Text

Attachment 1, Volume 14, San Onofre Nuclear Generating Station - Improved Technical Specifications Conversion - ITS Chapter 5.0, Administrative Controls
	ML11251A108
Person / Time
Site:	San Onofre
Issue date:	07/29/2011
From:	; Edison International Co, Southern California Edison Co
To:	; Office of Nuclear Reactor Regulation
References
	NUREG-1432, Rev. 3.0
	Download: ML11251A108 (209)
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Attachment 1, Volume 14, Rev. 0, Page 1 of 209 ATTACHMENT 1 VOLUME 14 SAN ONOFRE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Attachment 1, Volume 14, Rev. 0, Page 1 of 209

Attachment 1, Volume 14, Rev. 0, Page 2 of 209 LIST OF ATTACHMENTS

1. ITS 5.1 - Responsibility

2. ITS 5.2 - Organization

3. ITS 5.3 - Unit Staff Qualifications

4. ITS 5.4 - Procedures

5. ITS 5.5 - Programs and Manuals

6. ITS 5.7 - Reporting Requirements

7. ITS 5.8 - High Radiation Area Note: There is no ITS 5.6 Attachment 1, Volume 14, Rev. 0, Page 2 of 209

, Volume 14, Rev. 0, Page 3 of 209 ATTACHMENT 1 ITS 5.1, RESPONSIBILITY , Volume 14, Rev. 0, Page 3 of 209

, Volume 14, Rev. 0, Page 4 of 209 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 4 of 209

Attachment 1, Volume 14, Rev. 0, Page 5 of 209 ITS Responsibility A01 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility plant manager 5.1.1 5.1.1 The corporate officer with direct responsibility for the plant shall LA01 be responsible for overall unit operation and maintenance of Units 2 See ITS 5.2 and 3 at San Onofre Nuclear Generating Station, and all site support functions. He shall delegate in writing the succession to this M01 responsibility during his absence.

INSERT 1 5.1.2 5.1.2 The Shift Manager shall be responsible for the ultimate command A04 decision authority for all unit activities and operations which affect the safety of the plant, site personnel, and/or the general LA01 Not used. public. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant shall be reissued to all site/station personnel on an annual basis. A02 5.1.3 5.1.3 The Control Room Supervisor (CRS) shall be responsible for the Control Room command function. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant, shall be issued annually to all site/station A02 personnel. The confines of the Control Room Area shall be defined as depicted in the Licensee Controlled Specification (LCS). During any absence of the CRS from the Control Room Area while the Unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator's (SRO) license shall be designated to assume the Control A03 Room command function. During any absence of the CRS from the Control Room Area while the Unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator's license shall be designated to assume the Control Room command function.

(continued)

SAN ONOFRE--UNIT 2 5.0-1 Amendment No. 207 Attachment 1, Volume 14, Rev. 0, Page 5 of 209

Attachment 1, Volume 14, Rev. 0, Page 6 of 209 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 5.0-1 Attachment 1, Volume 14, Rev. 0, Page 6 of 209

Attachment 1, Volume 14, Rev. 0, Page 7 of 209 ITS Responsibility A01 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility plant manager 5.1.1 5.1.1 The corporate officer with direct responsibility for the plant shall LA01 be responsible for overall unit operation and maintenance of Units 2 See ITS 5.2 and 3 at San Onofre Nuclear Generating Station, and all site support functions. He shall delegate in writing the succession to this M01 responsibility during his absence.

INSERT 1 5.1.2 5.1.2 The Shift Manager shall be responsible for the ultimate command A04 decision authority for all unit activities and operations which affect the safety of the plant, site personnel, and/or the general LA01 Not used. public. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant shall be reissued to all site/station personnel on an annual basis.

A02 5.1.3 5.1.3 The Control Room Supervisor (CRS) shall be responsible for the Control Room command function. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant, shall be issued annually to all site/station A02 personnel. The confines of the Control Room Area shall be defined as depicted in the Licensee Controlled Specification (LCS). During any absence of the CRS from the Control Room Area while the Unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator's (SRO) license shall be designated to assume the Control A03 Room command function. During any absence of the CRS from the Control Room Area while the Unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator's license shall be designated to assume the Control Room command function.

(continued)

SAN ONOFRE--UNIT 3 5.0-1 Amendment No. 199 Attachment 1, Volume 14, Rev. 0, Page 7 of 209

Attachment 1, Volume 14, Rev. 0, Page 8 of 209 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 5.0-1 Attachment 1, Volume 14, Rev. 0, Page 8 of 209

Attachment 1, Volume 14, Rev. 0, Page 9 of 209 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY ADMINISTRATIVE CHANGES A01 In the conversion of the San Onofre Nuclear Generating Station (SONGS)

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-1432, Rev. 3.0, "Standard Technical Specifications-Combustion Engineering Plants" (ISTS) and additional approved 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 5.1.2 states, in part, "A management directive to this effect, signed by the corporate officer with direct responsibility for the plant shall be reissued to all site/station personnel on an annual basis." CTS 5.1.3 states, in part, "A management directive to this effect, signed by the corporate officer with direct responsibility for the plant, shall be issued annually to all site/station personnel.

The confines of the Control Room Area shall be defined as depicted in the Licensee Controlled Specification (LCS)." This statement refers to the Control Room Supervisor (CRS) being responsible for the Control Room command function. ITS 5.1.1 and ITS 5.1.3 do not include these requirements. This changes the CTS by deleting the requirement to issue these management directives annually and the reference to what and where the control room area encompasses.

The purpose of CTS 5.1.2 and CTS 5.1.3 is to specify the responsibilities of the Shift Manager and Control Room Supervisor. Since the corporate officer is responsible for overall unit operation, as directed by CTS 5.1.1, then there is no need for these requirements in the Technical Specifications. In addition, the control room command function requirement has not changed. Furthermore, the location that describes what the control room area encompasses (i.e., the LCS) is also not required in the ITS since this description is more properly located in plant requirements (like the LCS) and the ITS does not need to specifically delineate where the description is located. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 5.1.3 uses the term "Senior Reactor Operator" and "Reactor Operator." ITS 5.1.3 uses the term "Senior Operator" and "Operator." This changes the CTS by modifying the terms for the two licensed individuals.

The new terms are being used since they are consistent with 10 CFR 55.4 and 10 CFR 50.54(m). These two regulations define the licensed individuals as "Senior Operator" and "Operator," not "Senior Reactor Operator" and "Reactor Operator." Therefore, this change is acceptable and is designated as administrative because it does not result in a technical change to the CTS.

A04 CTS 5.1.2 states, "The Shift Manager shall be responsible for the ultimategeneral public." This portion is being relocated to the UFSAR as described in DOC LA01. CTS 5.1.2 also states, "A management directive to this effectannual basis." This portion is being deleted as described in DOC A02.

San Onofre Unit 2 and 3 Page 1 of 3 Attachment 1, Volume 14, Rev. 0, Page 9 of 209

Attachment 1, Volume 14, Rev. 0, Page 10 of 209 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY With these changes, all CTS 5.1.2 has either been deleted or relocated. ITS 5.1.2 will state "Not used." This changes the CTS by replacing the current wording with "Not used."

The purpose of CTS 5.1.2 is to specify the responsibilities of the Shift Manager and Control Room Supervisor. These requirements are not being included in the ITS as justified in DOC A02 and LA01. The proposed change adds, "Not Used,"

to replace the previous CTS 5.1.2 wording. This change is acceptable because retaining the numbering consistent with the CTS will avoid the unnecessary administrative burden of changing Section numbers in plant procedures. This change is designated as administrative because it does not technically change the Specifications.

MORE RESTRICTIVE CHANGES M01 ITS 5.1.1 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. The CTS does not include this requirement.

This changes the CTS by adding an approved requirement for the plant manager (see DOC LA01 for the title change to plant manager) with direct responsibility for the plant or his designee.

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

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 5.1.1 provides the responsibilities of the corporate officer. CTS 5.1.2 provides the responsibilities of the shift manager.

ITS 5.1.1 provides the responsibilities of the plant manager. This changes the CTS by moving the specific organizational titles to the UFSAR and replacing it with a generic title.

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 San Onofre Unit 2 and 3 Page 2 of 3 Attachment 1, Volume 14, Rev. 0, Page 10 of 209

Attachment 1, Volume 14, Rev. 0, Page 11 of 209 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY allowance to relocate the specific 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.

Furthermore, both CTS and ITS 5.2.1.a require the plant-specific titles of those personnel fulfilling the responsibilities for the positions delineated in these Technical Specifications to be documented in the UFSAR. Also, this type of change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because information relating to meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None San Onofre Unit 2 and 3 Page 3 of 3 Attachment 1, Volume 14, Rev. 0, Page 11 of 209

Attachment 1, Volume 14, Rev. 0, Page 12 of 209 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 12 of 209

Attachment 1, Volume 14, Rev. 0, Page 13 of 209 U2/U3 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.

1

2. The ANSI Standard shall be the same ANSI Standard referenced in Section 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.

5.1.1 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.

The plant manager or his designee shall approve, prior to implementation, each 5.1.2 Not used. proposed test, experiment or modification to systems or equipment that affect nuclear safety.

4 Control Room Supervisor (CRS) 5.1.3 5.1.2 The [Shift Supervisor (SS)] shall be responsible for the control room command 2 CRS function. During any absence of the [SS] from the control room while the unit is 3

in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator 3 (SRO) license shall be designated to assume the control room command CRS 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 3 license shall be designated to assume the control room command function.

CEOG STS 5.1-1 Rev. 3.0, 03/31/04 5 San Onofre -- Draft Amendment XXX Attachment 1, Volume 14, Rev. 0, Page 13 of 209

Attachment 1, Volume 14, Rev. 0, Page 14 of 209 JUSTIFICATION FOR DEVIATIONS ITS 5.1, RESPONSIBILITY

1. The Reviewers Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This 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 all Combustion Engineering vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

3. The terms in 10 CFR 55.4 and 10 CFR 50.54(m) are "Senior Operator" and "Operator," not "Senior Reactor Operator" and "Reactor Operator."

4. ITS 5.1.2 is being added as a "not used" Section so the subsequent Section 5.1 numbering (ITS 5.1.2 was changed to ITS 5.1.3) can remain consistent with the CTS Section 5.1 numbering. The numbering changes are being made to be consistent with the CTS to avoid the unnecessary administrative burden of changing Section numbers in plant procedures.

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

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 14 of 209

Attachment 1, Volume 14, Rev. 0, Page 15 of 209 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 15 of 209

Attachment 1, Volume 14, Rev. 0, Page 16 of 209 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY There are no specific NSHC discussions for this Specification.

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 16 of 209

, Volume 14, Rev. 0, Page 17 of 209 ATTACHMENT 2 ITS 5.2, ORGANIZATION , Volume 14, Rev. 0, Page 17 of 209

, Volume 14, Rev. 0, Page 18 of 209 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 18 of 209

Attachment 1, Volume 14, Rev. 0, Page 19 of 209 ITS Organization A01 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 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 the safety of the nuclear power plant.

5.2.1.a a. Lines of authority, responsibility, and communication shall be established and defined for 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 organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These relationships, including the plant-specific titles of those personnel fulfilling the responsibilities for the positions delineated in these Technical Specifications, are documented in the UFSAR.

plant manager 5.2.1.b b. The corporate officer with direct responsibility for the LA01 plant shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

5.2.1.c c. A specified corporate officer (or officers) 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.

5.2.1.d 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.

(continued)

SAN ONOFRE--UNIT 2 5.0-2 Amendment No. 207 Attachment 1, Volume 14, Rev. 0, Page 19 of 209

Attachment 1, Volume 14, Rev. 0, Page 20 of 209 ITS Organization 5.2 5.2 Organization (continued) 5.2.2 5.2.2 UNIT STAFF The unit staff organization shall include the following:

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 when 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.

Not used.

b. At least one licensed Reactor Operator (RO) shall be in the 5.2.2.b Control Room when fuel is in the reactor. In addition, while A02 the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Reactor Operator (SRO) shall be in the Control Room Area.

and 5.2.2.g 5.2.2.c c. Shift crew composition may be less than the minimum L01 requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

5.2.2.d d. 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Deleted 5.2.2.e operations manager or assistant operations manager LA01 5.2.2.f f. The Manager, Plant Operations (at time of appointment), Shift Managers, and Control Room Supervisors shall hold a Senior LA02 Reactor Operator's license.

An individual A04 5.2.2.g g. The Shift Technical Advisor (STA) shall provide advisory unit operations technical support to the Shift Manager in the areas of thermal shift crew hydraulics, reactor engineering, and plant analysis with A03 regard to the safe operation of the unit. The STA shall have This individual a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and analysis of the plant for transients and accidents.

meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift (continued)

SAN ONOFRE--UNIT 2 5.0-3 Amendment No. 221 Attachment 1, Volume 14, Rev. 0, Page 20 of 209

Attachment 1, Volume 14, Rev. 0, Page 21 of 209 This Page Intentionally Left Blank SAN ONOFRE--UNIT 2 5.0-4 Amendment No.207 Attachment 1, Volume 14, Rev. 0, Page 21 of 209

Attachment 1, Volume 14, Rev. 0, Page 22 of 209 ITS Responsibility A01 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The corporate officer with direct responsibility for the plant shall 5.2.1.b be responsible for overall unit operation and maintenance of Units 2 See ITS 5.1 and 3 at San Onofre Nuclear Generating Station, and all site support functions. He shall delegate in writing the succession to this responsibility during his absence.

5.1.2 The Shift Manager shall be responsible for the ultimate command decision authority for all unit activities and operations which affect the safety of the plant, site personnel, and/or the general public. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant shall be reissued to all site/station personnel on an annual basis.

5.1.3 The Control Room Supervisor (CRS) shall be responsible for the Control Room command function. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant, shall be issued annually to all site/station personnel. The confines of the Control Room Area shall be defined as depicted in the Licensee Controlled Specification (LCS). During any absence of the CRS from the Control Room Area while the Unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator's (SRO) license shall be designated to assume the Control Room command function. During any absence of the CRS from the Control Room Area while the Unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator's license shall be designated to assume the Control Room command function.

(continued)

SAN ONOFRE--UNIT 2 5.0-1 Amendment No. 207 Attachment 1, Volume 14, Rev. 0, Page 22 of 209

Attachment 1, Volume 14, Rev. 0, Page 23 of 209 ITS Organization A01 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 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 the safety of the nuclear power plant.

5.2.1.a a. Lines of authority, responsibility, and communication shall be established and defined for 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 organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These relationships, including the plant-specific titles of those personnel fulfilling the responsibilities for the positions delineated in these Technical Specifications, are documented in the UFSAR.

plant manager 5.2.1.b b. The corporate officer with direct responsibility for the plant LA01 shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

5.2.1.c c. A specified corporate officer (or officers) 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.

5.2.1.d 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.

(continued)

SAN ONOFRE--UNIT 3 5.0-2 Amendment No. 199 Attachment 1, Volume 14, Rev. 0, Page 23 of 209

Attachment 1, Volume 14, Rev. 0, Page 24 of 209 ITS Organization 5.2 5.2 Organization (continued) 5.2.2 5.2.2 UNIT STAFF The unit staff organization shall include the following:

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 when 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.

Not used.

b. At least one licensed Reactor Operator (RO) shall be in the 5.2.2.b Control Room when fuel is in the reactor. In addition, while A02 the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Reactor Operator (SRO) shall be in the Control Room Area.

and 5.2.2.g 5.2.2.c , c. Shift crew composition may be less than the minimum L01 requirement of 10 CFR 50.54(m)(2)(I) and 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

5.2.2.d d. 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Deleted 5.2.2.e operations manager or assistant operations manager LA01 5.2.2.f f. The Manager, Plant Operations (at time of appointment), Shift Managers, and Control Room Supervisors shall hold a Senior LA02 Reactor Operator's license.

An individual A04 5.2.2.g g. The Shift Technical Advisor (STA) shall provide advisory unit operations technical support to the Shift Manager in the areas of thermal shift crew hydraulics, reactor engineering, and plant analysis with A03 regard to the safe operation of the unit. The STA shall have This individual a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and analysis of the plant for transients and accidents.

meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift (continued)

SAN ONOFRE--UNIT 3 5.0-3 Amendment No. 214 Attachment 1, Volume 14, Rev. 0, Page 24 of 209

Attachment 1, Volume 14, Rev. 0, Page 25 of 209 This Page Intentionally Left Blank SAN ONOFRE--UNIT 3 5.0-4 Amendment No.214 Attachment 1, Volume 14, Rev. 0, Page 25 of 209

Attachment 1, Volume 14, Rev. 0, Page 26 of 209 ITS Responsibility A01 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The corporate officer with direct responsibility for the plant shall 5.2.1.b be responsible for overall unit operation and maintenance of Units 2 See ITS 5.1 and 3 at San Onofre Nuclear Generating Station, and all site support functions. He shall delegate in writing the succession to this responsibility during his absence.

5.1.2 The Shift Manager shall be responsible for the ultimate command decision authority for all unit activities and operations which affect the safety of the plant, site personnel, and/or the general public. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant shall be reissued to all site/station personnel on an annual basis.

5.1.3 The Control Room Supervisor (CRS) shall be responsible for the Control Room command function. A management directive to this effect, signed by the corporate officer with direct responsibility for the plant, shall be issued annually to all site/station personnel. The confines of the Control Room Area shall be defined as depicted in the Licensee Controlled Specification (LCS). During any absence of the CRS from the Control Room Area while the Unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator's (SRO) license shall be designated to assume the Control Room command function. During any absence of the CRS from the Control Room Area while the Unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator's license shall be designated to assume the Control Room command function.

(continued)

SAN ONOFRE--UNIT 3 5.0-1 Amendment No. 199 Attachment 1, Volume 14, Rev. 0, Page 26 of 209

Attachment 1, Volume 14, Rev. 0, Page 27 of 209 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION ADMINISTRATIVE CHANGES A01 In the conversion of the San Onofre Nuclear Generating Station (SONGS)

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-1432, Rev. 3.0, "Standard Technical Specifications-Combustion Engineering Plants" (ISTS) and additional approved 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 5.2.2.b requires that at least one licensed Reactor Operator (RO) shall be in the control room when fuel is in the reactor. It also requires that when the unit is in MODE 1, 2, 3 or 4 that at least one licensed Senior Reactor Operator (SRO) shall be in the Control Room Area. The ITS does not include these requirements. This changes the CTS by deleting these requirements.

10 CFR 50.54(m)(2)(iii) states "When a nuclear power unit is in operational mode other than cold shutdown or refueling, as defined by a unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be at the controls at all times." 10 CFR 50.54(m)(2)(iv) states "Each licensee shall have present, during alteration of the core of a nuclear power plant (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person." This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.54(m)(2)(iii) and 10 CFR 50.54(m)(2)(iv). This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 5.2.2.g states, in part, that the STA provides advisory technical support to the Shift Manager. Additionally it provides qualification requirements for the Shift Technical Advisor (STA), and requires the STA to have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and analysis of the plant for transients and accidents.

ISTS 5.2.2.f (ITS 5.2.2.g) requires the individual to provide advisory technical support to the unit operation shift crew and to meet the qualification requirements of the Commission Policy on Engineering Expertise on Shift for qualification requirements instead of listing the specific qualification requirements.

The purpose of the CTS 5.2.2.g STA requirements is to specify the minimum qualification requirements for the STA. This change is acceptable because the qualification requirements included in the Commission Policy Statement on Engineering Expertise on Shift (Generic Letter 86-04, dated February 13, 1986) encompass the current STA qualification requirements. This change is designated as administrative because it does not result in technical changes to the CTS.

San Onofre Unit 2 and 3 Page 1 of 4 Attachment 1, Volume 14, Rev. 0, Page 27 of 209

Attachment 1, Volume 14, Rev. 0, Page 28 of 209 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION A04 CTS 5.2.2.f uses the term "Senior Reactor Operator." ITS 5.2.2.f uses the term "Senior Operator." This changes the CTS by modifying the terms for the two licensed individuals.

The new term is being used since it is consistent with 10 CFR 55.4 and 10 CFR 50.54(m). These two regulations define the licensed individuals as "Senior Operator," not "Senior Reactor Operator." Therefore, this change is acceptable and is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 5.2.1.b provides the responsibilities of the corporate officer. CTS 5.2.2.f requires, in part, the Manager, Plant Operations to hold a Senior Reactor Operator's license (DOC LA02 discusses the requirement for the Shift Managers and the Control Room Supervisors to hold a Senior Reactor Operator's license). ITS 5.2.1.b provides the responsibilities of the plant manager and ITS 5.2.2.f discusses the requirement for the operations manager or assistant manager to hold a Senior Operator's license (the change in the license name from Senior Reactor Operator to Senior Operator is discussed in DOC A04). This changes the CTS by moving the specific organizational titles to the UFSAR and replacing them with a generic title.

The removal of this detail, which is 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 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. Furthermore, both CTS and ITS 5.2.1.a require the plant specific titles of those personnel fulfilling the responsibilities for the positions delineated in these Technical Specifications to be documented in the UFSAR. Also, this type of change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because information relating to meeting Technical Specification requirement is being removed from the Technical Specifications.

San Onofre Unit 2 and 3 Page 2 of 4 Attachment 1, Volume 14, Rev. 0, Page 28 of 209

Attachment 1, Volume 14, Rev. 0, Page 29 of 209 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 5.2.2.f requires, in part, that the Shift Manager and Control Room Supervisors hold a Senior Reactor Operator's license (DOC LA01 discusses the requirement for the Manager, Plant Operations to hold a Senior Reactor Operator's license). ITS 5.2.2.f does not require the Shift Managers and Control Room Supervisors to hold a Senior Reactor Operator's license. This changes CTS by moving the requirement for the Shift Managers and Control Room Supervisor to hold a Senior Operator license (the change in the license name from Senior Reactor Operator to Senior Operator is discussed in DOC A04) to the UFSAR.

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 requirement for shift supervision to hold Senior Operator licenses is contained in 10 CFR 50.54(m), and does not need to be repeated in the Technical Specifications. The relocation of the details of the shift supervisor personnel that are required to hold Senior Operator licenses to the UFSAR is acceptable considering the controls provided by regulations. Furthermore, both CTS and ITS 5.2.1.a require the plant specific titles of those personnel fulfilling the responsibilities for the positions delineated in these Technical Specifications to be documented in the UFSAR. Also, this type of change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e). This change is designated as a less restrictive removal of detail change because information relating to meeting Technical Specification requirement is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirement) CTS 5.2.2.g requires a Shift Technical Advisor (STA) to provide technical support to the operating crew. ITS 5.2.2.g includes the same requirement, but ITS 5.2.2.c allows the position to be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position. This changes the CTS by allowing the STA position to be vacant for a short time due to unexpected circumstances.

The purpose of CTS 5.2.2.g is to ensure an individual, trained in areas of thermal hydraulics, reactor engineering and plant analysis, is onsite to provide expertise to the plant with regard to these fields. However, under unusual circumstances, such as an unexpected and sudden illness of the onsite individual, the STA may not be available. This change allows a short time, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , to not meet the requirement, provided immediate action is taken to fill the position (e.g., call in a replacement STA). This allowance is similar to that allowed in CTS 5.2.2.c for an unexpected absence in the shift operating crew requirements. Therefore, since the time allowed is short, and immediate action to rectify the problem is required, this change is considered acceptable. This change is designated as less San Onofre Unit 2 and 3 Page 3 of 4 Attachment 1, Volume 14, Rev. 0, Page 29 of 209

Attachment 1, Volume 14, Rev. 0, Page 30 of 209 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION restrictive because a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowance is provided to not meet the STA position requirement.

San Onofre Unit 2 and 3 Page 4 of 4 Attachment 1, Volume 14, Rev. 0, Page 30 of 209

Attachment 1, Volume 14, Rev. 0, Page 31 of 209 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 31 of 209

Attachment 1, Volume 14, Rev. 0, Page 32 of 209 U2/U3 CTS Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 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.

5.2.1.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 UFSAR.

5.1.1, b. The plant manager shall be responsible for overall safe operation of the 5.2.1.b plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant, 3 (or officers) .

2 5.2.1.c 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 3 5.2.1.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.

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

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 control room from which a reactor is operating in MODES 1, 2, 3, or 4. ; 3

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

are required Two unit sites with both units shutdown or defueled require a total of three non- 4 licensed operators for the two units. ,

5.2.2.b b. Not used.

CEOG STS San Onofre -- Draft 5.2-1 Amendment XXX Rev. 3.0, 03/31/04 9 Attachment 1, Volume 14, Rev. 0, Page 32 of 209

Attachment 1, Volume 14, Rev. 0, Page 33 of 209 U2/U3 CTS Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) 7 5.2.2.c c b. Shift crew composition may be less than the minimum requirement of g TSTF-511-A 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.f for a period of time not to 3

, exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

7 5.2.2.d d 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

d. Administrative procedures shall be developed and implemented to limit the working hours of personnel who perform safety related functions 8 (e.g., [licensed Senior Reactor Operators (SROs), licensed Reactor Operators (ROs), health physicists, auxiliary operators, and key maintenance personnel]).

The controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.

Any deviation from the above guidelines shall be authorized in advance by the plant manager or the plant manager's designee, in accordance with approved administrative procedures, and with documentation of the basis for granting the deviation. Routine deviation from the working hour guidelines shall not be authorized.

Controls shall be included in the procedures to require a periodic independent review be conducted to ensure that excessive hours have not been assigned.

e. Not used. 7 5.2.2.e 5

5.2.2.f e. The operations manager or assistant operations manager shall hold an 7 f SRO license. (at time of appointment) 6 Senior Operator's 7

5.2.2.g f. An individual shall provide advisory technical support to the unit operations g 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.

CEOG STS San Onofre -- Draft 5.2-2 Amendment XXX Rev. 3.0, 03/31/04 9 Attachment 1, Volume 14, Rev. 0, Page 33 of 209

Attachment 1, Volume 14, Rev. 0, Page 34 of 209 JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION

1. The ISTS contains bracketed information and/or values that are generic to all Combustion Engineering vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

2. ISTS 5.2.1.c has been modified to allow more than one corporate officer to be responsible for these requirements. This information was previously approved in Amendment 127 for Unit 2 and Amendment 116 for Unit 3 (ML021990684).

3. Changes are made to use correct punctuation, correct typographical errors or to make corrections consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01.

4. 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. Since SONGS is a dual unit site, the requirement specified in the Reviewer's Note has been retained, consistent with the CTS requirements.

5. ISTS 5.2.2.e states that the operations manager or assistant operations manager shall hold an SRO license. ITS 5.2.2.f modifies this statement to allow the operations manager or assistant operations manager to have a Senior Operator's license (changed from SRO as discussed in JFD 6) at the time of appointment. This information was previously approved in Amendment 127 for Unit 2 and Amendment 116 for Unit 3 (ML021990684).

6. The generic positions have been used. Also, terms in 10 CFR 55.4 and 10 CFR 50.54(m) are "Senior Operator" and "Operator," not "Senior Reactor Operator" and "Reactor Operator."

7. The Specification number has been changed to be consistent with the Specification number in the SONGS CTS. SCE has decided not to renumber the CTS to be consistent with the ISTS because by doing so would result in the unnecessary administrative burden of changing TS numbers in plant procedures.

8. The working hour restrictions in ISTS 5.2.2.d have not been included since they have been deleted from the SONGS Units 2 and 3 CTS as documented in the NRC Safety Evaluation for Amendments 221 and 214, respectively, dated 10/20/09 (ADAMS Accession No. ML092880169). This is also consistent with TSTF-511.

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

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 34 of 209

Attachment 1, Volume 14, Rev. 0, Page 35 of 209 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 35 of 209

Attachment 1, Volume 14, Rev. 0, Page 36 of 209 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION There are no specific NSHC discussions for this Specification.

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 36 of 209

, Volume 14, Rev. 0, Page 37 of 209 ATTACHMENT 3 ITS 5.3, UNIT STAFF QUALIFICATIONS , Volume 14, Rev. 0, Page 37 of 209

, Volume 14, Rev. 0, Page 38 of 209 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 38 of 209

Attachment 1, Volume 14, Rev. 0, Page 39 of 209 Unit Staff Qualifications A01 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except a) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and b) multi-discipline supervisors who shall meet or exceed the qualifications listed below.

In addition, the Shift Technical Advisor shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

Multi-discipline supervisors shall meet or exceed the following requirements:

a. Education: Minimum of a high school diploma or equivalent.

b. Experience: Minimum of four years of related technical experience which shall include three years power plant experience of which one year is at a nuclear plant.

c. Training: Complete the multi-discipline supervisor training program.

INSERT 1 A02 SAN ONOFRE--UNIT 2 5.0-5 Amendment No. 207 Attachment 1, Volume 14, Rev. 0, Page 39 of 209

Attachment 1, Volume 14, Rev. 0, Page 40 of 209 5.3 INSERT 1 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Operator and a licensed Operator 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).

Insert Page 5.0-5 Attachment 1, Volume 14, Rev. 0, Page 40 of 209

Attachment 1, Volume 14, Rev. 0, Page 41 of 209 Unit Staff Qualifications A01 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except a) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and b) multi-discipline supervisors who shall meet or exceed the qualifications listed below.

In addition, the Shift Technical Advisor shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

Multi-discipline supervisors shall meet or exceed the following requirements:

a. Education: Minimum of a high school diploma or equivalent.

b. Experience: Minimum of four years of related technical experience which shall include three years power plant experience of which one year is at a nuclear plant.

c. Training: Complete the multi-discipline supervisor training program.

INSERT 1 A02 SAN ONOFRE--UNIT 3 5.0-5 Amendment No. 199 Attachment 1, Volume 14, Rev. 0, Page 41 of 209

Attachment 1, Volume 14, Rev. 0, Page 42 of 209 5.3 INSERT 1 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Operator and a licensed Operator 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).

Insert Page 5.0-5 Attachment 1, Volume 14, Rev. 0, Page 42 of 209

Attachment 1, Volume 14, Rev. 0, Page 43 of 209 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS ADMINISTRATIVE CHANGES A01 In the conversion of the San Onofre Nuclear Generating Station (SONGS)

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-1432, Rev. 3.0, "Standard Technical Specifications-Combustion Engineering Plants" (ISTS) and additional approved 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 ITS 5.3.2 states "For the purpose of 10 CFR 55.4, a licensed Senior Operator and a licensed Operator 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)." The CTS does not include such a statement. This changes the CTS by clarifying the functions Senior Operators and Operators perform (i.e.,

those described in 10 CFR 50.54(m).

This change is acceptable because it clarifies the existing relationship between the Technical Specifications and regulations regarding licensed Senior Operator and Operator qualification requirements. SONGS Units 2 and 3 are already required to meet 10 CFR 50.54(m) requirements, since it is a regulation. This change 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 San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 43 of 209

Attachment 1, Volume 14, Rev. 0, Page 44 of 209 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 44 of 209

Attachment 1, Volume 14, Rev. 0, Page 45 of 209 U2/U3 CTS Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 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 2 specifying individual position qualifications. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special qualification statements because of unique organizational structures.

5.3.1 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 1 INSERT 1 Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff].

5.3.2 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and 3 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).

CEOG STS 5.3-1 Rev. 3.0, 03/31/04 4 San Onofre -- Draft Amendment XXX Attachment 1, Volume 14, Rev. 0, Page 45 of 209

Attachment 1, Volume 14, Rev. 0, Page 46 of 209 5.3 1

INSERT 1 N18.1-1971 for comparable positions, except: a) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975; and b) multi-discipline supervisors who shall meet or exceed the qualifications listed below.

In addition, the Shift Technical Advisor shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

Multi-discipline supervisors shall meet or exceed the following qualifications:

a. Education: Minimum of a high school diploma or equivalent;

b. Experience: Minimum of four years of related technical experience which shall include three years power plant experience of which one year is at a nuclear plant; and

c. Training: Complete the multi-discipline supervisor training program.

Insert Page 5.3-1 Attachment 1, Volume 14, Rev. 0, Page 46 of 209

Attachment 1, Volume 14, Rev. 0, Page 47 of 209 JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS

1. The ISTS contains bracketed information and/or values that are generic to all Combustion Engineering vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

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

3. Terms in 10 CFR 55.4 ad 10 CFR 50.54(m) are "Senior Operator" and "Operator,"

not "Senior Reactor Operator" and "Reactor Operator."

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

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 47 of 209

Attachment 1, Volume 14, Rev. 0, Page 48 of 209 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 48 of 209

Attachment 1, Volume 14, Rev. 0, Page 49 of 209 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS There are no specific NSHC discussions for this Specification.

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 49 of 209

, Volume 14, Rev. 0, Page 50 of 209 ATTACHMENT 4 ITS 5.4, PROCEDURES , Volume 14, Rev. 0, Page 50 of 209

, Volume 14, Rev. 0, Page 51 of 209 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 51 of 209

Attachment 1, Volume 14, Rev. 0, Page 52 of 209 ITS A01 Procedures, Programs, and Manuals 5.5 4

5.0 ADMINISTRATIVE CONTROLS 5.5 Procedures, Programs, and Manuals 4

5.5.1 Procedures 5.4 5.5.1.1 Scope 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

5.4.1.a a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; 5.4.1.b b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; 5.4.1.c c. Quality assurance for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1, 1979; LA01 5.4.1.d d. Fire Protection Program implementation; and 5.4.1.e e. Programs, as specified in Specification 5.5.2.

5.4.1.f f. Modification of core protection calculator (CPC) addressable constants. These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with CPCs during reactor operation.

Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P, which has been determined to be applicable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.

(continued)

SAN ONOFRE--UNIT 2 5.0-7 Amendment No. 207 Attachment 1, Volume 14, Rev. 0, Page 52 of 209

Attachment 1, Volume 14, Rev. 0, Page 53 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 4

5.0 ADMINISTRATIVE CONTROLS 5.5 Procedures, Programs, and Manuals 4

5.5.1 Procedures 5.4 5.5.1.1 Scope 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

5.4.1.a a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; 5.4.1.b b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; 5.4.1.c c. Quality assurance for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1, 1979; LA01 5.4.1.d d. Fire Protection Program implementation; and 5.4.1.e e. Programs, as specified in Specification 5.5.2.

5.4.1.f f. Modification of core protection calculator (CPC) addressable constants. These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with CPCs during reactor operation.

Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P, which has been determined to be applicable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.

(continued)

SAN ONOFRE--UNIT 3 5.0-7 Amendment No. 116 Attachment 1, Volume 14, Rev. 0, Page 53 of 209

Attachment 1, Volume 14, Rev. 0, Page 54 of 209 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES ADMINISTRATIVE CHANGES A01 In the conversion of the San Onofre Nuclear Generating Station (SONGS)

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-1432, Rev. 3.0, "Standard Technical Specifications-Combustion Engineering Plants" (ISTS) and additional approved 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 LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 5.5.1.1.c requires written procedures to be established, implemented and maintained for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1, 1979. ITS 5.4.1.c does not include the Regulatory Guide references. This changes the CTS by moving the reference to the Regulatory Guide to the UFSAR.

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 for effluent and environmental monitoring. Also, this change is acceptable because these types of procedural details will be adequately controlled in the UFSAR. This change is designated as less restrictive removal of detail because references for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 54 of 209

Attachment 1, Volume 14, Rev. 0, Page 55 of 209 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 55 of 209

Attachment 1, Volume 14, Rev. 0, Page 56 of 209 U2/U3 CTS Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.5.1 5.4 Procedures 5.5.1.1 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

5.5.1.1.a a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, 2 5.5.1.1.b 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 2 5.5.1.1.c c. Quality assurance for effluent and environmental monitoring, 2 5.5.1.1.d d. Fire Protection Program implementation, and 2 5.5.1.1.e e. All programs specified in Specification 5.5. 1

.2 5.5.1.1.f [ f. Modification of core protection calculator (CPC) addressable constants. 3 These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with CPCs during reactor operation.

Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure,"

CEN-39(A)-P, which has been determined to be applicable to the facility.

Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval. ] 3 CEOG STS 5.4-1 Rev. 3.0, 03/31/04 4 San Onofre -- Draft Amendment XXX Attachment 1, Volume 14, Rev. 0, Page 56 of 209

Attachment 1, Volume 14, Rev. 0, Page 57 of 209 JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES

1. The Specification number has been changed to be consistent with the Specification number in the SONGS CTS. SCE has decided not to renumber the CTS to be consistent with the ISTS because by doing so would result in the unnecessary administrative burden of changing TS numbers in plant procedures.

2. Changes are made to use correct punctuation, correct typographical errors or to make corrections consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01.

3. The ISTS contains bracketed information and/or values that are generic to all Combustion Engineering vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the information/value is changed to reflect the current licensing basis.

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

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 57 of 209

Attachment 1, Volume 14, Rev. 0, Page 58 of 209 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 58 of 209

Attachment 1, Volume 14, Rev. 0, Page 59 of 209 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES There are no specific NSHC discussions for this Specification.

San Onofre Unit 2 and 3 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 59 of 209

, Volume 14, Rev. 0, Page 60 of 209 ATTACHMENT 5 ITS 5.5, PROGRAMS AND MANUALS , Volume 14, Rev. 0, Page 60 of 209

, Volume 14, Rev. 0, Page 61 of 209 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 61 of 209

Attachment 1, Volume 14, Rev. 0, Page 62 of 209 ITS TS Bases Control A01 5.4 Programs and Manuals 5.5 Programs and Manuals 5

5.0 ADMINISTRATIVE CONTROLS 5.5.2.2 5.4 Technical Specifications (TS) Bases Control 5.5.2.2 Program a.

5.4.1 Changes to the Bases of the TS shall be made under appropriate administrative controls.

5.4.2 Changes to the Bases may be made without prior NRC approval provided the changes do not require either of the following:

b.

a. A change in the TS incorporated in the license; or 1.

b. A change to the updated FSAR or Bases that requires NRC

2. approval pursuant to 10 CFR 50.59.

5.4.3 The Bases Control Program shall contain provisions to ensure that c.

the Bases are maintained consistent with the UFSAR.

5.4.4 Proposed changes that meet the criteria of (a) or (b) above shall be reviewed and approved by the NRC prior to implementation. Changes to

d. the Bases implemented without prior NRC approval shall be provided to the NRC within 6 months following every Unit 3 refueling, not to exceed 24 months. This schedule is consistent with SCEs submittal of UFSAR updates as allowed by the NRC approved exemption from 10 CFR 50.71(e) dated April 27, 1999.

SAN ONOFRE--UNIT 2 5.0-6 Amendment No. 127, 168,184 Attachment 1, Volume 14, Rev. 0, Page 62 of 209

Attachment 1, Volume 14, Rev. 0, Page 63 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Procedures, Programs, and Manuals 5.5.1 Procedures 5.5.1.1 Scope Written procedures shall be established, implemented, and maintained covering the following activities:

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

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Quality assurance for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1, 1979;

d. Fire Protection Program implementation; and See ITS 5.4 e. Programs, as specified in Specification 5.5.2.

f. Modification of core protection calculator (CPC) addressable constants. These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with CPCs during reactor operation.

Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P, which has been determined to be applicable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.

(continued)

SAN ONOFRE--UNIT 2 5.0-7 Amendment No. 127 Attachment 1, Volume 14, Rev. 0, Page 63 of 209

Attachment 1, Volume 14, Rev. 0, Page 64 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2 Programs and Manuals The following programs and manuals shall be established, implemented, and maintained.

5.5.2.1 5.5.2.1 Offsite Dose Calculation Manual (ODCM)

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;

b. The ODCM shall also contain the Radioactive Effluent Controls required by Specification 5.5.2.3 and the Radiological Environmental Monitoring programs required by the LCS, and descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and the Radioactive Effluent Release Report required by Specification 5.7.1.2 and Specification 5.7.1.3.

5.5.2.1.1 5.5.2.1.1 Licensee-initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s);

2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.106, 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.

3. Documentation of the fact that the change has been plant manager reviewed and found acceptable.

LA01

b. Shall become effective upon review and approval by the corporate officer with direct responsibility for the plant or designee. LA05 (continued)

SAN ONOFRE--UNIT 2 5.0-8 Amendment No. 207 Attachment 1, Volume 14, Rev. 0, Page 64 of 209

Attachment 1, Volume 14, Rev. 0, Page 65 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.1.1 5.5.2.1.1 Licensee-initiated changes to the ODCM: (continued)

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.

5.5.2.2 Deleted 5.5.2.3 5.5.2.3 Radioactive Effluent Controls Program This program conforming to 10 CFR 50.36a provides 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 operating procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

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;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 CFR 20, Appendix B, Table II, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM; (continued)

SAN ONOFRE--UNIT 2 5.0-9 Amendment No. 127,178 Attachment 1, Volume 14, Rev. 0, Page 65 of 209

Attachment 1, Volume 14, Rev. 0, Page 66 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.3 5.5.2.3 Radioactive Effluent Controls Program (continued)

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;

e. Determination of cumulative and projected 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;

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 percent of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table II, Column 1;

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;

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 greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the A03 Radioactive Effluents Controls Program surveillance frequency.

5.5.2.4 5.5.2.4 Component Cyclic or Transient Limit Program This program provides controls to track the UFSAR Table 3.9-1 cyclic and transient occurrences to ensure that components are maintained within the design limits.

(continued)

SAN ONOFRE--UNIT 2 5.0-10 Amendment No. 127 Attachment 1, Volume 14, Rev. 0, Page 66 of 209

Attachment 1, Volume 14, Rev. 0, Page 67 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.5 5.5.2.5 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. A07 Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the flywheels each 10 years.

5.5.2.6 5.5.2.6 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:

a. Identification of a sampling schedule for the critical parameters and control points for these parameters;

b. Identification of the procedures used to measure the values of the critical parameters; which shall include monitoring the discharge of the condensate M05

c. Identification of process sampling points; pumps for evidence of condenser inleakage

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off-control point chemistry conditions; and

f. A procedure identifying (a) the authority responsible for interpretation of data and (b) the sequence and timing of administrative events, required to initiate corrective action.

5.5.2.7 5.5.2.7 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Gaseous Radwaste System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The gaseous radioactivity quantities shall be determined following methodology comparable with Branch Technical Position (BTP)

ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures".

(continued)

SAN ONOFRE--UNIT 2 5.0-11 Amendment No.127170 Attachment 1, Volume 14, Rev. 0, Page 67 of 209

Attachment 1, Volume 14, Rev. 0, Page 68 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.7 5.5.2.7 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for the concentrations of hydrogen and oxygen in the Gaseous Radwaste 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);

and

b. A surveillance program to ensure that the quantity of radioactivity contained in each waste gas decay tank and fed into the gaseous radwaste vent system is less than the amount that would result in a whole body exposure of greater than or equal to 0.5 rem to any individual in the unrestricted area, in the event of an uncontrolled release of the tanks contents; and

c. A surveillance program to ensure that the quantity of radioactivity contained 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 overflows and surrounding area drains connected to the Liquid Waste Management System is less than the amount that would result in concentrations less than the limits of 10 CFR Part 20, Appendix B, Table II, 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.

5.5.2.8 5.5.2.8 Primary Coolant Sources Outside Containment Program 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 high pressure safety injection recirculation, the shutdown cooling system, the reactor coolant sampling system (post-accident sampling piping only until such time as a modification eliminates the post-accident piping as a potential leakage path), the containment spray system, the radioactive waste gas system (post-accident sampling return piping only until such time as a modification eliminates the post-accident piping as a potential leakage path), and the liquid radwaste (continued)

SAN ONOFRE--UNIT 2 5.0-12 Amendment No. 127,178 Attachment 1, Volume 14, Rev. 0, Page 68 of 209

Attachment 1, Volume 14, Rev. 0, Page 69 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.8 5.5.2.8 Primary Coolant Sources Outside Containment Program (continued) system (post-accident sampling return piping only until such time as a modification eliminates the post-accident piping as a potential leakage path). The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and least once per 24 months A08

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

The provisions of SR 3.0.2 are applicable. A09 5.5.2.9 5.5.2.9 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containment, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. Program itself is relocated to the LCS.

INSERT 1 5.5.2.10 5.5.2.10 Inservice Inspection and Testing Program M01 This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components and Code Class CC and MC components LA02 including applicable supports. The program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program itself is located in the LCS.

INSERT 2 5.5.2.11 5.5.2.11 Steam Generator (SG) Program M02 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:

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.

(continued)

SAN ONOFRE--UNIT 2 5.0-13 Amendment No. 178,204, 220 Attachment 1, Volume 14, Rev. 0, Page 69 of 209

Attachment 1, Volume 14, Rev. 0, Page 70 of 209 5.5 M01 INSERT 1 The program shall include baseline measurements prior to initial operations. The Tendon 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 10CFR 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.

M02 INSERT 2 The program shall include the following:

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 to

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and other normal and accelerated Frequencies specified in the Inservice Testing Program for performing inservice testing activities; as 2 years or less

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

Insert Page 5.0-13 Attachment 1, Volume 14, Rev. 0, Page 70 of 209

Attachment 1, Volume 14, Rev. 0, Page 71 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 5.5.2.11 Steam Generator (SG) Program (continued)

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.

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 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.

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 exceed 0.5 gpm per SG and 1 gpm through both SGs.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.

(continued)

SAN ONOFRE--UNIT 2 5.0-14 Amendment No. 140,204 Attachment 1, Volume 14, Rev. 0, Page 71 of 209

Attachment 1, Volume 14, Rev. 0, Page 72 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 5.5.2.11 Steam Generator (SG) Program (continued)

c. Provisions for SG tube repair criteria.

1. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 35% of the nominal tube wall thickness shall be plugged.

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 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.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. 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.

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.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

SAN ONOFRE--UNIT 2 5.0-15 Amendment No. 220 Attachment 1, Volume 14, Rev. 0, Page 72 of 209

Attachment 1, Volume 14, Rev. 0, Page 73 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued)

THIS PAGE INTENTIONALLY BLANK SAN ONOFRE--UNIT 2 5.0-16 Amendment No. 215, 220 Attachment 1, Volume 14, Rev. 0, Page 73 of 209

Attachment 1, Volume 14, Rev. 0, Page 74 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 5.5.2.12 Ventilation Filter Testing Program (VFTP)

This Program establishes the required testing of the Engineered Safety Feature filter ventilation system Control Room Emergency Air Cleanup System." The frequency of testing shall be in accordance with Regulatory Guide 1.52, Revision 2. As a minimum the VFTP program shall include the following:

5.5.2.12.a a. Inplace testing of the high efficiency particulate air (HEPA) filters to demonstrate acceptable penetration and system

< 0.05 % of 35,705 bypass when tested at the appropriate system flowrate in cfm M07 accordance with Regulatory Guide 1.52, Revision 2, and ANSI + 10%

N510-1975 (see Note 1); and 5.5.2.12.b b. Inplace testing of the charcoal adsorber to demonstrate

< 0.05 % of 35,705 acceptable penetration and system bypass when tested at the cfm appropriate system flowrate in accordance with Regulatory + 10% M07 Guide 1.52, Revision 2, and ANSI N510-1975 (see Note 1); and 5.5.2.12.c c. Laboratory testing of charcoal adsorber samples obtained in accordance with Regulatory Guide 1.52, Revision 2 and tested per the methodology of ASTM D3803-1989 at 30EC and 70%

relative humidity to show acceptable methyl iodide M07 penetration; and < 1%

5.5.2.12.d d. Testing to demonstrate the pressure drop across the combined of 35,705 cfm HEPA filters, the prefilters, and the charcoal adsorbers, when

+ 10% and a tested at the appropriate system flowrate. M07 5.5.2.12.a, Delta P of 8.3 5.2.2.12.b inches of water Note 1: Sample and injection points shall be qualified per ANSI N510-1975 unless manifolds have been qualified per ASME N510-1989.

HEPA testing will be conducted with DOP aerosol or suitable alternate.

(continued)

SAN ONOFRE--UNIT 2 5.0-17 Amendment No. 204,208 Attachment 1, Volume 14, Rev. 0, Page 74 of 209

Attachment 1, Volume 14, Rev. 0, Page 75 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 5.5.2.12 Ventilation Filter Testing Program (VFTP) (continued)

The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.

5.5.2.13 5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil. 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:

a. Acceptability of new fuel oil use 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 water and sediment content within limits.

b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance A04 Requirement 3.8.3.3; and,

c. Total particulate concentration of fuel oil is # 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A. LA03 5.5.2.14 Deleted The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies. L01 (continued)

SAN ONOFRE--UNIT 2 5.0-18 Amendment No. 189,204 Attachment 1, Volume 14, Rev. 0, Page 75 of 209

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SAN ONOFRE--UNIT 2 5.0-19 Amendment No. 140,204 Attachment 1, Volume 14, Rev. 0, Page 76 of 209

Attachment 1, Volume 14, Rev. 0, Page 77 of 209 ITS Procedures, Programs, and Manuals 5.5 A01 5.5 Procedures, Programs, and Manuals (continued)

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SAN ONOFRE--UNIT 2 5.0-20 Amendment No. 197,204 Attachment 1, Volume 14, Rev. 0, Page 77 of 209

Attachment 1, Volume 14, Rev. 0, Page 78 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.15 5.5.2.15 Containment Leakage Rate Testing Program 5.5.2.15.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 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 exception: M03 INSERT 3 NEI 94 1995, Section 9.2.3: The first Type A Test performed after the March 31, 1995 Type A Test shall be performed no later A02 than March 30, 2010.

5.5.2.15.b The calculated peak containment internal pressure related to the design basis loss-of-coolant accident, Pa, is 48.0 psig (Pa will conservatively be assumed to be equal to the calculated peak containment internal pressure for the design basis Main Steam Line Break (51.5 psig) for the purpose of containment testing in accordance with this Technical Specification).

The containment design pressure is 60 psig. M06 5.5.2.15.c The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

5.5.2.15.d Leakage rate acceptance criteria are:

5.5.2.15.d.1

a. The Containment overall 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 Type C tests and # 0.75 La for the Type A tests; 5.5.2.15.d.2 b. Air lock testing acceptance criteria are:

5.5.2.15.d.2.a)

1) Overall air lock leakage rate is # 0.05 La when tested at $ Pa.

5.5.2.15.d.2.b) 2) For each door, the leakage rate is # 0.01 La when pressurized to $ 9.0 psig.

SAN ONOFRE--UNIT 2 5.0-20a Amendment No. 198, 220 Attachment 1, Volume 14, Rev. 0, Page 78 of 209

Attachment 1, Volume 14, Rev. 0, Page 79 of 209 5.5 M03 INSERT 3

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.

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.

Insert Page 5.0-20a Attachment 1, Volume 14, Rev. 0, Page 79 of 209

Attachment 1, Volume 14, Rev. 0, Page 80 of 209 ITS ITS Procedures, Programs, and Manuals 5.5 A01 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.15 5.5.2.15 Containment Leakage Rate Testing Program (Continued)

The provisions of Surveillance Requirement 3.0.2 do not apply to the A06 test frequencies specified in the Containment Leakage Rate Testing 5.5.2.15.f Program. However, test frequencies specified in this Program may be extended consistent with the guidance provided in NEI 94-01, "Industry Guideline For Implementing Performance-Based Option Of 10CFR 50, Appendix J," as endorsed by Regulatory Guide 1.163.

Specifically, NEI 94-01 has these provisions for test frequencies extension:

1. Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals for recommended Type A testing may be extended by up to 15 months. This option should be used only in cases where refueling schedules have been changed to accommodate other factors.

2. Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals for the recommended surveillance frequency for Type B and Type C testing may be extended by up to 25 percent of the test interval, not to exceed 15 months.

5.5.2.15.e The provisions of Surveillance Requirement 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

5.5.2.16 5.5.2.16 Control Room Envelope Habitability Program 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 Air Cleanup System (CREACUS), 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 total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

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

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

(continued)

SAN ONOFRE--UNIT 2 5.0-20b Amendment No. 214 Attachment 1, Volume 14, Rev. 0, Page 80 of 209

Attachment 1, Volume 14, Rev. 0, Page 81 of 209 ITS Procedures, Programs, and Manuals A01 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.16 5.5.2.16 Control Room Envelope Habitability Program (Continued)

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 is an exception to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

Appropriate application of ASTM E-741 shall include the ability to take minor exceptions to the test methodology. These exceptions shall be documented in the test report.

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 CREACUS, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

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.

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.

(continued)

SAN ONOFRE--UNIT 2 5.0-20c Amendment No. 218 Attachment 1, Volume 14, Rev. 0, Page 81 of 209

Attachment 1, Volume 14, Rev. 0, Page 82 of 209 ITS Procedures, Programs, and Manuals 5.5 A01 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.17 5.5.2.17 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, which includes the following:

a. Actions to restore battery cells with float voltage

< 2.13 V, and

b. Actions to verify that the remaining cells are above 2.07 V when a battery cell or cells have been found less than 2.13 V, and

c. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates. M04 INSERT 4 SAN ONOFRE--UNIT 2 5.0-20d Amendment No. 218 Attachment 1, Volume 14, Rev. 0, Page 82 of 209

Attachment 1, Volume 14, Rev. 0, Page 83 of 209 5.5 M04 INSERT 4 5.5.2.18 Surveillance Frequency Control Program 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.

Insert Page 5.0-20d Attachment 1, Volume 14, Rev. 0, Page 83 of 209

Attachment 1, Volume 14, Rev. 0, Page 84 of 209 ITS Programs and Manuals SFDP A01 5.6 5.5 Programs and Manuals 5 5.0 ADMINISTRATIVE CONTROLS 5.6 Safety Function Determination Program (SFDP) 5.5.2.14 5.5.2.14 5.6.1 This program ensures loss of safety function is detected and entry into appropriate actions taken. Upon failure to meet two or more LCOs at A05 LCO 3.0.6 the same time, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to 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.

5.5.2.14.a 5.6.2 The SFDP shall contain the following:

a. Provisions for cross-train checks to ensure a loss of the 5.5.2.14.a.1 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 5.5.2.14.a.2 condition if a loss of function condition exists.

c. Provisions to ensure that an inoperable supported system's 5.5.2.14.a.3 Completion Time is not inappropriately extended as a result of multiple support system inoperabilities.

5.5.2.14.a.4 d. Other appropriate limitations and remedial or compensatory actions.

5.5.2.14.b 5.6.3 A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot A05 no concurrent loss of offsite power, or be performed. For the purpose of this program, a loss of safety no concurrent loss function may exist when a support system is inoperable, and:

of onsite diesel generator(s),

5.5.2.14.b.1 a. A required system redundant to system(s) supported by the inoperable support system is also inoperable (Case A); or LA04 5.5.2.14.b.2

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable (Case B); LA04 or 5.5.2.14.b.3 c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable (Case LA04 C).

(continued)

SAN ONOFRE--UNIT 2 5.0-21 Amendment No. 127 Attachment 1, Volume 14, Rev. 0, Page 84 of 209

Attachment 1, Volume 14, Rev. 0, Page 85 of 209 ITS 5.5 Programs and Manuals SFDP Programs and Manuals 5.6 5

5.6 Safety Function Determination Program (SFDP) 5.6.3 (continued)

Generic Example:

Train A Train B System i System i 7 Case C 9 9 LA04 System ii 7 (Support System) System ii 9 Inoperable 9 System iii System iii 7 Case A 9 9 System iv System iv 7 Case B 5.5.2.14.c 5.6.4 The Safety Function Determination Program 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 A05 system, the appropriate Conditions and Required Actions to enter are those of the support system.

(continued)

SAN ONOFRE--UNIT 2 5.0-22 Amendment No. 127 Attachment 1, Volume 14, Rev. 0, Page 85 of 209

Attachment 1, Volume 14, Rev. 0, Page 86 of 209 ITS TS Bases Control A01 5.4 5.5 Programs and Manuals Programs and Manuals 5

5.0 ADMINISTRATIVE CONTROLS 5.5.2.2 5.4 Technical Specifications (TS) Bases Control 5.5.2.2 Program a.

5.4.1 Changes to the Bases of the TS shall be made under appropriate administrative controls.

5.4.2 Changes to the Bases may be made without prior NRC approval provided the changes do not require either of the following:

b.

a. A change in the TS incorporated in the license; or 1.

b. A change to the updated FSAR or Bases that requires NRC

2. approval pursuant to 10 CFR 50.59.

5.4.3 The Bases Control Program shall contain provisions to ensure that c.

the Bases are maintained consistent with the UFSAR.

5.4.4 Proposed changes that meet the criteria of (a) or (b) above shall be reviewed and approved by the NRC prior to implementation. Changes to

d. the Bases implemented without prior NRC approval shall be provided to the NRC within 6 months following every Unit 3 refueling, not to exceed 24 months. This schedule is consistent with SCEs submittal of UFSAR updates as allowed by the NRC approved exemption for 10 CFR 50.71(e) dated April 27, 1999.

SAN ONOFRE--UNIT 3 5.0-6 Amendment No. 116, 159,175 Attachment 1, Volume 14, Rev. 0, Page 86 of 209

Attachment 1, Volume 14, Rev. 0, Page 87 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Procedures, Programs, and Manuals 5.5.1 Procedures 5.5.1.1 Scope Written procedures shall be established, implemented, and maintained covering the following activities:

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

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Quality assurance for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1, 1979;

d. Fire Protection Program implementation; and See ITS 5.4 e. Programs, as specified in Specification 5.5.2.

f. Modification of core protection calculator (CPC) addressable constants. These procedures shall include provisions to ensure that sufficient margin is maintained in CPC type I addressable constants to avoid excessive operator interaction with CPCs during reactor operation.

Modifications to the CPC software (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with the most recent version of "CPC Protection Algorithm Software Change Procedure," CEN-39(A)-P, which has been determined to be applicable to the facility. Additions or deletions to CPC addressable constants or changes to addressable constant software limit values shall not be implemented without prior NRC approval.

(continued)

SAN ONOFRE--UNIT 3 5.0-7 Amendment No. 116 Attachment 1, Volume 14, Rev. 0, Page 87 of 209

Attachment 1, Volume 14, Rev. 0, Page 88 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2 Programs and Manuals The following programs and manuals shall be established, implemented, and maintained.

5.5.2.1 5.5.2.1 Offsite Dose Calculation Manual (ODCM)

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;

b. The ODCM shall also contain the Radioactive Effluent Controls required by Specification 5.5.2.3 and the Radiological Environmental Monitoring programs required by the LCS, and descriptions of the information that should be included in the Annual Radiological Environmental Operating Report and the Radioactive Effluent Release Report required by Specification 5.7.1.2 and Specification 5.7.1.3.

5.5.2.1.1 5.5.2.1.1 Licensee-initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s);

2. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.106, 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.

3. Documentation of the fact that the change has been plant manager reviewed and found acceptable. LA01

b. Shall become effective upon review and approval by the corporate officer with direct responsibility for the plant or designee. LA05 (continued)

SAN ONOFRE--UNIT 3 5.0-8 Amendment No. 199 Attachment 1, Volume 14, Rev. 0, Page 88 of 209

Attachment 1, Volume 14, Rev. 0, Page 89 of 209 ITS A01 Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.1.1 5.5.2.1.1 Licensee-initiated changes to the ODCM: (continued)

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.

5.5.2.2 Deleted 5.5.2.3 5.5.2.3 Radioactive Effluent Controls Program This program conforming to 10 CFR 50.36a provides 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 operating procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

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;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 CFR 20, Appendix B, Table II, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM; (continued)

SAN ONOFRE--UNIT 3 5.0-9 Amendment No. 116,169 Attachment 1, Volume 14, Rev. 0, Page 89 of 209

Attachment 1, Volume 14, Rev. 0, Page 90 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.3 5.5.2.3 Radioactive Effluent Controls Program (continued)

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;

e. Determination of cumulative and projected 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;

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 percent of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table II, Column 1;

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;

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 greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Controls Program surveillance frequency. A03 5.5.2.4 5.5.2.4 Component Cyclic or Transient Limit Program This program provides controls to track the UFSAR Table 3.9-1 cyclic and transient occurrences to ensure that components are maintained within the design limits.

(continued)

SAN ONOFRE--UNIT 3 5.0-10 Amendment No. 116 Attachment 1, Volume 14, Rev. 0, Page 90 of 209

Attachment 1, Volume 14, Rev. 0, Page 91 of 209 ITS A01 Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.5 5.5.2.5 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel. A07 Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the flywheels each 10 years.

5.5.2.6 5.5.2.6 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:

a. Identification of a sampling schedule for the critical parameters and control points for these parameters;

b. Identification of the procedures used to measure the values of the critical parameters; which shall include monitoring the discharge of the condensate pumps for evidence of M05

c. Identification of process sampling points; condenser inleakage

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off-control point chemistry conditions; and

f. A procedure identifying (a) the authority responsible for interpretation of data and (b) the sequence and timing of administrative events, required to initiate corrective action.

5.5.2.7 5.5.2.7 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Gaseous Radwaste System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The gaseous radioactivity quantities shall be determined following methodology comparable with Branch Technical Position (BTP)

ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures".

(continued)

SAN ONOFRE--UNIT 3 5.0-11 Amendment No.116, 161 Attachment 1, Volume 14, Rev. 0, Page 91 of 209

Attachment 1, Volume 14, Rev. 0, Page 92 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.7 5.5.2.7 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for the concentrations of hydrogen and oxygen in the Gaseous Radwaste 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);

and

b. A surveillance program to ensure that the quantity of radioactivity contained in each waste gas decay tank and fed into the gaseous radwaste vent system is less than the amount that would result in a whole body exposure of greater than or equal to 0.5 rem to any individual in the unrestricted area, in the event of an uncontrolled release of the tanks contents; and

c. A surveillance program to ensure that the quantity of radioactivity contained 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 overflows and surrounding area drains connected to the Liquid Waste Management System is less than the amount that would result in concentrations less than the limits of 10 CFR Part 20, Appendix B, Table II, 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.

5.5.2.8 5.5.2.8 Primary Coolant Sources Outside Containment Program 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 high pressure safety injection recirculation, the shutdown cooling system, the reactor coolant sampling system (post-accident sampling piping only until such time as a modification eliminates the post-accident piping as a potential leakage path), the containment spray system, the radioactive waste gas system (post-accident sampling return piping only until such time as a modification eliminates the post-accident piping as a potential leakage path), and the liquid radwaste (continued)

SAN ONOFRE--UNIT 3 5.0-12 Amendment No. 116,169 Attachment 1, Volume 14, Rev. 0, Page 92 of 209

Attachment 1, Volume 14, Rev. 0, Page 93 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.8 5.5.2.8 Primary Coolant Sources Outside Containment Program (continued) system (post-accident sampling return piping only until such time as a modification eliminates the post-accident piping as a potential leakage path). The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and least once per 24 months A08

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

The provisions of SR 3.0.2 are applicable. A09 5.5.2.9 5.5.2.9 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containment, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. Program itself is relocated to the LCS. M01 INSERT 1 5.5.2.10 5.5.2.10 Inservice Inspection and Testing Program This program provides controls for inservice inspection of ASME Code LA02 Class 1, 2, and 3 components and Code Class CC and MC components including applicable supports. The program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program itself is located in the LCS. M02 INSERT 2 5.5.2.11 5.5.2.11 Steam Generator (SG) Program 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:

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.

(continued)

SAN ONOFRE--UNIT 3 5.0-13 Amendment No. 213 Attachment 1, Volume 14, Rev. 0, Page 93 of 209

Attachment 1, Volume 14, Rev. 0, Page 94 of 209 5.5 M01 INSERT 1 The program shall include baseline measurements prior to initial operations. The Tendon 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 10CFR 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.

M02 INSERT 2 The program shall include the following:

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 to

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and other normal and accelerated Frequencies specified in the Inservice Testing Program for performing inservice testing activities; as 2 years or less

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

Insert Page 5.0-13 Attachment 1, Volume 14, Rev. 0, Page 94 of 209

Attachment 1, Volume 14, Rev. 0, Page 95 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 5.5.2.11 Steam Generator (SG) Program (continued)

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.

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 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.

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 exceed 0.5 gpm per SG and 1 gpm through both SGs.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.

(continued)

SAN ONOFRE--UNIT 3 5.0-14 Amendment No. 132,196 Attachment 1, Volume 14, Rev. 0, Page 95 of 209

Attachment 1, Volume 14, Rev. 0, Page 96 of 209 A01 ITS Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 5.5.2.11 Steam Generator (SG) Program (continued)