. 0, Page 131 of 270

Enclosure 2, Volume 16, Rev. 0, Page 132 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) (DELETED) 5.5.1 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.5.1 6.14.1 Changes to the ODCM:

5.5.1.a 1. Shall be documented and records of reviews performed shall be retained in a manner convenient for review. This documentation shall contain:

5.5.1.a.1 a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and 5.5.1.a.2 b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

plant manager M01 5.5.1.b 2. Shall become effective after review and acceptance by the process described in TVA-NQA-PLN89-A.

LA04 5.5.1.c 3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

July 1, 1998 SEQUOYAH - UNIT 2 6-17 Amendment Nos. 34, 50, 66, 134, 159, 165, 169, 223 Page 61 of 64 Enclosure 2, Volume 16, Rev. 0, Page 132 of 270

Enclosure 2, Volume 16, Rev. 0, Page 133 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid)** (DELETED)

February 11, 2003 SEQUOYAH - UNIT 2 6-18 Amendment Nos. 34, 50, 66, 134, 159, 165, 223, 272 Page 62 of 64 Enclosure 2, Volume 16, Rev. 0, Page 133 of 270

Enclosure 2, Volume 16, Rev. 0, Page 134 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.11 6.16 DIESEL FUEL OIL TESTING PROGRAM A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

5.5.11.a a. Acceptability of new fuel oil prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. A clear and bright appearance with proper color; or a water and sediment content within limits L02 5.5.11.b b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and 5.5.11.c c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days in accordance LA06 with ASTM D-2276, Method A. A12 ASTM D6217-11 The provisions of SR 3.0.2 and SR 3.0.3 are A08 5.5.16 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM applicable to the Diesel Fuel Oil Testing A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:

5.5.16.a a. The definition of the CRE and the CRE boundary.

5.5.16.b b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

5.5.16.c c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

required by the VFTP A09 5.5.16.d d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate of 4000 cubic feet per minute plus or minus 10 percent, at a Frequency of 36 months A10 on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

October 28, 2008 SEQUOYAH - UNIT 2 6-19 Amendment No. 252, 313 Page 63 of 64 Enclosure 2, Volume 16, Rev. 0, Page 134 of 270

Enclosure 2, Volume 16, Rev. 0, Page 135 of 270 A01 ITS ITS 5.5 ADMINISTRATIVE CONTROLS 5.5.16 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM (continued) 5.5.16.e e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

5.5.16.f f. The provisions of SR 4.0.2 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

5.5.13 Add proposed program 5.5.13 M02 5.5.15 Add proposed program 5.5.15 M03 5.5.17 Add proposed program 5.5.17 M04 October 28, 2008 SEQUOYAH - UNIT 2 6-20 Amendment No. 313 Page 64 of 64 Enclosure 2, Volume 16, Rev. 0, Page 135 of 270

Enclosure 2, Volume 16, Rev. 0, Page 136 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.)

are made to obtain consistency with NUREG - 1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 4.0.5.c states in part, that the provisions of CTS SR 4.0.2 are not applicable for the 20 year interval reactor coolant pump flywheel inspection. ITS 5.5.5 requires a program to provide for the inspection of each reactor coolant pump flywheel. This changes the CTS by not stating that the allowance of ITS SR 3.0.2 is not applicable.

This change is acceptable because no changes have been made to the existing requirements. The CTS and proposed ITS 5.5.5 continue to require the same reactor coolant pump flywheel inspections to be performed. A statement that ITS SR 3.0.2 is not applicable is not needed, as the provisions of SR 3.0.2 do not apply to the programs in ITS Section 5.5, unless specified. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 The Surveillances associated with the ventilation filter testing for the Control Room Ventilation System (CREVS), the Emergency Gas Treatment System (EGTS), and the Auxiliary Building Gas Treatment System (ABGTS) have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.9). As such, a general program statement has been added as ITS 5.5.9. Also, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extensions do apply (as allowed in the CTS). This changes the CTS by moving the ventilation filter testing Surveillances associated with the CREVS, EGTS, and ABGTS to a program in ITS 5.5 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program.

The addition of the program statement is acceptable because it is describing the intent of the CTS Surveillances. The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the CTS, therefore, it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 The liquid holdup tank requirements in CTS 3.11.1.4, the explosive gas mixture requirements of CTS 3.11.2.5, and the gas decay tanks requirements in CTS 3.11.2.6 have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.10). As such, a general program statement has been added. Also, a statement of applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify the allowances for Surveillance Frequency extensions do apply. This changes the CTS by moving the liquid holdup tank, the explosive gas mixture, and the gas decay tanks requirements to a program in ITS 5.5.10 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program.

Sequoyah Unit 1 and Unit 2 Page 1 of 12 Enclosure 2, Volume 16, Rev. 0, Page 136 of 270

Enclosure 2, Volume 16, Rev. 0, Page 137 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS The addition of the program statement is acceptable because it is describing the intent of the CTS Specification. The addition of ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS.

A05 CTS 3.11.1.4 requires that the quantity of radioactive material contained in the condensate storage tank, steam generator layup tank and outside temporary tanks for radioactive liquid shall be less than or equal to 6700 effluent concentration limit (ECL).

CTS 4.11.1.4 requires a determination that the radioactive material contained in each of the tanks listed in CTS 3.11.1.4 is within limits on a prescribed frequency. ITS 5.5.10.c requires a surveillance program to ensure that the quantity of radioactive material contained in all outdoor temporary liquid radwaste storage tanks, Condensate Storage tank, and Steam Generator Layup tank is less than the amount that would result in concentrations exceeding the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents. This changes the CTS by specifically stating that the program shall meet the 10 CFR 20 requirements (See DOC LA03 for discussion of the removal of the effluent concentration limit).

The addition of the 10 CFR 20 limitations is acceptable because 10 CFR 20.1002 states that this part applies to persons licensed by the Commission to receive, possess, use, transfer, or dispose of byproduct, source, or special nuclear material or to operate a production or utilization facility under parts 30 through 36, 39, 40, 50, 52, 60, 61, 63, 70, or 72 of this chapter. SQN Units 1 and 2 are licensed by the Nuclear Regulatory Commission, in part, under 10 CFR Parts 30, 40, 50, and 70. 10 CFR 20.1302 requires, in part, that the annual average concentrations of radioactive material released in gaseous and liquid effluents at the boundary of the unrestricted area to not exceed the values specified in table 2 of appendix B to part 20. 10 CFR 20, Appendix B, Table 2 refers to effluent concentrations and Column 2 of this table lists limitations associated with water (liquid). Therefore, SQN Units 1 and 2 are currently required to limit effluent releases to within these concentrations. Additionally, restricting the quantity to less than or equal to 6700 ECL (See DOC LA03) provides assurance that the resulting concentrations would be less than the limits of 10 CFR 20. This change is designated as administrative because it does not result in a technical change to the CTS.

A06 CTS 6.8.4.k states the requirements of the Steam Generator (SG) program. ITS 5.5.7 specifies the requirements of the Steam Generator (SG) program based on the latest revision of TSTF-510. This changes CTS 6.8.4.k.c and CTS 6.8.4.k.d by replacing the word "repair" with "plugging" and replacing the word "replacement" with "installation."

CTS 6.8.4.k.d.2 has been revised to reflect TSTF-510-A.

This change is acceptable because no changes have been made to the existing requirements. ITS 5.5.7 continues to require the same steam generator inspections to be performed in accordance with approved TSTF-510-A. This change is designated as administrative because it does not result in technical changes to the CTS.

A07 The first sentence of CTS 6.8.4.k.d.3 states, "If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less)." The first sentence of ITS 5.5.7.d.3 states, "If crack Sequoyah Unit 1 and Unit 2 Page 2 of 12 Enclosure 2, Volume 16, Rev. 0, Page 137 of 270

Enclosure 2, Volume 16, Rev. 0, Page 138 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections)." The proposed change is replacing the words for each SG with the words for each affected and potentially affected SG, and is replacing the parenthetical statement "(whichever is less)" with "(whichever results in more frequent inspections)".

The purpose of CTS 6.8.4.k.d.3 is to restrict the allowable interval to the next scheduled inspection to 24 EFPM or one refueling outage (whichever is less) once cracks have been found in any SG tube. The intent of this requirement is that it applies to the affected SG and to any other SG which may be and affected by the degradation mechanism that caused the known crack(s). This change was made to reflect changes made under TSTF-510 and is acceptable because it clarifies the intent of the paragraph.

This change is designated as administrative because it does not result in a technical change to the CTS.

A08 The Diesel fuel oil testing program (CTS 6.16) has been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.11). As such, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extension do apply. This changes the CTS by specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program.

The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS.

A09 CTS 6.17.d requires, in part, that one train of the Control Room Emergency Ventilation System (CREVS) operates at a flow rate of 4000 cubic feet per minute plus or minus 10 percent. ITS 5.5.16.d requires, in part that one train of the CREVS operates at the flow rate required by the Ventilation Filter Testing Program (VFTP). This changes the CTS by requiring the CREVS to operate at the flow rate required by the VFTP.

The change is acceptable because no change to the existing requirements have been made. ITS 5.5.9 contains the flow requirements for a OPERABLE CREVS train. This change is designated as administrative because it does not result in a technical change to CTS.

A10 CTS 6.17.d requires, in part, measurement of the Control Room Envelope (CRE) boundary be tested using one train of the Control Room Emergency Ventilation System (CREVS) every 36 months on a STAGGERED TEST BASIS. CTS 1.35 defines a STAGGERED TEST BASIS as, "a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval." ITS 5.5.16.d requires a similar test of the CRE boundary with use of one CREVS train every 18 months "on a STAGGERED TEST BASIS." In ITS, a STAGGERED TEST BASIS consists of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, Sequoyah Unit 1 and Unit 2 Page 3 of 12 Enclosure 2, Volume 16, Rev. 0, Page 138 of 270

Enclosure 2, Volume 16, Rev. 0, Page 139 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS where n is the total number of systems, subsystems, channels, or other designated components in the associated function. This changes the CTS by utilizing the ITS definition of a STAGGERED TEST BASIS.

This change is acceptable because the requirements for CRE boundary testing remain unchanged. The ITS definition of STAGGERED TEST BASIS and its application in this requirement do not change the CTS 6.17.d testing Frequency requirements. CTS 6.17.d requires each train of CREVS to be tested at least once per 36 months (one train each 18 months). ITS 5.5.16.d requires a train of CREVS to be tested each 18 months, alternating between the trains each interval. Therefore, the CTS and ITS testing Frequencies are the same. This change is designated as administrative because it does not result in technical changes to the CTS.

A11 SQN Unit 2 CTS 6.8.4.k.b.2, Steam Generator (SG) Program - Accident Induced Leakage Performance Criterion, states, in part, that the accident-induced leakage is not to exceed 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs. Both Unit 1 and Unit 2 CTS 6.8.4.k.b.3 contain criterion for operational leakage referencing the CTS 3.4.6.2 criterion of a maximum primary to secondary leakage of 150 gallons per day (gpd) through any one steam generator. ITS 5.5.7.b.2, Steam Generator (SG)

Program - Accident Induced Leakage Performance Criterion, states, in part, that Leakage is not to exceed 1 gpm per SG while ITS 5.5.7.b.3 states that the operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE," 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG). This changes the CTS by removing the duplicative non-faulted SG leakage criterion.

The purpose of CTS 6.8.4.k.b is to provide provisions for performance criteria for SG tube integrity. The provisions are provided such that SG tube integrity is maintained by meeting performance criteria for tube structural integrity, accident induced leakage, and operational leakage. CTS 6.8.4.k.b.2 for Unit 1 provides one criterion for accident induced leakage, 1 gpm for the faulted SG; whereas CTS 6.8.4.k.b.2 for Unit 2 provides two criterion for accident induced leakage, 1.0 gpm for the faulted SG and 0.1 gpm for each of the non-faulted SGs. Both Unit 1 and Unit 2 CTS 6.8.4.k.b.3 provide an operational leakage performance criterion is specified in Limiting Condition for Operation (LCO) 3.4.6.2, Reactor Coolant System, Operational Leakage. The CTS 3.4.6.2 requirement for the maximum primary to secondary leakage is 150 gallons per day (gpd)

(0.1 gpm) through any one steam generator. The Unit 2 CTS 6.8.4.k.b.2 criterion of 0.1 gpm for each of the non-faulted SGs is duplicated in the Unit 2 CTS 6.8.4.k.b.3 criterion of the operational leakage performance criterion as reference to Limiting Condition for Operation (LCO) 3.4.6.2, Reactor Coolant System, Operational Leakage, of 150 gpd through any one SG. This change is acceptable because duplicative leakage criterion from the Unit 2 CTS is being removed while the leakage criterion is being maintained.

This change is designated as administrative because it does not result in a technical change to the CTS.

A12 CTS 6.16 c requires the total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A. ITS 5.5.11 c requires that the total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days but does not include a specific test method. TVA is proposing to change the test method for determining total particulate concentration for SQN to ASTM D6217-Sequoyah Unit 1 and Unit 2 Page 4 of 12 Enclosure 2, Volume 16, Rev. 0, Page 139 of 270

Enclosure 2, Volume 16, Rev. 0, Page 140 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS

11. This changes the CTS by requiring the testing of fuel oil total particulate concentration to be in accordance with ASTM D6217-11.

The purpose of CTS 6.16 c is to provide the requirements for testing of total particulate concentration of the fuel oil. Regulatory Guide 1.137, Revision 2, "Fuel Oil Systems for Emergency Power Supplies," describes methods that the NRC considers acceptable for use in complying with the NRC requirements regarding fuel oil systems. Based on the guidance of Regulatory Guide 1.137 ANSI/ANS-59.51 to ANSI/ASTM D2276-94 for manual sampling of the stored fuel should be changed to ASTM D6217-11. This change is acceptable because testing of total particulate concentration of the fuel oil will be done in accordance with the approved NRC method of ASTM D6217-11. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.14.1.2 states, in part, that the ODCM becomes effective after review and acceptance by the process described in TVA-NQA-PLN89-A. ITS 5.5.1.c.2 states, in part, that the ODCM becomes effective after review and acceptance by the plant manager. This changes the CTS by requiring the plant manager approval for the ODCM.

The purpose of CTS 6.14.1.2 is to ensure that the ODCM has been properly reviewed by the process described in TVA-NQA-PLN89-A. ITS 5.5.1 still requires that the review process described in TVA-NQA-PLN89-A is performed (see DOC LA04 for the exclusion of the process described in TVA-NQA-PLN89-A from the ITS 5.5.1), but also includes an additional acceptance that the plant manager must review and approve the ODCM. This change is designated as more restrictive since a higher level of approval is required in the ITS than was required in the CTS.

M02 The CTS does not include program requirements for the Safety Function Determination Program. The ITS includes a program for the Safety Function Determination Program.

This change the CTS by adding the Safety Function Determination Program (SFDP).

The Safety Function Determination Program is included to support implementation of the support system OPERABILITY characteristics of the Technical Specifications. The specific wording associated with this program is found in ITS 5.5.13. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

M03 The CTS does not include a requirement for the Battery Monitoring and Maintenance Program. The ITS includes a requirement for this program. This changes the CTS by adding the ITS 5.5.15, "Battery Monitoring and Maintenance Program."

The Battery Monitoring and Maintenance Program is included to provide for battery restoration and maintenance. The specific wording associated with this program may be found in ITS 5.5.15. This change is acceptable because it supports implementation of the requirements of the ITS. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

Sequoyah Unit 1 and Unit 2 Page 5 of 12 Enclosure 2, Volume 16, Rev. 0, Page 140 of 270

Enclosure 2, Volume 16, Rev. 0, Page 141 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS M04 The CTS does not have a Surveillance Frequency Control Program. ITS 5.5.17 requires a program to satisfy the relocation of the Surveillance Frequency from the individual specifications. This changes the CTS by incorporating the requirements of ITS 5.5.17.

The NRC has been reviewing and granting improvements to the Improved Standard Technical Specifications (ISTS) based, at least in part, on probabilistic risk analysis insights. Typically, the proposed improvements involved a relaxation of one or more Completion Times or Surveillance Frequencies in the TS. In August 1995, the NRC adopted a final policy statement on the use of probabilistic risk assessment (PRA) methods, which included the following regarding the expanded use of PRA.

• The use of PRA technology should be increased in all regulatory matters to the extent supported by the state of the art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

• PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state of the art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, licensee commitments, and staff practices. Where appropriate, PRA should be used to support the proposal of additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule).

Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

• PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

• The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on need for proposing and backfitting new generic requirements on nuclear power plant licensees.

In its approval of the policy statement, the Commission articulated its expectation that implementation of the policy statement will improve the regulatory process in three areas: foremost, through safety decision-making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees. This change is consistent with TSTF-425-A. TSTF- 425-A required that licensees who adopted this TSTF confirm that the plant PRA is consistent with Section 4.2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment results for Risk-Informed Activities." SQN has performed an assessment on the Sequoyah Units 1 and 2 PRA, and confirmed that it is consistent with the guidance in Section 4.2 of Regulatory Guide 1.200 (See Enclosure 10). Future model updates (internal model or external model) will be evaluated to determine any impact on the conclusions of the assessment that was performed in support of adopting this change. For each individual Surveillance Frequency relocation, see each of the associated Technical Specifications for the Sequoyah Unit 1 and Unit 2 Page 6 of 12 Enclosure 2, Volume 16, Rev. 0, Page 141 of 270

Enclosure 2, Volume 16, Rev. 0, Page 142 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Discussion of Changes (DOC) justifying the individual relocations. This change is considered more restrictive since a new program is being added to the Technical Specifications.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAP, CLRT Program, IST Program, ISI Program, or Surveillance Frequency Control Program) CTS 4.0.5 provides requirements for the Inservice Inspection Program. The ITS does not include Inservice Inspection Program requirements. In addition, since the Inservice Testing Program is the only requirement remaining, the reference to ASME Code Class 1, 2, and 3 "components" has been changed to "pumps and valves" for clarity. Pumps and valves are the only components related to the Inservice Testing Program (as described in CTS 4.0.5). This changes the CTS by moving these requirements from the Technical Specifications to the Inservice Inspection (ISI) Program.

The removal of these requirements is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The Technical Specifications still retain requirements for the affected components to be OPERABLE. Also, this change is acceptable because these requirements will be adequately controlled by the ISI, which is required by 10 CFR 50.55a. Compliance with 10 CFR 50.55a is required by the SQN Units 1 and 2 Operating Licenses. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications.

LA02 (Type 3 - Removing Procedural Details for meeting TS Requirements or Reporting Requirements) CTS 4.6.1.8.b.2, CTS 4.6.1.8.c, CTS 4.7.7.c.2, CTS 4.7.7.d, CTS 4.7.8.b.2, CTS 4.7.8.c, CTS 4.9.12.b.2, CTS 4.9.12.c require that within 31 days after removal of a carbon sample the laboratory analysis results are shown to be within limit.

ITS 5.5.9.c requires the same analysis to be performed however the detail of "within 31 days after removal of a carbon sample" is not included. This changes the CTS by moving these procedural details from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to perform the testing at the appropriate Frequencies. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

Sequoyah Unit 1 and Unit 2 Page 7 of 12 Enclosure 2, Volume 16, Rev. 0, Page 142 of 270

Enclosure 2, Volume 16, Rev. 0, Page 143 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS LA03 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.11.1.4 includes the details for implementing the requirements for the liquid holdup tank. CTS 3.11.2.5 includes the details for implementing the requirements for the explosive gas mixture. CTS 3.11.2.6 includes the details for implementing the requirements for the gas decay tanks. The details for implementing these requirements, including the specific limits, are not included in the ITS.

CTS 3.11.2.6 Bases requires, in part in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest exclusion boundary will not exceed 0.5 rem. ITS 5.5.10.b requires the curie content in the gas decay tank to be less than the amount that would result in whole body exposure of greater than or equal to 0.5 rem at the exclusion boundary. This changes the CTS by moving the boundary exposure limit of 0.5 rem from the Bases to ITS 5.5.10 and moving those procedural details for implementing the requirements, including the specific limits, from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details for the specific limits, Applicability, Actions, and Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.10 still retains the requirement to include a program, which provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in condensate storage tank, steam generator layup tank and outdoor temporary liquid radwaste storage tanks. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA04 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, NQAP, CLRT Program, IST Program, ISI Program) CTS 6.14.1.2 requires changes to the ODCM to be effective after review and acceptance by the process described in TVA-NQA-PLN89-A. ITS 5.5.1.b requires changes to the ODCM to become effective after the approval of the plant manager. This changes the CTS by moving the process described in TVA-NQA-PLN89-A to the Nuclear Facility Quality Assurance Program Description (NFQAPD). DOC M01 describes the addition of the plant manager approval.

The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is acceptable because these types of procedural details will be adequately controlled in the NFQAPD. Any changes to the NFQAPD are made under 10 CFR 50.54(a), which ensure changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specifications requirements are being removed from the Technical Specifications.

LA05 (Type 3 - Removing Procedural Details for meeting TS Requirements or Reporting Requirements) CTS Table 4.8.1a and Table 4.8.2 Unit 1 footnote (c) and CTS Table 4.8-1a and Table 4.8-2 Unit 2 footnote (c) states, in part the float voltage of 2.13 volts Sequoyah Unit 1 and Unit 2 Page 8 of 12 Enclosure 2, Volume 16, Rev. 0, Page 143 of 270

Enclosure 2, Volume 16, Rev. 0, Page 144 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS is corrected for average electrolyte temperature. ITS 5.5.15 b.1 requires a program with actions to restore battery cells with float voltage < 2.13 V and ITS 5.5.15 b.2 requires a program with actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cells has been found to be < 2.13 V.

This changes the CTS by by moving information from the specification to the Battery Monitoring and Maintenance Program implementing document.

The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.15 still retains the requirement for float voltage 2.13 V. Also, this change is acceptable because these types of procedural details will be adequately controlled by the requirements of a program required by ITS Chapter 5. ITS 5.5.15, Battery Monitoring and Maintenance program is controlled by Chapter 5 of the Technical Specifications. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA06 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 6.16 c requires total particulate of the fuel oil to be less than or equal to 10 mg/l when tested every 31 days in accordance with ASTM D-2276, Method A. ITS 5.5.11.c requires the total particulate concentration of the fuel oil is less than or equal to 10 mg/l when tested every 31 days. This changes the CTS by moving the details of using of particulate testing standard ASTM D-2276, Method A from the CTS to the ITS SR 3.8.3.3 Bases.

The removal of these details related to testing standards from the Technical Specification is acceptable, because this type of information is not necessary to be included in the Technical Specification to provide adequate protection of the public health and safety. The ITS retains the requirement for fuel oil particulate testing every 31 days. Also, this change is acceptable because the removed details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to the Bases to ensure the Bases are properly controlled.

This change is designated as less restrictive removal of detail change, because information relating to testing standards is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 7 - Relaxation of Surveillance Frequency) CTS 6.8.4.k.d.2 states, "Inspect 100% of the tubes at sequential periods of 144, 108, 72 and thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected." ITS 5.5.7.d.2 states, "After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of Sequoyah Unit 1 and Unit 2 Page 9 of 12 Enclosure 2, Volume 16, Rev. 0, Page 144 of 270

Enclosure 2, Volume 16, Rev. 0, Page 145 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage." ITS 5.5.7.d.2 goes on to describe the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d by stating, "a)

After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b)

During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d)

During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods." This changes the CTS by relaxing the surveillance frequency for inspecting SG tubes.

The purpose of CTS 6.8.4.k is to ensure that SG tube integrity is maintained by providing provisions regarding the scope, frequency, and methods of SG tube inspections. These changes to when inspections are performed are considered marginal increases for consistency with typical fuel cycle lengths that better accommodate the scheduling of inspections and reflect the improved resistance of alloy 690 TT SG tubes to stress corrosion cracking. This change is acceptable because TVA has reviewed TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," (ADAMS Accession No. ML110610350) and the model safety evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513) as part of the Federal Register Notice for Comment. As described in the subsequent paragraphs, TVA has concluded that the justifications presented in TSTF-510 and the model safety evaluation prepared by the NRC staff are applicable to SQN Unit 1 and Unit 2 and justify the incorporation of the changes to the SQN Unit 1 and SQN Unit 2 ITS. TVA is proposing the following variations from the TS changes described in the TSTF-510, Revision 2, or the applicable parts of the NRC staff's model safety evaluation dated October 19, 2011. SQN Unit 1 and Unit 2 ITS utilize different numbering than the Standard Technical Specifications (NUREG 1431, Revision 4.0, "Standard Technical Specifications Westinghouse Plants") on which TSTF-510 was based. The specific numbering differences are: 1) TSTF-510 Rev. 2, TS 3.4.20, "Steam Generator Tube Integrity," is ITS 3.4.17, "Steam Generator Tube Integrity"; 2) TSTF-510 Rev. 2 TS 5.5.9, "Steam Generator (SG) Program is ITS 5.5.7, "Steam Generator (SG) Program"; and 3)

TSTF-510 Rev. 2 TS 5.6.7, "Steam Generator Tube Inspection Report," is ITS 5.6.6, "Steam Generator Tube Inspection Report." This change is designated as less restrictive because the SG tube inspections will be performed less frequently in ITS than they were in CTS.

Sequoyah Unit 1 and Unit 2 Page 10 of 12 Enclosure 2, Volume 16, Rev. 0, Page 145 of 270

Enclosure 2, Volume 16, Rev. 0, Page 146 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS L02 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria)

CTS 6.16.a.3 requires performance of the "clear and bright" test, used to establish the acceptability of new fuel oil for use prior to addition to storage tanks. ITS 5.5.11.a.3 requires a determination that the fuel oil has a clear and bright appearance with proper color or that water and sediment content is within limits. This changes the CTS by allowing a "water and sediment content" test to be performed to establish the acceptability of new fuel oil instead of only allowing a "clear and bright" test.

CTS 6.16.a.3 requires performance of the "clear and bright" test, to establish the acceptability of new fuel oil for use prior to addition to storage tanks. ITS 5.5.11.a.3 is proposed to be expanded to allow a water and sediment content test to be performed to establish the acceptability of new fuel oil instead of the "clear and bright" test. ASTM D4176-86,"Standard Test Method for Free Water and Particulate Contamination in Distillate Fuels (Clear and Bright Pass/Fail Procedures)," verifies that the new fuel oil has a clear and bright appearance with proper color. The "clear and bright" test is only applicable to fuel oils that meet the ASTM D4176 color rating requirements (i.e., an ASTM D1500, "Test Method for ASTM Color of Petroleum Products (ASTM Color Scale)," color rating of five or less). The clear and bright test is a qualitative test for determining free water and particulate contamination in distillate fuels and is, therefore, subject to human interpretation. For example, if an attempt is made to use the qualitative "clear and bright" test with darker colored fuels (e.g., for high sulfur fuel oil that has been dyed in accordance with EPA mandated requirements), the presence of free water or particulate could be obscured and missed by the viewer. Therefore, ITS 5.5.11.a.3 has been expanded to allow a water and sediment content test. The water and sediment content test is a quantitative test using centrifuge methods. In ASTM D975-90, ASTM D1796, Standard Method for Water and Sediment in Fuel Oils by the Centrifuge Method (Laboratory Procedure), is an acceptable standard for the water and sediment content test. In addition, the use of ASTM D1796-83 was endorsed by the NRC in Amendment No. 101 for the Wolf Creek Generating Station. ASTM D1796-83 is the same ASTM Standard used to verify the water and sediment content is within limits within 31 days following sampling and addition to the storage tanks as required by CTS 6.16.b. Therefore, since ASTM D1796 is currently used to verify the acceptability of new fuel oil for use after addition to the storage tanks, the use of these quantitative methods (i.e., water and sediment content) in lieu of ASTM D4176 (i.e., "clear and bright" test) does not introduce a different method for determining the acceptability of new fuel oil.

This change is designated as less restrictive because Surveillance acceptance criteria required in the CTS will have alternative acceptance criteria allowed in ITS.

L03 (Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria) CTS 6.8.4.h specifies the limit for peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 12.0 psig. Also, CTS 6.8.4.h specifies, in part, that bypass leakage paths to the auxiliary building from the isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (13.2 psig). CTS 6.8.4.h.d.2 requires penetrations not individually testable to have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (12 psig) during each Type A test. ITS 5.5.14.a specifies, in part, that bypass leakage paths to the auxiliary building from the isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined Sequoyah Unit 1 and Unit 2 Page 11 of 12 Enclosure 2, Volume 16, Rev. 0, Page 146 of 270

Enclosure 2, Volume 16, Rev. 0, Page 147 of 270 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (12.46 psig). ITS 5.5.14.b specifies the calculated peak containment internal pressure for the design basis loss of coolant accident, Pa is 11.33 psig and the containment design pressure is 12.0 psig. ITS 5.5.14.d.4.b requires penetrations not individually testable to have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (11.33 psig) during each Type A test. This changes the CTS by reducing the calculated peak containment internal pressure for the design basis loss of coolant accident, Pa to 11.33 psig and specifying the containment design pressure is 12.0 psig.

The purpose of ITS 5.5.14 is to ensure the appropriate limits are specified for the Containment Leakage Rate Testing Program. This change is acceptable because the acceptable limits continue to ensure the containment leakage is within the value assumed in the accident analysis as described in the recent application to modify Ice Condenser Technical Specifications (ML13199A281). Currently, SQN is using the containment design pressure value of 12.0 psig as Pa. In the ITS, the value of Pa (11.33 psig) is the calculated peak containment internal pressure for the design basis loss of coolant accident. This is acceptable because the value of Pa (11.33 psig) is the value assumed in the accident analyses. This change is designated as less restrictive because a lower pressure will be used for Pa in the Containment Leakage Rate Testing Program.

Sequoyah Unit 1 and Unit 2 Page 12 of 12 Enclosure 2, Volume 16, Rev. 0, Page 147 of 270

Enclosure 2, Volume 16, Rev. 0, Page 148 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 148 of 270

Enclosure 2, Volume 16, Rev. 0, Page 149 of 270 CTS Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

1.18 5.5.1 Offsite Dose Calculation Manual (ODCM) 1.18 a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program, and 1 1.18 b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.1] and Specification [5.6.2]. 2 6.14.1.1 Licensee initiated changes to the ODCM:

6.14.1.1 a. Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

6.14.1.1.a

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s) and 6.14.1.1.b 2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations, 6.14.1.2 b. Shall become effective after the approval of the plant manager, and 6.14.1.3 c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-1 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 149 of 270

Enclosure 2, Volume 16, Rev. 0, Page 150 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.a 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include

[Recirculation Spray, Safety Injection, Chemical and Volume Control, gas stripper, and Hydrogen Recombiner]. The program shall include the following: 2 Residual Heat Removal, Containment Spray, and RCS Sampling 6.8.4.a a. Preventive maintenance and periodic visual inspection requirements and 6.8.4.b b. Integrated leak test requirements for each system at least once per

[18] months.

The provisions of SR 3.0.2 are applicable.

5.5.3 [ Post Accident Sampling

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REVIEWER'S NOTE----------------------------------------

This program may be eliminated based on the implementation of WCAP-14986, Rev. 1, "Post Accident Sampling System Requirements: A Technical Basis," and the associated NRC Safety Evaluation dated June 14, 2000, and implementation of the following commitments:

1. [Licensee] has developed contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere. The contingency plans will be contained in emergency plan implementing procedures and implemented with the implementation of the License amendment. Establishment of contingency plans is considered a regulatory commitment.

2. The capability for classifying fuel damage events at the Alert level threshold 4 has been established for [Plant] at radioactivity levels of 300 mCi/cc dose equivalent iodine. This capability may utilize the normal sampling system and/or correlations of sampling or letdown line dose rates to coolant concentrations. This capability will be described in emergency plan implementing procedures and implemented with the implementation of the License amendment. The capability for classifying fuel damage events is considered a regulatory commitment.

3. [Licensee] has established the capability to monitor radioactive iodines that have been released to offsite environs. This capability is described in our emergency plan implementing procedures. The capability to monitor radioactive iodines is considered a regulatory commitment.

This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-2 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 150 of 270

Enclosure 2, Volume 16, Rev. 0, Page 151 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Post Accident Sampling (continued) and containment atmosphere samples under accident conditions. The program shall include the following:

4

a. Training of personnel,

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analysis equipment. ]

4 6.8.4.f 5.5.4 Radioactive Effluent Controls Program 3

6.8.4.f This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

6.8.4.f.1) a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 6.8.4.f.2) b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, 6.8.4.f.3) c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, 6.8.4.f.4) d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I, 6.8.4.f.5) e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days, 6.8.4.f.6) f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I, SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-3 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 151 of 270

Enclosure 2, Volume 16, Rev. 0, Page 152 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.f 5.5.4 Radioactive Effluent Controls Program (continued) 4 3

6.8.4.f.7) g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

6.8.4.f.7)).1 1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin and 6.8.4.f.7)).2 2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ, 6.8.4.f.8) h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, 6.8.4.f.9) i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, and 6.8.4.f.10) j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

6.8.4.f The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

6.8.4.l 5.5.5 Component Cyclic or Transient Limit 4 U 5.2.1 4

This program provides controls to track the FSAR, Section [ ], cyclic and 2 transient occurrences to ensure that components are maintained within the design limits.

5.5.6 [ Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon 4

Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC.

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies. ]

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-4 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 152 of 270

Enclosure 2, Volume 16, Rev. 0, Page 153 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4.0.5.c 5.5.7 Reactor Coolant Pump Flywheel Inspection Program 4 5

This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

ultrasonic In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over 3 magnetic the volume from the inner bore of the flywheel to the circle one-half of the outer particle radius or a surface examination (MT and/or PT) of exposed surfaces of the penetrant liquid 3 removed flywheels may be conducted at 20 year intervals.

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REVIEWER'S NOTE----------------------------------------

The inspection interval and scope for RCP flywheels stated above can be applied to plants that satisfy the requirements in WCAP-15666, "Extension of Reactor 5 Coolant Pump Motor Flywheel Examination."

4.0.5 5.5.8 Inservice Testing Program 4 6

4.0.5 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following: 6 pumps and valves 4.0.5.b a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice testing inservice testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days 4.0.5.c b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities, 4.0.5.d c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-5 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 153 of 270

Enclosure 2, Volume 16, Rev. 0, Page 154 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4.0.5 5.5.8 Inservice Testing Program (continued) 4 6

4.0.5.e d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

6.8.4.k 5.5.9 Steam Generator (SG) Program 4 7

A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

6.8.4.k.a a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging [or repair] of 2 tubes. Condition monitoring assessments shall be conducted during each or outage during which the SG tubes are inspected, plugged, [or repaired] to 1 2 confirm that the performance criteria are being met.

6.8.4.k.b b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

6.8.4.k.b.1 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients TSTF-

), 510-A included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

6.8.4.k.b.2

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-6 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 154 of 270

Enclosure 2, Volume 16, Rev. 0, Page 155 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) 4 7

exceed [1 gpm] per SG [, except for specific types of degradation at 2

specific locations as described in paragraph c of the Steam Generator TSTF-Program. 510-A

]

6.8.4.k.b.3

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

]

TSTF-6.8.4.k.c c. Provisions for SG tube repair criteria. Tubes found by inservice inspection 510-A plugging [or to contain flaws with a depth equal to or exceeding [40%] of the nominal 2 tube wall thickness shall be plugged [or repaired].

]

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REVIEWER'S NOTE----------------------------------------

Alternate tube repair criteria currently permitted by plant technical specifications are plugging [or listed here. The description of these alternate tube repair criteria should be ]

TSTF-equivalent to the descriptions in current technical specifications and should also 510-A plugging [or include any allowed accident induced leakage rates for specific types of degradation at specific locations associated with tube repair criteria. ] 5

]

plugging [or

[The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. . . .]

6.8.4.k.d d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the plugging [or tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may TSTF-satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not 2 510-A

]

part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals assessment shall be such as to ensure that SG tube integrity is maintained until the next TSTF-SG inspection. An assessment of degradation shall be performed to 7 510-A determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-7 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 155 of 270

Enclosure 2, Volume 16, Rev. 0, Page 156 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) 4 7

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REVIEWER'S NOTE----------------------------------------

Plants are to include the appropriate Frequency (e.g., select the appropriate Item 2.) for their SG design. The first Item 2 is applicable to SGs with Alloy 600 mill annealed tubing. The second Item 2 is applicable to SGs with Alloy 600 5 thermally treated tubing. The third Item 2 is applicable to SGs with Alloy 690 thermally treated tubing.

6.8.4.k.d.1 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

installation

[2. Inspect 100% of the tubes at sequential periods of 60 effective full INSERT 1 power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.]

[2. Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period INSERT 2 shall be considered to begin after the first inservice inspection of the TSTF-SGs. In addition, inspect 50% of the tubes by the refueling outage 510-A nearest the midpoint of the period and the remaining 50% by the 2

refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.]

6.8.4.k.d.2 [2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, INSERT 3 thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.]

6.8.4.k.d.3 3. If crack indications are found in any SG tube, then the next inspection affected and for each SG for the degradation mechanism that caused the crack potentially affected TSTF-indication shall not exceed 24 effective full power months or one 510-A results in more refueling outage (whichever is less). If definitive information, such as frequent inspections from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

6.8.4.k.e e. Provisions for monitoring operational primary to secondary LEAKAGE.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-8 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 156 of 270

Enclosure 2, Volume 16, Rev. 0, Page 157 of 270 CTS 5.5 2

INSERT 1 After the first refueling outage following SG installation, inspect each steam generator at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

2 INSERT 2 TSTF-After the first refueling outage following SG installation, inspect each SG at least every 48 510-A effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

---

Reviewer's Note ------------------------------------------------------

A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; Insert Page 5.5-8 Enclosure 2, Volume 16, Rev. 0, Page 157 of 270

Enclosure 2, Volume 16, Rev. 0, Page 158 of 270 CTS 5.5 b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.

INSERT 3 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations 2 inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the TSTF-510-A determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

---

Reviewer's Note ------------------------------------------------------

A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

Insert Page 5.5-8 Enclosure 2, Volume 16, Rev. 0, Page 158 of 270

Enclosure 2, Volume 16, Rev. 0, Page 159 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) 4 7

[f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the 2 purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

---

REVIEWER'S NOTE----------------------------------------

Tube repair methods currently permitted by plant technical specifications are to be listed here. The description of these tube repair methods should be 5 equivalent to the descriptions in current technical specifications. If there are no approved tube repair methods, this section should not be used.

1. . . .] 2 4

6.8.4.c 5.5.10 Secondary Water Chemistry Program 8

6.8.4.c This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion 8 cracking. The program shall include:

6.8.4.c.(i) a. Identification of a sampling schedule for the critical variables and control points for these variables, 6.8.4.c.(ii) b. Identification of the procedures used to measure the values of the critical variables, 6.8.4.c.(iii) c. Identification of process sampling points, which shall include monitoring the 11 discharge of the condensate pumps for evidence of condenser in leakage, 6.8.4.c.(iv) d. Procedures for the recording and management of data, 6.8.4.c.(v) e. Procedures defining corrective actions for all off control point chemistry conditions, and 6.8.4.c.(vi) f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-9 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 159 of 270

Enclosure 2, Volume 16, Rev. 0, Page 160 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) Regulatory Positions C.5.a, C.5.c, 4

C.5.d and C.6.b of 3 9

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, 2 Revision 2, ASME N510-1989, and AG-1]. removal efficiently of 99.95% of dioctyl phthalate (DOP) 3 INSERT 4 ANSI N510-1975 ASTM D3803-1989 4.6.1.8.e a. Demonstrate for each of the ESF systems that an inplace test of the high 4.7.7.c.3 4.7.7.f 4.6.1.8.b.1 efficiency particulate air (HEPA) filters shows a penetration and system 4.7.8.b.3 4.7.7.c.1 4.7.8.b.1 bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, 2 4.7.8.e 4.9.12.b.3 4.9.12.b.1 Revision 2, and ASME N510-1989] at the system flowrate specified below 4.9.12.e [+/- 10%]. ANSI N510-1975 (except for the provisions of Sections 8 and 9) Regulatory Positions C.5.a and C.5.c of 3 2 ESF Ventilation System Flowrate INSERT 5 2

[ ] [ ]

removal efficiently of 99.95% of a halogenated 3 hydrocarbon refrigerant test gas 4.6.1.8.f b. Demonstrate for each of the ESF systems that an inplace test of the 4.7.7.c.3 4.7.7.g charcoal adsorber shows a penetration and system bypass < [0.05]% when 4.7.8.b.3 tested in accordance with [Regulatory Guide 1.52, Revision 2, and 2 4.7.8.f 4.9.12.b.3 ASME N510-1989] at the system flowrate specified below [+/- 10%]. 2 4.9.12.f ANSI N510-1975 (except for the provisions of Sections 8 and 9) Regulatory Positions C.5.a and C.5.d of 3 ESF Ventilation System Flowrate INSERT 6 3

[ ] [ ]

Regulatory Position C.6.b of 3 4.6.1.8.b.2 c. Demonstrate for each of the ESF systems that a laboratory test of a sample 2 4.6.1.8.c 4.7.7.c..2 of the charcoal adsorber, when obtained as described in [Regulatory < 2.5%

4.7.7.d Guide 1.52, Revision 2], shows the methyl iodide penetration less than the 3 4.7.8.b.2 4.7.8.c value specified below when tested in accordance with ASTM D3803-1989 at 4.9.12.b.2 a temperature of 30°C (86°F) and the relative humidity specified below. 3 4.9.12.c of 70%

ESF Ventilation System Penetration RH Face Velocity (fps)

[ ] [See Reviewer's [See [See Reviewer's EGTS Note] Reviewer's Note] 2 ABGTS Note] 3 CREVS

---

REVIEWER'S NOTE----------------------------------------

The use of any standard other than ASTM D3803-1989 to test the charcoal sample may result in an overestimation of the capability of the charcoal to adsorb 5 radioiodine. As a result, the ability of the charcoal filters to perform in a manner consistent with the licensing basis for the facility is indeterminate.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-10 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 160 of 270

Enclosure 2, Volume 16, Rev. 0, Page 161 of 270 CTS 5.5 4.6.1.8.b 3 INSERT 4 4.6.1.8.e 4.6.1.8.f 4.7.7.f The test described in Specification 5.5.9.a and 5.5.9.b shall be performed once per 18 months; 4.7.7.g after any structural maintenance on the high efficiency particulate air (HEPA) filter bank or 4.7.8.e 4.7.8.f charcoal adsorber bank housing; following painting, fire, or chemical release in any ventilation zone communicating with the system; and after each complete or partial replacement of a HEPA 4.6.1.8.c filter bank or charcoal adsorber bank.

4.6.1.8.b 4.7.7.c 4.7.7.d The test described in Specification 5.5.9.c shall be performed once per 18 months or after 720 4.7.8.b 4.7.8.c hours of filter operation; after any structural maintenance on the HEPA filter bank or charcoal 4.9.12.b adsorber bank housing; and following painting, fire, or chemical release in any ventilation zone 4.9.12.c communicating with the system.

4.6.1.8.d 4.7.7.e The test described in Specification 5.5.9.d and 5.5.9.e shall be performed once per 18 months.

4.7.8.d 4.9.12.d 3 INSERT 5 ESF Ventilation System Flow Rate (cfm)

Emergency Gas Treatment System (EGTS) 4000 Auxiliary Building Gas Treatment System (ABGTS) 9000 Control Room Emergency Ventilation System (CREVS) 4000 3

INSERT 6 ESF Ventilation System Flow Rate (cfm)

EGTS 4000 ABGTS 9000 CREVS 4000 Insert Page 5.5-10 Enclosure 2, Volume 16, Rev. 0, Page 161 of 270

Enclosure 2, Volume 16, Rev. 0, Page 162 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (continued) 4 9

ASTM D 3803-1989 is a more stringent testing standard because it does not differentiate between used and new charcoal, it has a longer equilibration period performed at a temperature of 30°C (86°F) and a relative humidity (RH) of 95%

(or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test.

Allowable Penetration = [(100% - Methyl Iodide Efficiency

• for Charcoal Credited in Licensee's Accident Analysis) / Safety Factor]

When ASTM D3803-1989 is used with 30°C (86°F) and 95% RH (or 70% RH with humidity control) is used, the staff will accept the following:

5 Safety factor 2 for systems with or without humidity control.

Humidity control can be provided by heaters or an NRC-approved analysis that demonstrates that the air entering the charcoal will be maintained less than or equal to 70 percent RH under worst-case design-basis conditions.

If the system has a face velocity greater than 110 percent of 0.203 m/s (40 ft/min), the face velocity should be specified.

• This value should be the efficiency that was incorporated in the licensee's accident analysis which was reviewed and approved by the staff in a safety evaluation.

4.6.1.8.d.1 d. Demonstrate for each of the ESF systems that the pressure drop across the 4.7.7.e.1 4.7.8.b.3 combined HEPA filters, the prefilters, and the charcoal adsorbers is less 3 4.7.8.d.1 than the value specified below when tested in accordance with [Regulatory 2 4.9.12.b.3 4.9.12.d.1 Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate 4.7.7.c.3 specified below [+/- 10%]. 1975 ESF Ventilation System Delta P Flowrate INSERT 7 3

[ ] [ ] [ ]

4.7.8.d.4 [ e. Demonstrate that the heaters for each of the ESF systems dissipate the 4.9.12.d.3 value specified below [+/- 10%] when tested in accordance with

[ASME N510-1989].

ANSI N510-1975 2

ESF Ventilation System Wattage ]

[ ] [ ]

Auxiliary Building Gas Treatment System 32 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-11 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 162 of 270

Enclosure 2, Volume 16, Rev. 0, Page 163 of 270 CTS 5.5 3 INSERT 7 ESF Ventilation System Combined Flowrate (cfm)

Delta P (inches water gauge)

EGTS 5 4000 ABGTS 3 9000 CREVS 3 4000 Insert Page 5.5-11 Enclosure 2, Volume 16, Rev. 0, Page 163 of 270

Enclosure 2, Volume 16, Rev. 0, Page 164 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 3.11.1.4 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program 4 3.11.2.5 3.11.2.6 10 This program provides controls for potentially explosive gas mixtures contained decay in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of temporary 2 the Condensate Storage Tank, radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous Steam Generator radioactivity quantities shall be determined following the methodology in [Branch Layup Tank, and Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be 9 determined in accordance with [Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures"].

The program shall include:

4.11.2.5 a. The limits for concentrations of hydrogen and oxygen in the [Waste Gas 2 Holdup System] and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion), 11 4.11.2.6 b. A surveillance program to ensure that the quantity of radioactivity contained decay in [each gas storage tank and fed into the offgas treatment system] is less 2 than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of [an uncontrolled 2

release of the tanks' contents], and , Condensate Storage Tank, and

Steam Generator Layup Tank 11 4.11.1.4 temporary

c. A surveillance program to ensure that the quantity of radioactivity contained storage in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank 3 overflows and surrounding area drains connected to the [Liquid Radwaste Treatment System] is less than the amount that would result in exceeding concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-12 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 164 of 270

Enclosure 2, Volume 16, Rev. 0, Page 165 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.16 5.5.13 Diesel Fuel Oil Testing Program 4 11 6.16 A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

6.16.a a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

6.16.a.1 1. An API gravity or an absolute specific gravity within limits, 6.16.a.2

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and 6.16.a.3 3. A clear and bright appearance with proper color or a water and sediment content within limits.

6.16.b b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a.,

above, are within limits for ASTM 2D fuel oil, and 6.16.c

c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

6.8.4.j 5.5.14 Technical Specifications (TS) Bases Control Program 4 12 6.8.4.j This program provides a means for processing changes to the Bases of these Technical Specifications.

6.8.4.j.a a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

6.8.4.j.b b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

6.8.4.j.b.1 1. A change in the TS incorporated in the license or 6.8.4.j.b.2 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-13 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 165 of 270

Enclosure 2, Volume 16, Rev. 0, Page 166 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.j.d 5.5.14 Technical Specifications (TS) Bases Control Program (continued) 4 12 6.8.4.j.d c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR. 3 U 12 6.8.4.j.d d. Proposed changes that meet the criteria of Specification 5.5.14b above shall 4 be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

DOC M02 5.5.15 Safety Function Determination Program (SFDP) 4 13 This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected,

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists,

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities, and

d. Other appropriate limitations and remedial or compensatory actions.

11 A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable, or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable, or SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-14 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 166 of 270

Enclosure 2, Volume 16, Rev. 0, Page 167 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC M02 5.5.15 Safety Function Determination Program (SFDP) (continued) 4 13

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5. 16 Containment Leakage Rate Testing Program 4 14

[OPTION A]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

b. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

c. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests. 10

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

d. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

e. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-15 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 167 of 270

Enclosure 2, Volume 16, Rev. 0, Page 168 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.h 5.5.16 Containment Leakage Rate Testing Program (continued) 4 14 10

[OPTION B]

6.8.4.h a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

INSERT 8

2. The visual examination of the steel liner plate inside containment 3 intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

[ 3. ...]

6.8.4.h b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is [45 psig]. The containment design pressure is 2

[50 psig]. 11.33 12 6.8.4.h c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of 2 containment air weight per day.

0.25 6.8.4.h d. Leakage rate acceptance criteria are:

6.8.4.h.a 1. Containment leakage rate acceptance criterion is 1.0 La. During the 3 first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C 3 tests and 0.75 La for Type A tests.

6.8.4.h.b 2. Air lock testing acceptance criteria are:

6.8.4.h.b.1 a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

2 6.8.4.h.b.2 b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig]. 3 6 for at least two minutes 3

INSERT 9 Amendment XXX SEQUOYAH UNIT 1 Westinghouse STS 5.5-16 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 168 of 270

Enclosure 2, Volume 16, Rev. 0, Page 169 of 270 CTS 5.5 3 INSERT 8 6.8.4.h 1. Bypass leakage paths to the auxiliary building leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (12.46 psig) and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

3 INSERT 9 6.8.4.h.c 3. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate to 0.05 La.

6.8.4.h.d 4. Bypass leakage paths to the auxiliary building acceptance criteria are:

6.8.4.h.d.1 a) The combined bypass leakage rate to the auxiliary building shall be 0.25 La by applicable Type B and C tests.

6.8.4.h.d.2 b) Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (11.33 psig) during each Type A test.

Insert Page 5.5-16 Enclosure 2, Volume 16, Rev. 0, Page 169 of 270

Enclosure 2, Volume 16, Rev. 0, Page 170 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.h 5.5.16 Containment Leakage Rate Testing Program (continued) 4 14 6.8.4.h e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

6.8.4.h f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

[OPTION A/B Combined]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. [Type A][Type B and C] test requirements are in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. [Type B and C][Type A] test requirements are in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The 10 CFR 50, Appendix J, Option B test requirements shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

10

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

[ 3. ...]

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, [45 psig]. The containment design pressure is

[50 psig].

c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

d. Leakage rate acceptance criteria are:

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-17 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 170 of 270

Enclosure 2, Volume 16, Rev. 0, Page 171 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) 4 14

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and [< 0.75 La for Option A Type A tests] [ 0.75 La for Option B Type A tests].

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa. 10 b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

DOC M03 5.5.17 Battery Monitoring and Maintenance Program 4 15

---

REVIEWERS NOTE------------------------------------------

This program and the corresponding requirements in LCO 3.8.4, LCO 3.8.5, and LCO 3.8.6 require providing the information and verifications requested in the 5

Notice of Availability for TSTF-500, Revision 2, "DC Electrical Rewrite - Update to TSTF-360," (76FR54510).

This Program provides controls for battery restoration and maintenance. The program shall be in accordance with IEEE Standard (Std) 450-2002, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed by Regulatory Guide 1.129, Revision 2 (RG), with RG exceptions and program provisions as identified below:

a. The program allows the following RG 1.129, Revision 2 exceptions:

1. Battery temperature correction may be performed before or after conducting discharge tests.

2. RG 1.129, Regulatory Position 1, Subsection 2, "References," is not applicable to this program.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-18 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 171 of 270

Enclosure 2, Volume 16, Rev. 0, Page 172 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4

DOC M01 5.5.17 Battery Monitoring and Maintenance Program (continued) 15

3. In lieu of RG 1.129, Regulatory Position 2, Subsection 5.2, "Inspections," the following shall be used: "Where reference is made to the pilot cell, pilot cell selection shall be based on the lowest voltage cell in the battery."

4 In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string."

b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage < [2.13] V; 2 Table 4.8.2 2. Actions to determine whether the float voltage of the remaining Float Voltage battery cells is [2.13] V when the float voltage of a battery cell has 2 been found to be < [2.13] V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

6.17 5.5.18 Control Room Envelope (CRE) Habitability Program 4 16 A Control Room Envelope (CRE) Habitability Program shall be established and Ventilation implemented to ensure that CRE habitability is maintained such that, with an V OPERABLE Control Room Emergency Filtration System (CREFS), CRE 3 occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in 2

excess of [5 rem whole body or its equivalent to any part of the body] [5 rem total SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-19 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 172 of 270

Enclosure 2, Volume 16, Rev. 0, Page 173 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.17 5.5.18 Control Room Envelope (CRE) Habitability Program (continued) 4 16 effective dose equivalent (TEDE)] for the duration of the accident. The program 2 shall include the following elements:

6.17.a a. The definition of the CRE and the CRE boundary.

6.17.b b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

6.17.c c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

[The following are exceptions to Sections C.1 and C.2 of Regulatory 2 Guide 1.197, Revision 0:

1. ;and] 2 6.17.d d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization V mode of operation by one train of the CREFS, operating at the flow rate 3 2

required by the VFTP, at a Frequency of [18] months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the

[18] month assessment of the CRE boundary. 2 6.17.e e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

6.17.f f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-20 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 173 of 270

Enclosure 2, Volume 16, Rev. 0, Page 174 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 [ Setpoint Control Program This program shall establish the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analyses, provides a means for processing changes to instrumentation setpoints, and identifies setpoint methodologies to ensure instrumentation will function as required. The program shall ensure that testing of automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A) verifies that instrumentation will function as required.

a. The program shall list the Functions in the following specifications to which it applies:

1. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation;"

2. LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation Functions;"

3. LCO 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation;"

4. LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation;"

5. LCO 3.3.7, "Control Room Emergency Filtration System (CREFS)

Actuation Instrumentation;" 4

6. LCO 3.3.8, "Fuel Building Air Cleanup System (FBACS) Actuation Instrumentation;" and

7. LCO 3.3.9, "Boron Dilution Protection System (BDPS)."

b. The program shall require the Nominal Trip Setpoint (NTSP), Allowable Value (AV), As-Found Tolerance (AFT), and As-Left Tolerance (ALT) (as applicable) of the Functions described in paragraph a. are calculated using the NRC approved setpoint methodology, as listed below. In addition, the program shall contain the value of the NTSP, AV, AFT, and ALT (as applicable) for each Function described in paragraph a. and shall identify the setpoint methodology used to calculate these values.

---

Reviewer's Note----------------------------------------

List the NRC safety evaluation report by letter, date, and ADAMS accession number (if available) that approved the setpoint methodologies.

1. [Insert reference to NRC safety evaluation that approved the setpoint methodology.]

c. The program shall establish methods to ensure that Functions described in paragraph a. will function as required by verifying the as-left and as-found settings are consistent with those established by the setpoint methodology.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-21 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 174 of 270

Enclosure 2, Volume 16, Rev. 0, Page 175 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 Setpoint Control Program (continued)

d. -----------------------------------REVIEWERS NOTE--------------------------------------

A license amendment request to implement a Setpoint Control Program must list the instrument functions to which the program requirements of paragraph d. will be applied. Paragraph d. shall apply to all Functions in the Reactor Trip System and Engineered Safety Feature Actuation System specifications unless one or more of the following exclusions apply:

1. Manual actuation circuits, automatic actuation logic circuits or to instrument functions that derive input from contacts which have no associated sensor or adjustable device, e.g., limit switches, breaker position switches, manual actuation switches, float switches, proximity detectors, etc. are excluded. In addition, those permissives and interlocks that derive input from a sensor or adjustable device that is tested as part of another TS function are excluded.

2. Settings associated with safety relief valves are excluded. The performance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program.

3. Functions and Surveillance Requirements which test only digital 4

components are normally excluded. There is no expected change in result between SR performances for these components. Where separate as-left and as-found tolerance is established for digital component SRs, the requirements would apply.

The program shall identify the Functions described in paragraph a. that are automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A). The NTSP of these Functions are Limiting Safety System Settings. These Functions shall be demonstrated to be functioning as required by applying the following requirements during CHANNEL CALIBRATIONS, CHANNEL OPERATIONAL TESTS, and TRIP ACTUATING DEVICE OPERATIONAL TESTS that verify the NTSP.

1 The as-found value of the instrument channel trip setting shall be compared with the previous as-left value or the specified NTSP.

2. If the as-found value of the instrument channel trip setting differs from the previous as-left value or the specified NTSP by more than the pre-defined test acceptance criteria band (i.e., the specified AFT), then the instrument channel shall be evaluated before declaring the SR met and returning the instrument channel to service. This condition shall be entered in the plant corrective action program.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-22 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 175 of 270

Enclosure 2, Volume 16, Rev. 0, Page 176 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 Setpoint Control Program (continued)

3. If the as-found value of the instrument channel trip setting is less conservative than the specified AV, then the SR is not met and the instrument channel shall be immediately declared inoperable.

4. The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NTSP at the completion of the 4 surveillance test; otherwise, the channel is inoperable (setpoints may be more conservative than the NTSP provided that the as-found and as-left tolerances apply to the actual setpoint used to confirm channel performance).

e. The program shall be specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference]. ]

DOC M04 5.5.20 [ Surveillance Frequency Control Program 4 2 17 This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. ] 2 SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.5-23 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 176 of 270

Enclosure 2, Volume 16, Rev. 0, Page 177 of 270 CTS Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

1.18 5.5.1 Offsite Dose Calculation Manual (ODCM) 1.18 a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program, and 1 1.18 b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.1] and Specification [5.6.2]. 2 6.14.1.1 Licensee initiated changes to the ODCM:

6.14.1.1 a. Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

6.14.1.1.a

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s) and 6.14.1.1.b 2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations, 6.14.1.2 b. Shall become effective after the approval of the plant manager, and 6.14.1.3 c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-1 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 177 of 270

Enclosure 2, Volume 16, Rev. 0, Page 178 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.a 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include

[Recirculation Spray, Safety Injection, Chemical and Volume Control, gas stripper, and Hydrogen Recombiner]. The program shall include the following: 2 Residual Heat Removal, Containment Spray, and RCS Sampling 6.8.4.a a. Preventive maintenance and periodic visual inspection requirements and 6.8.4.b b. Integrated leak test requirements for each system at least once per

[18] months.

The provisions of SR 3.0.2 are applicable.

5.5.3 [ Post Accident Sampling

---

REVIEWER'S NOTE----------------------------------------

This program may be eliminated based on the implementation of WCAP-14986, Rev. 1, "Post Accident Sampling System Requirements: A Technical Basis," and the associated NRC Safety Evaluation dated June 14, 2000, and implementation of the following commitments:

1. [Licensee] has developed contingency plans for obtaining and analyzing highly radioactive samples of reactor coolant, containment sump, and containment atmosphere. The contingency plans will be contained in emergency plan implementing procedures and implemented with the implementation of the License amendment. Establishment of contingency plans is considered a regulatory commitment.

2. The capability for classifying fuel damage events at the Alert level threshold 4 has been established for [Plant] at radioactivity levels of 300 mCi/cc dose equivalent iodine. This capability may utilize the normal sampling system and/or correlations of sampling or letdown line dose rates to coolant concentrations. This capability will be described in emergency plan implementing procedures and implemented with the implementation of the License amendment. The capability for classifying fuel damage events is considered a regulatory commitment.

3. [Licensee] has established the capability to monitor radioactive iodines that have been released to offsite environs. This capability is described in our emergency plan implementing procedures. The capability to monitor radioactive iodines is considered a regulatory commitment.

This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-2 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 178 of 270

Enclosure 2, Volume 16, Rev. 0, Page 179 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Post Accident Sampling (continued) and containment atmosphere samples under accident conditions. The program shall include the following:

4

a. Training of personnel,

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analysis equipment. ]

4 6.8.4.f 5.5.4 Radioactive Effluent Controls Program 3

6.8.4.f This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

6.8.4.f.1) a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM, 6.8.4.f.2) b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, 6.8.4.f.3) c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, 6.8.4.f.4) d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I, 6.8.4.f.5) e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days, 6.8.4.f.6) f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I, SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-3 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 179 of 270

Enclosure 2, Volume 16, Rev. 0, Page 180 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.f 5.5.4 Radioactive Effluent Controls Program (continued) 4 3

6.8.4.f.7) g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

6.8.4.f.7)).1 1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin and 6.8.4.f.7)).2 2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ, 6.8.4.f.8) h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, 6.8.4.f.9) i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, and 6.8.4.f.10) j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

6.8.4.f The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

6.8.4.l 5.5.5 Component Cyclic or Transient Limit 4 U 5.2.1 4

This program provides controls to track the FSAR, Section [ ], cyclic and 2 transient occurrences to ensure that components are maintained within the design limits.

5.5.6 [ Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon 4

Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC.

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies. ]

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-4 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 180 of 270

Enclosure 2, Volume 16, Rev. 0, Page 181 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4.0.5.c 5.5.7 Reactor Coolant Pump Flywheel Inspection Program 4 5

This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

ultrasonic In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over 3 magnetic the volume from the inner bore of the flywheel to the circle one-half of the outer particle radius or a surface examination (MT and/or PT) of exposed surfaces of the penetrant liquid 3 removed flywheels may be conducted at 20 year intervals.

---

REVIEWER'S NOTE----------------------------------------

The inspection interval and scope for RCP flywheels stated above can be applied to plants that satisfy the requirements in WCAP-15666, "Extension of Reactor 5 Coolant Pump Motor Flywheel Examination."

4.0.5 5.5.8 Inservice Testing Program 4 6

4.0.5 This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following: 6 pumps and valves 4.0.5.b a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice testing inservice testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days 4.0.5.c b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities, 4.0.5.d c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-5 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 181 of 270

Enclosure 2, Volume 16, Rev. 0, Page 182 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4.0.5 5.5.8 Inservice Testing Program (continued) 4 6

4.0.5.e d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

6.8.4.k 5.5.9 Steam Generator (SG) Program 4 7

A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

6.8.4.k.a a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging [or repair] of 2 tubes. Condition monitoring assessments shall be conducted during each or outage during which the SG tubes are inspected, plugged, [or repaired] to 1 2 confirm that the performance criteria are being met.

6.8.4.k.b b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

6.8.4.k.b.1 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients TSTF-

), 510-A included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

6.8.4.k.b.2

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-6 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 182 of 270

Enclosure 2, Volume 16, Rev. 0, Page 183 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) 4 7

exceed [1 gpm] per SG [, except for specific types of degradation at 2

specific locations as described in paragraph c of the Steam Generator TSTF-Program. 510-A

]

6.8.4.k.b.3

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

]

TSTF-6.8.4.k.c c. Provisions for SG tube repair criteria. Tubes found by inservice inspection 510-A plugging [or to contain flaws with a depth equal to or exceeding [40%] of the nominal 2 tube wall thickness shall be plugged [or repaired].

]

---

REVIEWER'S NOTE----------------------------------------

Alternate tube repair criteria currently permitted by plant technical specifications are plugging [or listed here. The description of these alternate tube repair criteria should be ]

TSTF-equivalent to the descriptions in current technical specifications and should also 510-A plugging [or include any allowed accident induced leakage rates for specific types of degradation at specific locations associated with tube repair criteria. ] 5

]

plugging [or

[The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. . . .]

6.8.4.k.d d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the plugging [or tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may TSTF-satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not 2 510-A

]

part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals assessment shall be such as to ensure that SG tube integrity is maintained until the next TSTF-SG inspection. An assessment of degradation shall be performed to 7 510-A determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-7 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 183 of 270

Enclosure 2, Volume 16, Rev. 0, Page 184 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) 4 7

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REVIEWER'S NOTE----------------------------------------

Plants are to include the appropriate Frequency (e.g., select the appropriate Item 2.) for their SG design. The first Item 2 is applicable to SGs with Alloy 600 mill annealed tubing. The second Item 2 is applicable to SGs with Alloy 600 5 thermally treated tubing. The third Item 2 is applicable to SGs with Alloy 690 thermally treated tubing.

6.8.4.k.d.1 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

installation

[2. Inspect 100% of the tubes at sequential periods of 60 effective full INSERT 1 power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.]

[2. Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period INSERT 2 shall be considered to begin after the first inservice inspection of the TSTF-SGs. In addition, inspect 50% of the tubes by the refueling outage 510-A nearest the midpoint of the period and the remaining 50% by the 2

refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.]

6.8.4.k.d.2 [2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, INSERT 3 thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.]

6.8.4.k.d.3 3. If crack indications are found in any SG tube, then the next inspection affected and for each SG for the degradation mechanism that caused the crack potentially affected TSTF-indication shall not exceed 24 effective full power months or one 510-A results in more refueling outage (whichever is less). If definitive information, such as frequent inspections from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

6.8.4.k.e e. Provisions for monitoring operational primary to secondary LEAKAGE.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-8 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 184 of 270

Enclosure 2, Volume 16, Rev. 0, Page 185 of 270 CTS 5.5 2

INSERT 1 After the first refueling outage following SG installation, inspect each steam generator at least every 24 effective full power months or at least every refueling outage (whichever results in more frequent inspections). In addition, inspect 100% of the tubes at sequential periods of 60 effective full power months beginning after the first refueling outage inspection following SG installation. Each 60 effective full power month inspection period may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

2 INSERT 2 TSTF-After the first refueling outage following SG installation, inspect each SG at least every 48 510-A effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

---

Reviewer's Note ------------------------------------------------------

A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; Insert Page 5.5-8 Enclosure 2, Volume 16, Rev. 0, Page 185 of 270

Enclosure 2, Volume 16, Rev. 0, Page 186 of 270 CTS 5.5 b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.

INSERT 3 After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging [or repair] criteria, the minimum number of locations 2 inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the TSTF-510-A determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

---

Reviewer's Note ------------------------------------------------------

A licensee may elect to retain historical and existing inspection period lengths in order to not revise those inspection periods.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.

Insert Page 5.5-8 Enclosure 2, Volume 16, Rev. 0, Page 186 of 270

Enclosure 2, Volume 16, Rev. 0, Page 187 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) 4 7

[f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the 2 purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

---

REVIEWER'S NOTE----------------------------------------

Tube repair methods currently permitted by plant technical specifications are to be listed here. The description of these tube repair methods should be 5 equivalent to the descriptions in current technical specifications. If there are no approved tube repair methods, this section should not be used.

1. . . .] 2 4

6.8.4.c 5.5.10 Secondary Water Chemistry Program 8

6.8.4.c This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion 8 cracking. The program shall include:

6.8.4.c.(i) a. Identification of a sampling schedule for the critical variables and control points for these variables, 6.8.4.c.(ii) b. Identification of the procedures used to measure the values of the critical variables, 6.8.4.c.(iii) c. Identification of process sampling points, which shall include monitoring the 11 discharge of the condensate pumps for evidence of condenser in leakage, 6.8.4.c.(iv) d. Procedures for the recording and management of data, 6.8.4.c.(v) e. Procedures defining corrective actions for all off control point chemistry conditions, and 6.8.4.c.(vi) f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-9 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 187 of 270

Enclosure 2, Volume 16, Rev. 0, Page 188 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) Regulatory Positions C.5.a, C.5.c, 4

C.5.d and C.6.b of 3 9

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, 2 Revision 2, ASME N510-1989, and AG-1]. removal efficiently of 99.95% of dioctyl phthalate (DOP) 3 INSERT 4 ANSI N510-1975 ASTM D3803-1989 4.6.1.8.e a. Demonstrate for each of the ESF systems that an inplace test of the high 4.7.7.c.3 4.7.7.f 4.6.1.8.b.1 efficiency particulate air (HEPA) filters shows a penetration and system 4.7.8.b.3 4.7.7.c.1 4.7.8.b.1 bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, 2 4.7.8.e 4.9.12.b.3 4.9.12.b.1 Revision 2, and ASME N510-1989] at the system flowrate specified below 4.9.12.e [+/- 10%]. ANSI N510-1975 (except for the provisions of Sections 8 and 9) Regulatory Positions C.5.a and C.5.c of 3 2 ESF Ventilation System Flowrate INSERT 5 2

[ ] [ ]

removal efficiently of 99.95% of a halogenated 3 hydrocarbon refrigerant test gas 4.6.1.8.f b. Demonstrate for each of the ESF systems that an inplace test of the 4.7.7.c.3 4.7.7.g charcoal adsorber shows a penetration and system bypass < [0.05]% when 4.7.8.b.3 tested in accordance with [Regulatory Guide 1.52, Revision 2, and 2 4.7.8.f 4.9.12.b.3 ASME N510-1989] at the system flowrate specified below [+/- 10%]. 2 4.9.12.f ANSI N510-1975 (except for the provisions of Sections 8 and 9) Regulatory Positions C.5.a and C.5.d of 3 ESF Ventilation System Flowrate INSERT 6 3

[ ] [ ]

Regulatory Position C.6.b of 3 4.6.1.8.b.2 c. Demonstrate for each of the ESF systems that a laboratory test of a sample 2 4.6.1.8.c 4.7.7.c..2 of the charcoal adsorber, when obtained as described in [Regulatory < 2.5%

4.7.7.d Guide 1.52, Revision 2], shows the methyl iodide penetration less than the 3 4.7.8.b.2 4.7.8.c value specified below when tested in accordance with ASTM D3803-1989 at 4.9.12.b.2 a temperature of 30°C (86°F) and the relative humidity specified below. 3 4.9.12.c of 70%

ESF Ventilation System Penetration RH Face Velocity (fps)

[ ] [See Reviewer's [See [See Reviewer's EGTS Note] Reviewer's Note] 2 ABGTS Note] 3 CREVS

---

REVIEWER'S NOTE----------------------------------------

The use of any standard other than ASTM D3803-1989 to test the charcoal sample may result in an overestimation of the capability of the charcoal to adsorb 5 radioiodine. As a result, the ability of the charcoal filters to perform in a manner consistent with the licensing basis for the facility is indeterminate.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-10 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 188 of 270

Enclosure 2, Volume 16, Rev. 0, Page 189 of 270 CTS 5.5 4.6.1.8.b 3 INSERT 4 4.6.1.8.e 4.6.1.8.f 4.7.7.f The test described in Specification 5.5.9.a and 5.5.9.b shall be performed once per 18 months; 4.7.7.g after any structural maintenance on the high efficiency particulate air (HEPA) filter bank or 4.7.8.e 4.7.8.f charcoal adsorber bank housing; following painting, fire, or chemical release in any ventilation zone communicating with the system; and after each complete or partial replacement of a HEPA 4.6.1.8.c filter bank or charcoal adsorber bank.

4.6.1.8.b 4.7.7.c 4.7.7.d The test described in Specification 5.5.9.c shall be performed once per 18 months or after 720 4.7.8.b 4.7.8.c hours of filter operation; after any structural maintenance on the HEPA filter bank or charcoal 4.9.12.b adsorber bank housing; and following painting, fire, or chemical release in any ventilation zone 4.9.12.c communicating with the system.

4.6.1.8.d 4.7.7.e The test described in Specification 5.5.9.d and 5.5.9.e shall be performed once per 18 months.

4.7.8.d 4.9.12.d 3 INSERT 5 ESF Ventilation System Flow Rate (cfm)

Emergency Gas Treatment System (EGTS) 4000 Auxiliary Building Gas Treatment System (ABGTS) 9000 Control Room Emergency Ventilation System (CREVS) 4000 3

INSERT 6 ESF Ventilation System Flow Rate (cfm)

EGTS 4000 ABGTS 9000 CREVS 4000 Insert Page 5.5-10 Enclosure 2, Volume 16, Rev. 0, Page 189 of 270

Enclosure 2, Volume 16, Rev. 0, Page 190 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (continued) 4 9

ASTM D 3803-1989 is a more stringent testing standard because it does not differentiate between used and new charcoal, it has a longer equilibration period performed at a temperature of 30°C (86°F) and a relative humidity (RH) of 95%

(or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test.

Allowable Penetration = [(100% - Methyl Iodide Efficiency

• for Charcoal Credited in Licensee's Accident Analysis) / Safety Factor]

When ASTM D3803-1989 is used with 30°C (86°F) and 95% RH (or 70% RH with humidity control) is used, the staff will accept the following:

5 Safety factor 2 for systems with or without humidity control.

Humidity control can be provided by heaters or an NRC-approved analysis that demonstrates that the air entering the charcoal will be maintained less than or equal to 70 percent RH under worst-case design-basis conditions.

If the system has a face velocity greater than 110 percent of 0.203 m/s (40 ft/min), the face velocity should be specified.

• This value should be the efficiency that was incorporated in the licensee's accident analysis which was reviewed and approved by the staff in a safety evaluation.

4.6.1.8.d.1 d. Demonstrate for each of the ESF systems that the pressure drop across the 4.7.7.e.1 4.7.8.b.3 combined HEPA filters, the prefilters, and the charcoal adsorbers is less 3 4.7.8.d.1 than the value specified below when tested in accordance with [Regulatory 2 4.9.12.b.3 4.9.12.d.1 Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate 4.7.7.c.3 specified below [+/- 10%]. 1975 ESF Ventilation System Delta P Flowrate INSERT 7 3

[ ] [ ] [ ]

4.7.8.d.4 [ e. Demonstrate that the heaters for each of the ESF systems dissipate the 4.9.12.d.3 value specified below [+/- 10%] when tested in accordance with

[ASME N510-1989].

ANSI N510-1975 2

ESF Ventilation System Wattage ]

[ ] [ ]

Auxiliary Building Gas Treatment System 32 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-11 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 190 of 270

Enclosure 2, Volume 16, Rev. 0, Page 191 of 270 CTS 5.5 3 INSERT 7 ESF Ventilation System Combined Flowrate (cfm)

Delta P (inches water gauge)

EGTS 5 4000 ABGTS 3 9000 CREVS 3 4000 Insert Page 5.5-11 Enclosure 2, Volume 16, Rev. 0, Page 191 of 270

Enclosure 2, Volume 16, Rev. 0, Page 192 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 3.11.1.4 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program 4 3.11.2.5 3.11.2.6 10 This program provides controls for potentially explosive gas mixtures contained decay in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of temporary 2 the Condensate Storage Tank, radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous Steam Generator radioactivity quantities shall be determined following the methodology in [Branch Layup Tank, and Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be 9 determined in accordance with [Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures"].

The program shall include:

4.11.2.5 a. The limits for concentrations of hydrogen and oxygen in the [Waste Gas 2 Holdup System] and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion), 11 4.11.2.6 b. A surveillance program to ensure that the quantity of radioactivity contained decay in [each gas storage tank and fed into the offgas treatment system] is less 2 than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of [an uncontrolled 2

release of the tanks' contents], and , Condensate Storage Tank, and

Steam Generator Layup Tank 11 4.11.1.4 temporary

c. A surveillance program to ensure that the quantity of radioactivity contained storage in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank 3 overflows and surrounding area drains connected to the [Liquid Radwaste Treatment System] is less than the amount that would result in exceeding concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-12 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 192 of 270

Enclosure 2, Volume 16, Rev. 0, Page 193 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.16 5.5.13 Diesel Fuel Oil Testing Program 4 11 6.16 A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

6.16.a a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

6.16.a.1 1. An API gravity or an absolute specific gravity within limits, 6.16.a.2

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and 6.16.a.3 3. A clear and bright appearance with proper color or a water and sediment content within limits.

6.16.b b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a.,

above, are within limits for ASTM 2D fuel oil, and 6.16.c

c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

6.8.4.j 5.5.14 Technical Specifications (TS) Bases Control Program 4 12 6.8.4.j This program provides a means for processing changes to the Bases of these Technical Specifications.

6.8.4.j.a a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

6.8.4.j.b b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

6.8.4.j.b.1 1. A change in the TS incorporated in the license or 6.8.4.j.b.2 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-13 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 193 of 270

Enclosure 2, Volume 16, Rev. 0, Page 194 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.j.d 5.5.14 Technical Specifications (TS) Bases Control Program (continued) 4 12 6.8.4.j.d c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR. 3 U 12 6.8.4.j.d d. Proposed changes that meet the criteria of Specification 5.5.14b above shall 4 be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

DOC M02 5.5.15 Safety Function Determination Program (SFDP) 4 13 This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected,

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists,

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities, and

d. Other appropriate limitations and remedial or compensatory actions.

11 A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable, or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable, or SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-14 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 194 of 270

Enclosure 2, Volume 16, Rev. 0, Page 195 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC M02 5.5.15 Safety Function Determination Program (SFDP) (continued) 4 13

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5. 16 Containment Leakage Rate Testing Program 4 14

[OPTION A]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

b. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

c. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests. 10

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

d. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

e. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-15 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 195 of 270

Enclosure 2, Volume 16, Rev. 0, Page 196 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.h 5.5.16 Containment Leakage Rate Testing Program (continued) 4 14 10

[OPTION B]

6.8.4.h a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

INSERT 8

2. The visual examination of the steel liner plate inside containment 3 intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

[ 3. ...]

6.8.4.h b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is [45 psig]. The containment design pressure is 2

[50 psig]. 11.33 12 6.8.4.h c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of 2 containment air weight per day.

0.25 6.8.4.h d. Leakage rate acceptance criteria are:

6.8.4.h.a 1. Containment leakage rate acceptance criterion is 1.0 La. During the 3 first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C 3 tests and 0.75 La for Type A tests.

6.8.4.h.b 2. Air lock testing acceptance criteria are:

6.8.4.h.b.1 a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

2 6.8.4.h.b.2 b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig]. 3 6 for at least two minutes 3

INSERT 9 Amendment XXX SEQUOYAH UNIT 2 Westinghouse STS 5.5-16 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 196 of 270

Enclosure 2, Volume 16, Rev. 0, Page 197 of 270 CTS 5.5 3 INSERT 8 6.8.4.h 1. Bypass leakage paths to the auxiliary building leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (12.46 psig) and the seal system capacity is adequate to maintain system pressure (or fluid head for the containment spray system and RHR spray system valves at penetrations 48A, 48B, 49A and 49B) for at least 30 days.

3 INSERT 9 6.8.4.h.c 3. For each containment purge supply and exhaust isolation valve, acceptance criteria is measured leakage rate to 0.05 La.

6.8.4.h.d 4. Bypass leakage paths to the auxiliary building acceptance criteria are:

6.8.4.h.d.1 a) The combined bypass leakage rate to the auxiliary building shall be 0.25 La by applicable Type B and C tests.

6.8.4.h.d.2 b) Penetrations not individually testable shall have no detectable leakage when tested with soap bubbles while the containment is pressurized to Pa (11.33 psig) during each Type A test.

Insert Page 5.5-16 Enclosure 2, Volume 16, Rev. 0, Page 197 of 270

Enclosure 2, Volume 16, Rev. 0, Page 198 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.h 5.5.16 Containment Leakage Rate Testing Program (continued) 4 14 6.8.4.h e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

6.8.4.h f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

[OPTION A/B Combined]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. [Type A][Type B and C] test requirements are in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. [Type B and C][Type A] test requirements are in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The 10 CFR 50, Appendix J, Option B test requirements shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

10

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

[ 3. ...]

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, [45 psig]. The containment design pressure is

[50 psig].

c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

d. Leakage rate acceptance criteria are:

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-17 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 198 of 270

Enclosure 2, Volume 16, Rev. 0, Page 199 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) 4 14

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and [< 0.75 La for Option A Type A tests] [ 0.75 La for Option B Type A tests].

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa. 10 b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

DOC M03 5.5.17 Battery Monitoring and Maintenance Program 4 15

---

REVIEWERS NOTE------------------------------------------

This program and the corresponding requirements in LCO 3.8.4, LCO 3.8.5, and LCO 3.8.6 require providing the information and verifications requested in the 5

Notice of Availability for TSTF-500, Revision 2, "DC Electrical Rewrite - Update to TSTF-360," (76FR54510).

This Program provides controls for battery restoration and maintenance. The program shall be in accordance with IEEE Standard (Std) 450-2002, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed by Regulatory Guide 1.129, Revision 2 (RG), with RG exceptions and program provisions as identified below:

a. The program allows the following RG 1.129, Revision 2 exceptions:

1. Battery temperature correction may be performed before or after conducting discharge tests.

2. RG 1.129, Regulatory Position 1, Subsection 2, "References," is not applicable to this program.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-18 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 199 of 270

Enclosure 2, Volume 16, Rev. 0, Page 200 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4

DOC M01 5.5.17 Battery Monitoring and Maintenance Program (continued) 15

3. In lieu of RG 1.129, Regulatory Position 2, Subsection 5.2, "Inspections," the following shall be used: "Where reference is made to the pilot cell, pilot cell selection shall be based on the lowest voltage cell in the battery."

4 In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration," the following may be used: "Following the test, record the float voltage of each cell of the string."

b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage < [2.13] V; 2 Table 4.8.2 2. Actions to determine whether the float voltage of the remaining Float Voltage battery cells is [2.13] V when the float voltage of a battery cell has 2 been found to be < [2.13] V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

6.17 5.5.18 Control Room Envelope (CRE) Habitability Program 4 16 A Control Room Envelope (CRE) Habitability Program shall be established and Ventilation implemented to ensure that CRE habitability is maintained such that, with an V OPERABLE Control Room Emergency Filtration System (CREFS), CRE 3 occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in 2

excess of [5 rem whole body or its equivalent to any part of the body] [5 rem total SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-19 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 200 of 270

Enclosure 2, Volume 16, Rev. 0, Page 201 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.17 5.5.18 Control Room Envelope (CRE) Habitability Program (continued) 4 16 effective dose equivalent (TEDE)] for the duration of the accident. The program 2 shall include the following elements:

6.17.a a. The definition of the CRE and the CRE boundary.

6.17.b b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

6.17.c c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

[The following are exceptions to Sections C.1 and C.2 of Regulatory 2 Guide 1.197, Revision 0:

1. ;and] 2 6.17.d d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization V mode of operation by one train of the CREFS, operating at the flow rate 3 2

required by the VFTP, at a Frequency of [18] months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the

[18] month assessment of the CRE boundary. 2 6.17.e e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

6.17.f f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-20 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 201 of 270

Enclosure 2, Volume 16, Rev. 0, Page 202 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 [ Setpoint Control Program This program shall establish the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analyses, provides a means for processing changes to instrumentation setpoints, and identifies setpoint methodologies to ensure instrumentation will function as required. The program shall ensure that testing of automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A) verifies that instrumentation will function as required.

a. The program shall list the Functions in the following specifications to which it applies:

1. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation;"

2. LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation Functions;"

3. LCO 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation;"

4. LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation;"

5. LCO 3.3.7, "Control Room Emergency Filtration System (CREFS)

Actuation Instrumentation;" 4

6. LCO 3.3.8, "Fuel Building Air Cleanup System (FBACS) Actuation Instrumentation;" and

7. LCO 3.3.9, "Boron Dilution Protection System (BDPS)."

b. The program shall require the Nominal Trip Setpoint (NTSP), Allowable Value (AV), As-Found Tolerance (AFT), and As-Left Tolerance (ALT) (as applicable) of the Functions described in paragraph a. are calculated using the NRC approved setpoint methodology, as listed below. In addition, the program shall contain the value of the NTSP, AV, AFT, and ALT (as applicable) for each Function described in paragraph a. and shall identify the setpoint methodology used to calculate these values.

---

Reviewer's Note----------------------------------------

List the NRC safety evaluation report by letter, date, and ADAMS accession number (if available) that approved the setpoint methodologies.

1. [Insert reference to NRC safety evaluation that approved the setpoint methodology.]

c. The program shall establish methods to ensure that Functions described in paragraph a. will function as required by verifying the as-left and as-found settings are consistent with those established by the setpoint methodology.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-21 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 202 of 270

Enclosure 2, Volume 16, Rev. 0, Page 203 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 Setpoint Control Program (continued)

d. -----------------------------------REVIEWERS NOTE--------------------------------------

A license amendment request to implement a Setpoint Control Program must list the instrument functions to which the program requirements of paragraph d. will be applied. Paragraph d. shall apply to all Functions in the Reactor Trip System and Engineered Safety Feature Actuation System specifications unless one or more of the following exclusions apply:

1. Manual actuation circuits, automatic actuation logic circuits or to instrument functions that derive input from contacts which have no associated sensor or adjustable device, e.g., limit switches, breaker position switches, manual actuation switches, float switches, proximity detectors, etc. are excluded. In addition, those permissives and interlocks that derive input from a sensor or adjustable device that is tested as part of another TS function are excluded.

2. Settings associated with safety relief valves are excluded. The performance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program.

3. Functions and Surveillance Requirements which test only digital 4

components are normally excluded. There is no expected change in result between SR performances for these components. Where separate as-left and as-found tolerance is established for digital component SRs, the requirements would apply.

The program shall identify the Functions described in paragraph a. that are automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A). The NTSP of these Functions are Limiting Safety System Settings. These Functions shall be demonstrated to be functioning as required by applying the following requirements during CHANNEL CALIBRATIONS, CHANNEL OPERATIONAL TESTS, and TRIP ACTUATING DEVICE OPERATIONAL TESTS that verify the NTSP.

1 The as-found value of the instrument channel trip setting shall be compared with the previous as-left value or the specified NTSP.

2. If the as-found value of the instrument channel trip setting differs from the previous as-left value or the specified NTSP by more than the pre-defined test acceptance criteria band (i.e., the specified AFT), then the instrument channel shall be evaluated before declaring the SR met and returning the instrument channel to service. This condition shall be entered in the plant corrective action program.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-22 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 203 of 270

Enclosure 2, Volume 16, Rev. 0, Page 204 of 270 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.19 Setpoint Control Program (continued)

3. If the as-found value of the instrument channel trip setting is less conservative than the specified AV, then the SR is not met and the instrument channel shall be immediately declared inoperable.

4. The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NTSP at the completion of the 4 surveillance test; otherwise, the channel is inoperable (setpoints may be more conservative than the NTSP provided that the as-found and as-left tolerances apply to the actual setpoint used to confirm channel performance).

e. The program shall be specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference]. ]

DOC M04 5.5.20 [ Surveillance Frequency Control Program 4 2 17 This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program. ] 2 SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.5-23 Rev. 4.0 3 Enclosure 2, Volume 16, Rev. 0, Page 204 of 270

Enclosure 2, Volume 16, Rev. 0, Page 205 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

1. Typographical/grammatical error corrected.

2. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

3. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

4. The bracketed ISTS 5.5.3, "Post Accident Sampling," ISTS 5.5.6, "Pre-Stressed Concrete Containment Tendon Surveillance Program," and ISTS 5.5.19, "Setpoint Control Program," are not included in the SQN ITS. Subsequent programs in the ITS Section 5.5 have been renumbered, as necessary.

5. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

6. The Inservice Testing (IST) Program (ISTS 5.5.8) has been modified to state that the IST Program provides control for ASME Code Class 1, 2, and 3 "pumps and valves" in place of the current "components." 10 CFR 50.55a(f) provides the regulatory requirements for an IST Program. It specifies that ASME Code Class 1, 2, and 3 pumps and valves are the only components covered by an IST Program.

10 CFR 50.55a(g) provides regulatory requirements for an Inservice Inspection (ISI) Program. It specifies that ASME Code Class 1, 2, and 3 components are covered by the ISI Program, and that pumps and valves are covered by the IST Program in 10 CFR 50.55a(f). The ISTS does not include ISI Program requirements as these requirements have been relocated to a plant specific document. Therefore, the components to which the IST Program applies (i.e.,

pumps and valves) have been added for clarity.

7. Changes made to improve clarity.

8. ISTS 5.5.10 (ITS 5.5.8) provides the requirements for the Secondary Water Chemistry Program. The program in the ISTS includes requirements to provide controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion. ITS 5.5.8 provides controls for monitoring secondary water chemistry only to inhibit SG tube degradation. This change is consistent with the current SQN licensing bases.

9. The program details of the Explosive Gas and Storage Tank Radioactivity Monitoring Program are described in ISTS 5.5.12 (ITS 5.5.10) parts a, b, and c.

Therefore, the sentence in the introductory paragraph that specifies a method to determine the explosive gas and storage tank radioactivity is not necessary.

10. SQN complies with Option B of 10 CFR 50, Appendix J. Therefore, the ISTS 5.5.16 Option A and combined Option A and B provisions have been deleted.

11. These punctuation corrections have been made consistent with the Writers Guide for the Improved Standard Technical Specifications NEI 01-03, Section 5.1.3.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 205 of 270

Enclosure 2, Volume 16, Rev. 0, Page 206 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 206 of 270

Enclosure 2, Volume 16, Rev. 0, Page 207 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 207 of 270

, Volume 16, Rev. 0, Page 208 of 270 ATTACHMENT 6 ITS 5.6, REPORTING REQUIREMENTS , Volume 16, Rev. 0, Page 208 of 270

, Volume 16, Rev. 0, Page 209 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 209 of 270

Enclosure 2, Volume 16, Rev. 0, Page 210 of 270 ITS A01 ITS 5.6 6.0 ADMINISTRATIVE CONTROLS

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack. See ITS 5.5

e. Provisions for monitoring operational primary-to-secondary leakage.

l. Component Cyclic and Transient Limit This program provides controls to track the FSAR, Section 5.2.1, cyclic and transient occurrences to ensure that components are maintained within the design limits.

5.6 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 5.6 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4.

STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED ANNUAL REPORTS 1/

6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to A02 March 1 of the year following initial criticality.

6.9.1.5 DELETED 1/ A single submittal may be made for a multiple unit station. The submittal should combine those A02 sections that are common to all units at the station.

August 2, 2006 SEQUOYAH - UNIT 1 6-11b Amendment No. 12, 32, 58, 72, 74, 148, 174, 233, 280, 300, 306, 309 Page 1 of 16 Enclosure 2, Volume 16, Rev. 0, Page 210 of 270

Enclosure 2, Volume 16, Rev. 0, Page 211 of 270 ITS A01 ITS 5.6 ADMINISTRATIVE CONTROLS 5.6.1 1/

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit by May 15 during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives L01 outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

INSERT 1 M01 6.9.1.7 (Relocated to the ODCM.)

5.6.2 1/

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT in accordance with A03 10 CFR 50.36a 6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

6.9.1.9 (Relocated to the ODCM or PCP.)

5.6.1 Note, 5.6.2 Note 1/ A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

April 5, 2005 SEQUOYAH - UNIT 1 6-12 Amendment No. 42, 58, 74, 117, 148, 169, 281, 300 Page 2 of 16 Enclosure 2, Volume 16, Rev. 0, Page 211 of 270

Enclosure 2, Volume 16, Rev. 0, Page 212 of 270 ITS 5.6 M01 INSERT 1 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

Insert Page 6-12 Page 3 of 16 Enclosure 2, Volume 16, Rev. 0, Page 212 of 270

Enclosure 2, Volume 16, Rev. 0, Page 213 of 270 ITS A01 ITS 5.6 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED.

5.6.3 CORE OPERATING LIMITS REPORT 5.6.3.a 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. f1(I) limits for Overtemperature Delta T Trip Setpoints and f2(I) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4. Control Bank Insertion Limits for Specification 3/4.1.3.6,

5. AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6. Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

5.6.3.b 6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents:

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). A04 Revision 1,

1. BAW-10180P-A, "NEMO - Nodal Expansion Method Optimized" Revision 0, , March 1993

2. BAW-10169P-A, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants" , June 1989 Revision 0, , October 1989

3. BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs A07

4. BAW-10168P-A, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants EMF-2328 (P)(A), PWR Small Break LOCA Evaluation Model, March 2001 SL 2.1.1, Reactor Core Safety Limits LCO 3.1.1, SHUTDOWN MARGIN LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T and Overpower T A05 Nominal Trip Setpoint denoted values LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.9.1, Boron Concentration November 16, 2006 SEQUOYAH - UNIT 1 6-13 Amendment No. 52, 58, 72, 74, 117, 152, 155, 156, 171, 216, 223, 281, 300, 314 Page 4 of 16 Enclosure 2, Volume 16, Rev. 0, Page 213 of 270

Enclosure 2, Volume 16, Rev. 0, Page 214 of 270 ITS A01 ITS 5.6 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

5. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code

6. WCAP-10266-P-A, The 1981 Revision of Westinghouse Evaluation Model Using BASH CODE" Revision 1, 5 7. BAW-10227P-A, Evaluation of Advanced Cladding and Structural Material (M5) in PWR A07 Reactor Fuel"

, June 2003 6 8. BAW-10186-A, Extended Burnup Evaluation Revision 2, , June 2003 7 9. EMF-2103P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors Revision 0,

, April 2003 INSERT 2 5.6.3.c 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

M02 5.6.3.d 6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC A06 Document Control Desk with copies to the Regional Administrator and Resident Inspector.

5.6.4 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 5.6.4.a 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, RCS Pressure and Temperature (P/T) Limits Specification 3.4.12, Low Temperature Over Pressure Protection (LTOP) System 5.6.4.b 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. Westinghouse Topical Report WCAP-14040-NP-A, Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

2. Westinghouse Topical Report WCAP-15293, Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation.

3. Westinghouse Topical Report WCAP-15984, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2.

5.6.4.c 6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

STEAM GENERATOR TUBE INSPECTION REPORT 5.6.6 6.9.1.16 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.8.4.k, Steam Generator (SG) Program. The report shall include:

September 24, 2008 SEQUOYAH - UNIT 1 6-13a Amendment No. 52, 58, 72, 74, 117, 155, 223, 241, 258, 294, 297, 306, 314, 320 Page 5 of 16 Enclosure 2, Volume 16, Rev. 0, Page 214 of 270

Enclosure 2, Volume 16, Rev. 0, Page 215 of 270 ITS 5.6 A07 INSERT 2

8. BAW-1 0241 P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005

9. BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10. BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11. BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12. BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004 Insert Page 6-13a Page 6 of 16 Enclosure 2, Volume 16, Rev. 0, Page 215 of 270

Enclosure 2, Volume 16, Rev. 0, Page 216 of 270 ITS A01 ITS 5.6 ADMINISTRATIVE CONTROLS 5.6.6.a a. The scope of inspections performed on each SG, 5.6.6.b

b. Active degradation mechanisms found, 5.6.6.c c. Nondestructive examination techniques utilized for each degradation mechanism, 5.6.6.d d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, 5.6.6.e e. Number of tubes plugged during the inspection outage for each active degradation mechanism, 5.6.6.f f. Total number and percentage of tubes plugged to date, 5.6.6.g g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and 5.6.6.f h. The effective plugging percentage for all plugging in each SG.

SPECIAL REPORTS 5.6 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This specification has been deleted.

6.10 RECORD RETENTION (DELETED)

February 23, 2006 SEQUOYAH - UNIT 1 6-14 Amendment No. 42, 52, 58, 72, 74, 117, 148, 155, 163, 174, 178, 223, 233, 241, 258, 294, 297, 306 Page 7 of 16 Enclosure 2, Volume 16, Rev. 0, Page 216 of 270

Enclosure 2, Volume 16, Rev. 0, Page 217 of 270 ITS A01 ITS 5.6 TABLE 3.3-10 (Continued)

ACTION STATEMENTS (Continued)

ACTION 4 - With the number of channels less than the minimum channels required, initiate an See ITS 3.3.3 alternate method of monitoring containment area radiation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and either restore 5.6.5 the inoperable channel(s) to OPERABLE status within 30 days, or prepare and submit a special report to the Commission pursuant to Specification 6.9.2.1 within the next 14 days that provides actions taken, cause of the inoperability, and plans and schedule for restoring the channels to OPERABLE status.

ACTION 5 - NOTE: Also refer to the applicable action requirements from LCO 3.3.3.5 since it may contain more restrictive actions.

a. With the number of channels on one or more steam generators less than the minimum channels required for either flow rate or valve position, restore the inoperable channel to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 See ITS 3.3.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

b. With the number of channels on one or more steam generators less than the minimum channels required for flow rate and valve position, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

April 11, 2005 SEQUOYAH - UNIT 1 3/4 3-57b Amendment No. 112, 149, 159, 301 Page 8 of 16 Enclosure 2, Volume 16, Rev. 0, Page 217 of 270

Enclosure 2, Volume 16, Rev. 0, Page 218 of 270 ITS A01 ITS 5.6.

ADMINISTRATIVE CONTROLS 5.6 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 5.6 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted in accordance with 10 CFR 50.4.

STARTUP REPORT 6.9.1.1 DELETED 6.9.1.2 DELETED 6.9.1.3 DELETED 1

/

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar A02 year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 DELETED 1

/A single submittal may be made for a multiple unit station. The submittal should combine those A02 sections that are common to all units at the station.

April 13, 2009 SEQUOYAH - UNIT 2 6-11 Amendment No. 28, 34, 50, 64, 66, 107, 134, 165, 207, 223, 231, 271, 272, 289, 298 Page 9 of 16 Enclosure 2, Volume 16, Rev. 0, Page 218 of 270

Enclosure 2, Volume 16, Rev. 0, Page 219 of 270 ITS A01 ITS 5.6.

ADMINISTRATIVE CONTROLS 1

/

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 5.6.1 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit by May 15 during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives L01 outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

INSERT 1 M01 6.9.1.7 (Relocated to the ODCM.)

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT in accordance with A03 10 CFR 50.36a 5.6.2 6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

6.9.1.9 (Relocated to the ODCM or PCP.)

1 5.6.1 Note, /A single submittal may be made for a multiple unit station. The submittal should combine those sections 5.6.2 Note that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

August 2, 1993 SEQUOYAH - UNIT 2 6-12 Amendment No. 34, 50, 66, 107, 134, 159 Page 10 of 16 Enclosure 2, Volume 16, Rev. 0, Page 219 of 270

Enclosure 2, Volume 16, Rev. 0, Page 220 of 270 ITS 5.6 M01 INSERT 1 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

Insert Page 6-12 Page 11 of 16 Enclosure 2, Volume 16, Rev. 0, Page 220 of 270

Enclosure 2, Volume 16, Rev. 0, Page 221 of 270 ITS A01 ITS 5.6.

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED 5.6.3 CORE OPERATING LIMITS REPORT 5.6.3.a 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. f1(I) limits for Overtemperature Delta T Trip Setpoints and f2(I) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4. Control Bank Insertion Limits for Specification 3/4.1.3.6,

5. AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6. Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

5.6.3.b 6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents:

The COLR will contain the complete identification for each of the TS referenced topical reports used to A04 prepare the COLR (i.e., report number, title, revision, date, and any supplements).

1. BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2. BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3. BAW-10163P-A, Revision 0, Core Operating Limit Methodology for Westinghouse Designed PWRs, June 1989

4. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, March 2001

5. BAW-10227P-A, Revision 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6. BAW-10186P-A, Revision 2, Extended Burnup Evaluation, June 2003

7. EMF-2103P-A, Revision 0, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, April 2003 SL 2.1.1, Reactor Core Safety Limits LCO 3.1.1, SHUTDOWN MARGIN LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T and Overpower A05 T Nominal Trip Setpoint denoted values LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB Limits LCO 3.9.1, Boron Concentration September 26, 2012 SEQUOYAH - UNIT 2 6-13 Amendment No. 44, 50, 64, 66, 107, 134, 142, 146, 161, 206, 214, 223, 272, 289, 303, 324 Page 12 of 16 Enclosure 2, Volume 16, Rev. 0, Page 221 of 270

Enclosure 2, Volume 16, Rev. 0, Page 222 of 270 ITS A01 ITS 5.6.

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

8. BAW-10241P-A, Revision 1, BHTP DNB Correlation Applied with LYNXT, July 2005

9. BAW-10199P-A, Revision 0, The BWU Critical Heat Flux Correlations, August 1996

10. BAW-10189P-A, CHF Testing and Analysis of the Mark-BW Fuel Assembly Design, January 1996

11. BAW-10159P-A, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, August 1990

12. BAW-10231(P)(A), Revision 1, COPERNIC Fuel Rod Design Computer Code, January 2004 5.6.3.c 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. M02 5.6.3.d 6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC A06 Document Control Desk with copies to the Regional Administrator and Resident Inspector.

5.6.4 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 5.6.4.a 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, RCS Pressure and Temperature (P/T) Limits Specification 3.4.12, Low Temperature Over Pressure Protection (LTOP) System 5.6.4.b 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. Westinghouse Topical Report WCAP-14040-NP-A, Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

2. Westinghouse Topical Report WCAP-15321, Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation.

3. Westinghouse Topical Report WCAP-15984, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2.

5.6.4.c 6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

September 26, 2012 SEQUOYAH - UNIT 2 6-14 Amendment No. 44, 50, 64, 66, 107, 134, 146, 206, 214, 231, 249, 284, 303, 305, 311, 324 Page 13 of 16 Enclosure 2, Volume 16, Rev. 0, Page 222 of 270

Enclosure 2, Volume 16, Rev. 0, Page 223 of 270 ITS A01 ITS 5.6.

ADMINISTRATIVE CONTROLS STEAM GENERATOR (SG) TUBE INSPECTION REPORT 5.6.6 6.9.1.16.1 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, Steam Generator (SG) Program. The report shall include:

5.6.6.a a. The scope of inspections performed on each SG, 5.6.6.b b. Active degradation mechanisms found, 5.6.6.c

c. Nondestructive examination techniques utilized for each degradation mechanism, 5.6.6.d

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, 5.6.6.e

e. Number of tubes plugged during the inspection outage for each active degradation mechanism, 5.6.6.f f. Total number and percentage of tubes plugged to date, 5.6.6.g

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and 5.6.6.f h. The effective plugging percentage for all plugging in each SG.

September 26, 2012 SEQUOYAH - UNIT 2 6-14a Amendment No. 305, 323, 324 Page 14 of 16 Enclosure 2, Volume 16, Rev. 0, Page 223 of 270

Enclosure 2, Volume 16, Rev. 0, Page 224 of 270 ITS A01 ITS 5.6.

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 5.6 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This specification has been deleted.

6.10 RECORD RETENTION (DELETED)

July 10, 2012 SEQUOYAH - UNIT 2 6-15 Amendment No. 28, 44, 50, 64, 66, 107, 134, 146, 153, 165, 169, 206, 214, 223, 231, 249, 284, 309, 323 Page 15 of 16 Enclosure 2, Volume 16, Rev. 0, Page 224 of 270

Enclosure 2, Volume 16, Rev. 0, Page 225 of 270 ITS A01 ITS 5.6.

TABLE 3.3-10 (Continued)

ACTION STATEMENTS (Continued)

ACTION 4 - With the number of channels less than the minimum channels required, initiate an See ITS 3.3.3 alternate method of monitoring containment area radiation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and either restore the inoperable channel(s) to OPERABLE status within 30 days, or prepare and 5.6.5 submit a special report to the Commission pursuant to Specification 6.9.2.1 within 14 days that provides actions taken, cause of the inoperability, and plans and schedule for restoring the channels to OPERABLE status.

ACTION 5 - NOTE: Also refer to the applicable action requirements from LCO 3.3.3.5 since it may contain more restrictive actions.

a. With the number of channels on one or more steam generators less than the minimum channels required for either flow rate or valve position, restore the inoperable channel to OPERABLE status within 30 days or be in at least HOT See ITS STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 3.3.3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the number of channels on one or more steam generators less than the minimum channels required for flow rate and valve position, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 3-58b Amendment Nos. 102, 135, 149, 290 Page 16 of 16 Enclosure 2, Volume 16, Rev. 0, Page 225 of 270

Enclosure 2, Volume 16, Rev. 0, Page 226 of 270 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.9.1.4 states that, annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year following initial criticality. ITS 5.6 does not include the requirement for annual reports. This changes the CTS by not including the requirements.

The purpose of CTS 6.9.1.4 is to specify submittal dates of annual reports for associated activities. This change is acceptable because no activities are associated with the current Specification. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 6.9.1.8 requires the Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year to be submitted prior to May 1 of each year. ITS 5.6.2 requires this report, the Radioactive Effluent Release Report, covering the operation of the unit in the previous year to be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. This changes the CTS by explicitly stating the report shall be submitted in accordance with 10 CFR 50.36a.

The purpose to CTS 6.9.1.8 is to provide the requirements associated with the Radioactive Effluent Release Report. 10 CFR 50.36a, "Technical Specifications on Effluents from Nuclear Power Reactors," also provides requirements for submission of a report to the Commission annually that specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents during the previous 12 months. 10 CFR 50.36a also states that the time between submissions of the reports must be no longer than 12 months. This change is acceptable because the CTS reporting requirements have not changed, ITS explicitly states that the reporting requirement of "prior to May 1 of each year," is also in accordance with 10 CFR 50.36a. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 6.9.1.14.a requires, in part, the COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e.,

report number, title, revision, date, and any supplements). ITS 5.6.3 b Reviewers Note states, licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete Sequoyah Unit 1 and Unit 2 Page 1 of 4 Enclosure 2, Volume 16, Rev. 0, Page 226 of 270

Enclosure 2, Volume 16, Rev. 0, Page 227 of 270 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). This changes the CTS by not including the requirement of referencing Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). SQN has received prior approval by the NRC to include reference Topical Reports used to prepare the COLR (i.e., report number, title revision, date, and any supplements) in the Specification.

This change is acceptable because the Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements) have been included in the Specification. This change is designated as administrative because it does not result in technical changes to the CTS.

A05 CTS 6.9.1.14 contains a list of the core operating limits established and documented in the COLR. ITS 5.6.5.a includes additional core operating limits established and documented in the COLR. These are Reactor Core Safety Limits; SHUTDOWN MARGIN; Reactor Trip System (RTS) Instrumentation, -

Overtemperature T and Overpower T Nominal Trip Setpoint denoted values; RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits; and Boron Concentration. These limits had previously been addressed in other parts of the CTS, but are being moved to the COLR in the ITS, and because of this are listed in ITS 5.6.5.a. This changes the CTS by adding core operating limits established and documented in the COLR because they are being moved there as part of changes to other parts of the CTS. Technical aspects of the changes are addressed in the Discussion of Changes for the respective individual ITS Specifications.

This change is acceptable because it administratively documents changes made to other parts of the CTS and the COLR. This change is designated as administrative because it does not result in technical changes to the CTS.

A06 CTS 6.9.1.14.c requires, in part the CORE OPERATING LIMITS REPORT (COLR) to be provided to the NRC document control desk with copies to the Regional Administrator and Resident Inspector. ITS 5.6.3.d requires the COLR to be provided to the NRC. This changes the CTS by removing the specifics regarding distribution of the report to the NRC.

10 CFR 50.4 provides distribution requirements for written communications to the NRC. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.4. This change is designated as administrative because it does not result in technical changes to the CTS.

A07 SQN Unit 1 CTS 6.9.1.14.a requires, in part that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, to be listed. TVA has received approval to change the list of approved documents used to determine the core operating limits. This changes the CTS by revising the list of approved documents to those approved in License Amendment 331 before it has been implemented at SQN Unit 1.

Sequoyah Unit 1 and Unit 2 Page 2 of 4 Enclosure 2, Volume 16, Rev. 0, Page 227 of 270

Enclosure 2, Volume 16, Rev. 0, Page 228 of 270 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS This change is acceptable because this change was approved by License Amendment 331/324 [Unit 1/Unit 2] in September of 2012 by letter titled, "Sequoyah Nuclear Plant, Units 1 and 2 Issuance of Amendments to Revise the Technical Specification to allow use of Areva Advanced W17 High Thermal Performance Fuel (TS-SQN-2011-07) (TAC NOS. ME6538 and ME6539)"

(ADAMS Accession No. ML12249A394). This amendment was effective as of its date of issuance, to be implemented on Unit 1 prior to startup from Unit 1 fall 2013 refueling outage and on Unit 2 prior to startup from Unit 2 fall 2012 refueling outage. SQN Unit 2 License Amendment 324 has been implemented on Unit 2 and is reflected in this license amendment request. Because the implementation of SQN Unit 1 License Amendment 331 is after the submittal of the SQN ITS conversion license amendment request, the values approved in License Amendment 331 are shown as being inserted. This change is designated as administrative because it does not result in technical changes to the CTS approved by the NRC.

MORE RESTRICTIVE CHANGES M01 The second paragraph of ITS 5.6.1 includes details required to be included in the Annual Radiological Environmental Operating Report. CTS 6.9.1 does not contain this level of detail. This changes the CTS by requiring additional detail to be included in the Annual Radiological Environmental Operating Report.

The purpose of the second paragraph of ITS 5.6.1 is to specify details to be included in the Annual Radiological Environmental Operating Report. This change is acceptable because the content requirements are consistent with the objectives outlined in the Offsite Dose Calculation Manual. This change is designated more restrictive because it adds new reporting requirements to the Technical Specifications.

M02 CTS 6.9.1.14.c states, in part, that the CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle. ITS 5.6.3.d states, in part, that the COLR shall be provided within 30 days of issuance for each reload cycle to the NRC. This changes the CTS by eliminating the allowance to wait until entering MODE 2 for the 30 day period to begin before requiring the COLR to be submitted to the NRC.

The purpose to CTS 6.9.1.14.c is to provide guidance on when the COLR is required to be submitted to the NRC. ITS 5.6.3.d provides similar guidance but requires the COLR to be submitted in less time than allowed by CTS. This change is acceptable because the ITS requirement for submission of the COLR continues to allow adequate time to process the submittal and is within the CTS requirements. This change is designated as more restrictive because less time is allowed in ITS to submit the COLR than is allowed in CTS.

RELOCATED SPECIFICATIONS None Sequoyah Unit 1 and Unit 2 Page 3 of 4 Enclosure 2, Volume 16, Rev. 0, Page 228 of 270

Enclosure 2, Volume 16, Rev. 0, Page 229 of 270 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 L01 (Category 1 -Relaxation of LCO Requirements) CTS 6.9.1.6 requires the Annual Radiological Environmental Operating Report to be submitted prior to May 1 of each year. ITS 5.6.1 requires the Annual Radiological Environmental Operating Report to be submitted by May 15 of each year. This changes the CTS by allowing additional time to submit this report each year.

The purpose of the due date for submitting the Annual Radiological Environmental Operating Report is to ensure that the report is provided in a reasonable period of time to the NRC for review. This change is acceptable because the report is still required to be provided to the NRC on or before May 15 and cover the previous calendar year, report completion and submittal is clearly not necessary to assure operation in a safe manner for the interval between May 1 and May 15. Additionally, there is no requirement for the NRC to approve the report. This change is designated as less restrictive because it allows more time to prepare and submit the report to the NRC.

Sequoyah Unit 1 and Unit 2 Page 4 of 4 Enclosure 2, Volume 16, Rev. 0, Page 229 of 270

Enclosure 2, Volume 16, Rev. 0, Page 230 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 230 of 270

Enclosure 2, Volume 16, Rev. 0, Page 231 of 270 CTS Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 6.9 5.6 Reporting Requirements 6.9.1 The following reports shall be submitted in accordance with 10 CFR 50.4.

6.9.1.6 5.6.1 Annual Radiological Environmental Operating Report

---

REVIEWER'S NOTE----------------------------------------

6.9.1 Note

[ A single submittal may be made for a multiple unit station. The submittal should 1 combine sections common to all units at the station. ]

6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation DOC L01 of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

DOC M01 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the 1 table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

6.9.1.8 5.6.2 Radioactive Effluent Release Report

---

REVIEWER'S NOTE----------------------------------------

6.9.1 Note [ A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with 1 separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. ]

6.9.1.8 The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

SEQUOYAH UNIT 1 Amendment XXX 2

Westinghouse STS 5.6-1 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 231 of 270

Enclosure 2, Volume 16, Rev. 0, Page 232 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 6.9.1.14 5.6.3 CORE OPERATING LIMITS REPORT 6.9.1.14

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[ The individual specifications that address core operating limits must be referenced here. ] INSERT 1 1 6.9.1.14.a b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

---

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical 3 Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ] INSERT 2 1

6.9.1.14.b c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.14.c d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. 2 within 30 days of 6.9.1.15 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT 6.9.1.15 a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature 1

limits must be referenced here. ] INSERT 3 SEQUOYAH UNIT 1 Amendment XXX 2

Westinghouse STS 5.6-2 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 232 of 270

Enclosure 2, Volume 16, Rev. 0, Page 233 of 270 ITS 5.6 1 INSERT 1

1. SL 2.1.1, "Reactor Core Safety Limits";

2. LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

3. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";

4. LCO 3.1.5, "Shutdown Bank Insertion Limits";

5. LCO 3.1.6, "Control Bank Insertion Limits";

6. LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(X, Y, Z))";

7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FH(X,Y))";

8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T and Overpower T Nominal Trip Setpoint denoted values;

10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits"; and

11. LCO 3.9.1, "Boron Concentration."

Insert Page 5.6-2a Enclosure 2, Volume 16, Rev. 0, Page 233 of 270

Enclosure 2, Volume 16, Rev. 0, Page 234 of 270 ITS 5.6 1 INSERT 2

1. BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2. BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3. BAW-10163P-A, Revision 0,"Core Operating Limit Methodology for Westinghouse-Designed PWRs," June 1989

4. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 2001

5. BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6. BAW-10186-A, Revision 2, "Extended Burnup Evaluation," June 2003

7. EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003

8. BAW-10241P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005

9. BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10. BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11. BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990 and

12. BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code" January 2004.

1 INSERT 3

1. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits";

2. LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System"; and

3. LCO 3.5.2, "ECCS - Operating".

Insert Page 5.6-2b Enclosure 2, Volume 16, Rev. 0, Page 234 of 270

Enclosure 2, Volume 16, Rev. 0, Page 235 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued) 6.9.1.15.a

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

---

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the PTLR will contain the complete identification for each of the Technical Specification referenced Topical 3 Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the NRC staff approval document by date.] INSERT 4 1 within 30 days of 2

6.9.1.15.b c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

---

REVIEWER'S NOTE----------------------------------------

The methodology for the calculation of the P-T limits for NRC approval should include the following provisions:

1. The methodology shall describe how the neutron fluence is calculated (reference new Regulatory Guide when issued).

2. The Reactor Vessel Material Surveillance Program shall comply with Appendix H to 10 CFR 50. The reactor vessel material irradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves.

3. Low Temperature Overpressure Protection (LTOP) System lift setting limits 3 for the Power Operated Relief Valves (PORVs), developed using NRC-approved methodologies may be included in the PTLR.

4. The adjusted reference temperature (ART) for each reactor beltline material shall be calculated, accounting for radiation embrittlement, in accordance with Regulatory Guide 1.99, Revision 2.

5. The limiting ART shall be incorporated into the calculation of the pressure and temperature limit curves in accordance with NUREG-0800 Standard Review Plan 5.3.2, Pressure-Temperature Limits.

6. LTOP arming temperature limit development methodology.

SEQUOYAH UNIT 1 Amendment XXX 2

Westinghouse STS 5.6-3 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 235 of 270

Enclosure 2, Volume 16, Rev. 0, Page 236 of 270 ITS 5.6 1 INSERT 4

1. Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves";

2. Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation"; and

3. Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

Insert Page 5.6-3 Enclosure 2, Volume 16, Rev. 0, Page 236 of 270

Enclosure 2, Volume 16, Rev. 0, Page 237 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

7. The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

8. Licensees who have removed two or more capsules should compare for each surveillance material the measured increase in reference temperature (RTNDT) to the predicted increase in RTNDT; where the predicted increase in 3 RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology.

3.3.10 ACTION 4 5.6.5 Post Accident Monitoring Report I

1 When a report is required by Condition B or F of LCO 3.3.[3], "Post Accident 5 Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 [ Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance 4 Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. ]

6.9.1.16 5.6.7 Steam Generator Tube Inspection Report 4 6

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include: 6 7

6.9.1.16.a a. The scope of inspections performed on each SG, 6.9.1.16.b b. Active degradation mechanisms found, 8 6.9.1.16.c c. Nondestructive examination techniques utilized for each degradation mechanism, SEQUOYAH UNIT 1 Amendment XXX 2

Westinghouse STS 5.6-4 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 237 of 270

Enclosure 2, Volume 16, Rev. 0, Page 238 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued) 4 6

6.9.1.16.d d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, 8 6.9.1.16.e e. Number of tubes plugged [or repaired] during the inspection outage for each active degradation mechanism, 6.9.1.16.f f. Total number and percentage of tubes plugged [or repaired] to date, 6.9.1.16.g g. The results of condition monitoring, including the results of tube pulls and in- 7 8 situ testing,

and 6.9.1.16.h [h. The effective plugging percentage for all plugging [and tube repairs] in each SG, and]

[i. Repair method utilized and the number of tubes repaired by each repair method.]

Amendment XXX SEQUOYAH UNIT 1 Westinghouse STS 5.6-5 Rev. 4.0 x Enclosure 2, Volume 16, Rev. 0, Page 238 of 270

Enclosure 2, Volume 16, Rev. 0, Page 239 of 270 CTS Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 6.9 5.6 Reporting Requirements 6.9.1 The following reports shall be submitted in accordance with 10 CFR 50.4.

6.9.1.6 5.6.1 Annual Radiological Environmental Operating Report

---

REVIEWER'S NOTE----------------------------------------

6.9.1 Note

[ A single submittal may be made for a multiple unit station. The submittal should 1 combine sections common to all units at the station. ]

6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation DOC L01 of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

DOC M01 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the 1 table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

6.9.1.8 5.6.2 Radioactive Effluent Release Report

---

REVIEWER'S NOTE----------------------------------------

6.9.1 Note [ A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with 1 separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. ]

6.9.1.8 The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

SEQUOYAH UNIT 2 Amendment XXX 2

Westinghouse STS 5.6-1 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 239 of 270

Enclosure 2, Volume 16, Rev. 0, Page 240 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 6.9.1.14 5.6.3 CORE OPERATING LIMITS REPORT 6.9.1.14

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[ The individual specifications that address core operating limits must be referenced here. ] INSERT 1 1 6.9.1.14.a b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

---

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical 3 Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ] INSERT 2 1

6.9.1.14.b c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.14.c d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. 2 within 30 days of 6.9.1.15 5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT 6.9.1.15 a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature 1

limits must be referenced here. ] INSERT 3 SEQUOYAH UNIT 2 Amendment XXX 2

Westinghouse STS 5.6-2 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 240 of 270

Enclosure 2, Volume 16, Rev. 0, Page 241 of 270 ITS 5.6 1 INSERT 1

1. SL 2.1.1, "Reactor Core Safety Limits";

2. LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

3. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";

4. LCO 3.1.5, "Shutdown Bank Insertion Limits";

5. LCO 3.1.6, "Control Bank Insertion Limits";

6. LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(X, Y, Z))";

7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FH(X,Y))";

8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," - Overtemperature T and Overpower T Nominal Trip Setpoint denoted values;

10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits"; and

11. LCO 3.9.1, "Boron Concentration."

Insert Page 5.6-2a Enclosure 2, Volume 16, Rev. 0, Page 241 of 270

Enclosure 2, Volume 16, Rev. 0, Page 242 of 270 ITS 5.6 1 INSERT 2

1. BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2. BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3. BAW-10163P-A, Revision 0,"Core Operating Limit Methodology for Westinghouse-Designed PWRs," June 1989

4. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 2001

5. BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6. BAW-10186-A, Revision 2, "Extended Burnup Evaluation," June 2003

7. EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003

8. BAW-10241P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005

9. BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10. BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11. BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990 and

12. BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code" January 2004.

1 INSERT 3

1. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits";

2. LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System"; and

3. LCO 3.5.2, "ECCS - Operating".

Insert Page 5.6-2b Enclosure 2, Volume 16, Rev. 0, Page 242 of 270

Enclosure 2, Volume 16, Rev. 0, Page 243 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued) 6.9.1.15.a

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

---

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the PTLR will contain the complete identification for each of the Technical Specification referenced Topical 3 Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the NRC staff approval document by date.] INSERT 4 1 within 30 days of 2

6.9.1.15.b c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

---

REVIEWER'S NOTE----------------------------------------

The methodology for the calculation of the P-T limits for NRC approval should include the following provisions:

1. The methodology shall describe how the neutron fluence is calculated (reference new Regulatory Guide when issued).

2. The Reactor Vessel Material Surveillance Program shall comply with Appendix H to 10 CFR 50. The reactor vessel material irradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves.

3. Low Temperature Overpressure Protection (LTOP) System lift setting limits 3 for the Power Operated Relief Valves (PORVs), developed using NRC-approved methodologies may be included in the PTLR.

4. The adjusted reference temperature (ART) for each reactor beltline material shall be calculated, accounting for radiation embrittlement, in accordance with Regulatory Guide 1.99, Revision 2.

5. The limiting ART shall be incorporated into the calculation of the pressure and temperature limit curves in accordance with NUREG-0800 Standard Review Plan 5.3.2, Pressure-Temperature Limits.

6. LTOP arming temperature limit development methodology.

SEQUOYAH UNIT 2 Amendment XXX 2

Westinghouse STS 5.6-3 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 243 of 270

Enclosure 2, Volume 16, Rev. 0, Page 244 of 270 ITS 5.6 1 INSERT 4

1. Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves";

2. Westinghouse Topical Report WCAP-15293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation"; and

3. Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

Insert Page 5.6-3 Enclosure 2, Volume 16, Rev. 0, Page 244 of 270

Enclosure 2, Volume 16, Rev. 0, Page 245 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

7. The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

8. Licensees who have removed two or more capsules should compare for each surveillance material the measured increase in reference temperature (RTNDT) to the predicted increase in RTNDT; where the predicted increase in 3 RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology.

3.3.10 ACTION 4 5.6.5 Post Accident Monitoring Report I

1 When a report is required by Condition B or F of LCO 3.3.[3], "Post Accident 5 Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 [ Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance 4 Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. ]

6.9.1.16 5.6.7 Steam Generator Tube Inspection Report 4 6

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include: 6 7

6.9.1.16.a a. The scope of inspections performed on each SG, 6.9.1.16.b b. Active degradation mechanisms found, 8 6.9.1.16.c c. Nondestructive examination techniques utilized for each degradation mechanism, SEQUOYAH UNIT 2 Amendment XXX 2

Westinghouse STS 5.6-4 Rev. 4.0 Enclosure 2, Volume 16, Rev. 0, Page 245 of 270

Enclosure 2, Volume 16, Rev. 0, Page 246 of 270 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued) 4 6

6.9.1.16.d d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, 8 6.9.1.16.e e. Number of tubes plugged [or repaired] during the inspection outage for each active degradation mechanism, 6.9.1.16.f f. Total number and percentage of tubes plugged [or repaired] to date, 6.9.1.16.g g. The results of condition monitoring, including the results of tube pulls and in- 7 8 situ testing,

and 6.9.1.16.h [h. The effective plugging percentage for all plugging [and tube repairs] in each SG, and]

[i. Repair method utilized and the number of tubes repaired by each repair method.]

Amendment XXX SEQUOYAH UNIT 2 Westinghouse STS 5.6-5 Rev. 4.0 x Enclosure 2, Volume 16, Rev. 0, Page 246 of 270

Enclosure 2, Volume 16, Rev. 0, Page 247 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.6, STEAM GENERATOR TUBE INSPECTION REPORT

1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant-specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This Note is not meant to be retained in the final version of the plant specific submittal.

4. ISTS 5.6.6 provides requirements for the Tendon Surveillance Report. The Containment design at SQN does not include pre-stressed concrete tendons.

Therefore, this report is not included in the SQN ITS, consistent with the current licensing basis. Subsequent Specifications are renumbered as a result of this deletion.

5. Changes made to reflect those changes made to ITS 3.3.3, "Post Accident Monitoring (PAM) Instrumentation."

6. Changes made to reflect those changes made to ITS 5.5.7, "Steam Generator (SG)

Program."

7. Sequoyah Unit 1 and 2 are not licensed for repair of SG tubes, so the bracketed allowance has been deleted.

8. These punctuation corrections have been made consistent with the Writers Guide for the Improved Standard Technical Specifications NEI 01-03, Section 5.1.3.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 247 of 270

Enclosure 2, Volume 16, Rev. 0, Page 248 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 248 of 270

Enclosure 2, Volume 16, Rev. 0, Page 249 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 249 of 270

, Volume 16, Rev. 0, Page 250 of 270 ATTACHMENT 7 ITS 5.7, HIGH RADIATION AREA , Volume 16, Rev. 0, Page 250 of 270

, Volume 16, Rev. 0, Page 251 of 270 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 251 of 270

Enclosure 2, Volume 16, Rev. 0, Page 252 of 270 A01 ITS ITS 5.7 ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM (DELETED) 5.7 6.12 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation 5.7.1.a a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

5.7.1.b b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures.

that includes specification of radiation dose rates in the immediate work area(s) A02 5.7.1.c c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

5.7.1.d d. Each individual or group entering such an area shall possess:

5.7.1.d.1 1. A radiation monitoring device that continuously displays radiation dose rates in the area; or 5.7.1.d.2 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 5.7.1.d.3 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 5.7.1.d.4

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 5.7.1.d.4.(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 5.7.1.d.4.(ii) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

5.7.1.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas.

This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

February 11, 2003 SEQUOYAH - UNIT 1 6-15 Amendment No. 42, 58, 74, 148, 152, 174, 178, 212, 233, 266, 281 Page 1 of 6 Enclosure 2, Volume 16, Rev. 0, Page 252 of 270

Enclosure 2, Volume 16, Rev. 0, Page 253 of 270 A01 ITS ITS 5.7 ADMINISTRATIVE CONTROLS 5.7.2 6.12.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation 5.7.2.a a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

5.7.2.a.1 1. All such door and gate keys shall be maintained under the administrative control of the shift manager, radiation protection manager, or his or her designee.

s and 5.7.2.a.2

2. Doors and gates shall remain locked except when needed for personnel or equipment access. during periods of entry or exit 5.7.2.b

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures. dose rates in the immediate work area(s) that includes specifications of A02 5.7.2.c c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

5.7.2.d d. Each individual or group entering such an area shall possess:

5.7.2.d.1

1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 5.7.2.d.2

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 5.7.2.d.3

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 5.7.2.d.3.(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or or 5.7.2.d.3.(ii) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

5.7.2.d.4 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

February 11, 2003 SEQUOYAH - UNIT 1 6-15a Amendment No. 281 Page 2 of 6 Enclosure 2, Volume 16, Rev. 0, Page 253 of 270

Enclosure 2, Volume 16, Rev. 0, Page 254 of 270 A01 ITS ITS 5.7 ADMINISTRATIVE CONTROLS 5.7.2.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas.

This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

5.7.2.f f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

February 11, 2003 SEQUOYAH - UNIT 1 6-15b Amendment No. 281 Page 3 of 6 Enclosure 2, Volume 16, Rev. 0, Page 254 of 270

Enclosure 2, Volume 16, Rev. 0, Page 255 of 270 A01 ITS ITS 5.7 ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM (DELETED 5.7 6.12 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 6.12.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation 5.7.1.a a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

5.7.1.b b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures.

that includes specification of radiation dose rates in the immediate work area(s) A02 5.7.1.c c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

5.7.1.d d. Each individual or group entering such an area shall possess:

5.7.1.d.1 1. A radiation monitoring device that continuously displays radiation dose rates in the area; or 5.7.1.d.2 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 5.7.1.d.3 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 5.7.1.d.4

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 5.7.1.d.4.(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 5.7.1.d.4.(ii) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

5.7.1.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas.

This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

February 11, 2003 SEQUOYAH - UNIT 2 6-16 Amendment No. 34, 50, 66, 134, 142, 165, 169, 202, 223, 257, 272 Page 4 of 6 Enclosure 2, Volume 16, Rev. 0, Page 255 of 270

Enclosure 2, Volume 16, Rev. 0, Page 256 of 270 A01 ITS ITS 5.7 ADMINISTRATIVE CONTROLS 5.7.2 6.12.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation 5.7.2.a a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

5.7.2.a.1 1. All such door and gate keys shall be maintained under the administrative control of the shift manager, radiation protection manager, or his or her designee.

s and 5.7.2.a.2

2. Doors and gates shall remain locked except when needed for personnel or equipment access. during periods of entry or exit 5.7.2.b

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent, associated radiation survey, and other appropriate radiation protection equipment and measures. dose rates in the immediate work area(s) that includes specifications of A02 5.7.2.c c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

5.7.2.d d. Each individual or group entering such an area shall possess:

5.7.2.d.1

1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 5.7.2.d.2

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 5.7.2.d.3

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 5.7.2.d.3.(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or or 5.7.2.d.3.(ii) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

5.7.2.d.4 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

February 11, 2003 SEQUOYAH - UNIT 2 6-16a Amendment No. 272 Page 5 of 6 Enclosure 2, Volume 16, Rev. 0, Page 256 of 270

Enclosure 2, Volume 16, Rev. 0, Page 257 of 270 A01 ITS ITS 5.7 ADMINISTRATIVE CONTROLS 5.7.2.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas.

This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

5.7.2.f f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

February 11, 2003 SEQUOYAH - UNIT 2 6-16b Amendment No. 272 Page 6 of 6 Enclosure 2, Volume 16, Rev. 0, Page 257 of 270

Enclosure 2, Volume 16, Rev. 0, Page 258 of 270 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA ADMINISTRATIVE CHANGES A01 In the conversion of the Sequoyah Nuclear Plant (SQN) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 4.0, "Standard Technical Specifications - Westinghouse Plants" (ISTS) and additional Technical Specification Task Force (TSTF) travelers included in this submittal.

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 6.12.1.b and CTS 6.12.2.b state, in part, access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent and associated radiation survey. ITS 5.7.1.b and ITS 5.7.2.b state, in part, that access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area (s). This changes the CTS by specifying the document equivalent to the RWP shall include specification of radiation dose rates in the immediate work area(s).

The purpose of CTS 6.12.1.b and CTS 6.12.2.b is to specify the controls needed to access high radiation areas. This change is acceptable because the additional wording that the RWP equivalent includes a specification of radiation dose rates in the immediate work area(s) clarifies the requirements of an RWP. This is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 258 of 270

Enclosure 2, Volume 16, Rev. 0, Page 259 of 270 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Enclosure 2, Volume 16, Rev. 0, Page 259 of 270

Enclosure 2, Volume 16, Rev. 0, Page 260 of 270 CTS High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 6.12 [ 5.7 High Radiation Area ] 1 As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

6.12.1 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation 6.12.1.a a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

6.12.1.b b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

6.12.1.c c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

6.12.1.d d. Each individual or group entering such an area shall possess:

6.12.1.d.1 1. A radiation monitoring device that continuously displays radiation dose rates in the area, or 6.12.1.d.2 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 6.12.1.d.3

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 6.12.1.d.4 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 6.12.1.d.4.(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.7-1 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 260 of 270

Enclosure 2, Volume 16, Rev. 0, Page 261 of 270 CTS High Radiation Area 5.7 5.7 High Radiation Area 6.12.1 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.12.1.d.4.(ii) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

6.12.1.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 or Centimeters from the Radiation Source of from any Surface Penetrated by the 2 Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation 6.12.2.a a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

6.12.2.a.1 1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection 2 manager, or his or her designees, and manager 6.12.2.a.2

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

6.12.2.b b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

6.12.2.c c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.7-2 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 261 of 270

Enclosure 2, Volume 16, Rev. 0, Page 262 of 270 CTS High Radiation Area 5.7 5.7 High Radiation Area 6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters 2 or from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) or 6.12.2.d

d. Each individual group entering such an area shall possess: 2 6.12.2.d.1 1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 6.12.2.d.2

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 6.12.2.d.3 3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 6.12.2.d.3.(i) (i) Be under surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 6.12.2.d.3.(ii) (ii) Be under surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

6.12.2.d.4 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displaces radiation dose rates in the area. 3 displays 6.12.2.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.7-3 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 262 of 270

Enclosure 2, Volume 16, Rev. 0, Page 263 of 270 CTS High Radiation Area 5.7 5.7 High Radiation Area 6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.12.2.f

f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

SEQUOYAH UNIT 1 Amendment XXX Westinghouse STS 5.7-4 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 263 of 270

Enclosure 2, Volume 16, Rev. 0, Page 264 of 270 CTS High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 6.12 [ 5.7 High Radiation Area ] 1 As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

6.12.1 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation 6.12.1.a a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

6.12.1.b b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

6.12.1.c c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

6.12.1.d d. Each individual or group entering such an area shall possess:

6.12.1.d.1 1. A radiation monitoring device that continuously displays radiation dose rates in the area, or 6.12.1.d.2 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 6.12.1.d.3

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or 6.12.1.d.4 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 6.12.1.d.4.(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.7-1 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 264 of 270

Enclosure 2, Volume 16, Rev. 0, Page 265 of 270 CTS High Radiation Area 5.7 5.7 High Radiation Area 6.12.1 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.12.1.d.4.(ii) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

6.12.1.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 or Centimeters from the Radiation Source of from any Surface Penetrated by the 2 Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation 6.12.2.a a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

6.12.2.a.1 1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection 2 manager, or his or her designees, and manager 6.12.2.a.2

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

6.12.2.b b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

6.12.2.c c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.7-2 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 265 of 270

Enclosure 2, Volume 16, Rev. 0, Page 266 of 270 CTS High Radiation Area 5.7 5.7 High Radiation Area 6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters 2 or from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) or 6.12.2.d

d. Each individual group entering such an area shall possess: 2 6.12.2.d.1 1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 6.12.2.d.2

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 6.12.2.d.3 3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 6.12.2.d.3.(i) (i) Be under surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 6.12.2.d.3.(ii) (ii) Be under surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

6.12.2.d.4 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displaces radiation dose rates in the area. 3 displays 6.12.2.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.7-3 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 266 of 270

Enclosure 2, Volume 16, Rev. 0, Page 267 of 270 CTS High Radiation Area 5.7 5.7 High Radiation Area 6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.12.2.f

f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

SEQUOYAH UNIT 2 Amendment XXX Westinghouse STS 5.7-4 Rev. 4.0 2 Enclosure 2, Volume 16, Rev. 0, Page 267 of 270

Enclosure 2, Volume 16, Rev. 0, Page 268 of 270 JUSTIFICATION FOR DEVIATIONS ITS 5.7, HIGH RADIATION AREA

1. The ISTS contains bracketed information and/or values that are generic to Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is inserted to reflect the current licensing basis.

2. Changes are made (additions, deletions, and/or changes) to the ISTS that reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Typographical/grammatical error corrected.

Sequoyah Unit 1 and Unit 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 268 of 270

Enclosure 2, Volume 16, Rev. 0, Page 269 of 270 Specific No Significant Hazards Considerations (NSHCs)

Enclosure 2, Volume 16, Rev. 0, Page 269 of 270

Enclosure 2, Volume 16, Rev. 0, Page 270 of 270 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.7, HIGH RADIATION AREA There are no specific No Significant Hazards Considerations for this Specification.

Sequoyah Unit 1 and 2 Page 1 of 1 Enclosure 2, Volume 16, Rev. 0, Page 270 of 270