Text

Attachment 1, Volume 16, Improved Technical Specifications Conversion (Chapter 5.0), Administrative Controls, Rev. 2, Summary of Changes
	ML102580706
Person / Time
Site:	Kewaunee
Issue date:	09/08/2010
From:	; Dominion, Dominion Energy Kewaunee
To:	; Office of Nuclear Reactor Regulation
References
	10-521, LAR 249, TAC ME2139
	Download: ML102580706 (172)
	v • d • e

Attachment 1, Volume 16, Rev. 2, Page i of i Summary of Changes ITS Chapter 5.0 Change Description Affected Pages The changes described in the KPS response to Pages 105, 106, and 112 question KAB-084 have been made. This change makes changes to the Setpoint Control Program.

In the DEK letter dated 8/18/2010, it was stated that Page 104 the Battery Monitoring and Maintenance Program change described in the last KPS response to GMW-007 would only be made in the clean typed ITS. This was stated since the Revision 1 version of ITC Chapter 5.0 had already been submitted prior to this response. However, since a new Revision 2 version of ITS Chapter 5.0 is being submitted (due to the change described above), this change is being submitted in this Revision 2 version for completeness of the ITS submittal.

A typographical error has been corrected in the ISTS Page 99 Markup for ITS 5.5.12. A Reference in ITS 5.5.12.d to Specification 5.5.12b has been changed to Specification 5.5.12.b (i.e., a period has been added between the 12 and the b).

Note that the black revision bars from Volume 16, Revision 1 are still present in Volume 16, Revision 2. The Revision 2 changes described above are shown by red revision bars.

Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page i of i

Attachment 1, Volume 16, Rev. 2, Page 1 of 171 ATTACHMENT 1 VOLUME 16 KEWAUNEE POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Revision 2 Attachment 1, Volume 16, Rev. 2, Page 1 of 171

Attachment 1, Volume 16, Rev. 2, Page 2 of 171 LIST OF ATTACHMENTS

1. ITS 5.1

2. ITS 5.2

3. ITS 5.3

4. ITS 5.4

5. ITS 5.5

6. ITS 5.6

7. ITS 5.7

8. Relocated/Deleted Current Technical Specifications Attachment 1, Volume 16, Rev. 2, Page 2 of 171

, Volume 16, Rev. 2, Page 3 of 171 ATTACHMENT 1 ITS 5.1, RESPONSIBILITY , Volume 16, Rev. 2, Page 3 of 171

, Volume 16, Rev. 2, Page 4 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 4 of 171

Attachment 1, Volume 16, Rev. 2, Page 5 of 171 ITS A01 ITS 5.1 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 5.1.1 a. The plant manager shall be responsible for overall plant operation and shall delegate in writing the succession of this responsibility during his absence.

5.1.1 b. The plant manager, or his designee, shall approve prior to implementation, each proposed test, experiment or modification to structures, systems or components that affect nuclear safety.

INSERT 1 M01 Amendment No. 193 TS 6.1-1 10/31/2007 Page 1 of 2 Attachment 1, Volume 16, Rev. 2, Page 5 of 171

Attachment 1, Volume 16, Rev. 2, Page 6 of 171 ITS ITS 5.1 M01 INSERT 1 5.1.2 The shift manager shall be responsible for the control room command function. During any absence of the shift manager from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Operator license shall be designated to assume the control room command function. During any absence of the shift manager from the control room while the unit is in MODE 5 or 6, an individual with an active Senior Operator license or Operator license shall be designated to assume the control room command function.

Insert Page 6.1-1 Page 2 of 2 Attachment 1, Volume 16, Rev. 2, Page 6 of 171

Attachment 1, Volume 16, Rev. 2, Page 7 of 171 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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

MORE RESTRICTIVE CHANGES M01 ITS 5.1.2 requires that the shift manager shall be responsible for the control room command function. During any absence of the shift manager from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Operator license shall be designated to assume the control room command function. During any absence of the shift manager from the control room while the unit is in MODE 5 or 6, an individual with an active Senior Operator license or Operator license shall be designated to assume the control room command function. The CTS does not include this requirement. This changes the CTS by adding an approved requirement for control room command.

The purpose of ITS 5.1.2 is to ensure that the control room command function is maintained. This change is acceptable because the additional requirements ensure that the individual assuming the control room command functions meets the appropriate qualification requirements. This change is designated as more restrictive because it adds qualification requirements for the designated individual that assumes the control room command function to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 7 of 171

Attachment 1, Volume 16, Rev. 2, Page 8 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 8 of 171

Attachment 1, Volume 16, Rev. 2, Page 9 of 171 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 1

this position. Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special titles because of unique organizational structures.

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

6.1.a, 5.1.1 The plant manager shall be responsible for overall unit operation and shall 6.1.b delegate in writing the succession to this responsibility during his absence.

The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. shift manager DOC 5.1.2 The [Shift Supervisor (SS)] shall be responsible for the control room command 2 M01 shift manager function. During any absence of the [SS] from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator 3

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

WOG STS 5.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 9 of 171

Attachment 1, Volume 16, Rev. 2, Page 10 of 171 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 Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current Technical Specifications.

3. Typographical error corrected. The terms in 10 CFR 55.4 and 50.54(m) are "Senior Operator" and "Operator."

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 10 of 171

Attachment 1, Volume 16, Rev. 2, Page 11 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 11 of 171

Attachment 1, Volume 16, Rev. 2, Page 12 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 12 of 171

, Volume 16, Rev. 2, Page 13 of 171 ATTACHMENT 2 ITS 5.2, ORGANIZATION , Volume 16, Rev. 2, Page 13 of 171

, Volume 16, Rev. 2, Page 14 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 14 of 171

Attachment 1, Volume 16, Rev. 2, Page 15 of 171 A01 ITS ITS 5.2 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY and have control over those onsite activities necessary for safe operation and maintenance of the plant M01 5.2.1.b a. The plant manager shall be responsible for overall plant operation and shall delegate in See ITS 5.1 writing the succession of this responsibility during his absence.

b. The plant manager, or his designee, shall approve prior to implementation, each See ITS proposed test, experiment or modification to structures, systems or components that 5.1 affect nuclear safety.

Amendment No. 193 TS 6.1-1 10/31/2007 Page 1 of 4 Attachment 1, Volume 16, Rev. 2, Page 15 of 171

Attachment 1, Volume 16, Rev. 2, Page 16 of 171 A01 ITS ITS 5.2 5.2 6.2 ORGANIZATION 5.2.1 a. Off-Site Staff 5.2.2.a The off-site organization for plant management and technical support shall be as described in the quality assurance program.

5.2.2 b. Facility Staff M01 5.2.2.a The plant organization shall be as described in the quality assurance program.

1. Each on-duty shift complement shall consist of at least: A02 A. One shift manager (SRO) L01 B. Two licensed reactor operators C. Two nuclear auxiliary operators 5.2.2.a D. Deleted 5.2.2.c E. One radiation technologist L02

2. While above COLD SHUTDOWN, the on-duty shift complement shall consist of the A02 personnel required by TS 6.2.b.1 and an additional SRO.

5.2.2.b, 3. In the event that one of the shift members becomes incapacitated due to illness or 5.2.2.c injury or the radiation technologist has to accompany an injured person to the hospital, reactor operations may continue with the reduced complement until a replacement arrives. In all but severe weather conditions, a replacement is required within two hours.

4. At least one licensed operator shall be in the control room when fuel is in the reactor.

5. Two licensed operators, one of which shall be an SRO, shall be present in the control room when the unit is in an operational MODE other than COLD SHUTDOWN or REFUELING. A02

6. REFUELING OPERATIONS shall be directed by a licensed SRO assigned to the REFUELING OPERATION who has no other concurrent responsibilities during the REFUELING OPERATION.

5.2.2.e 7. When the reactor is above the COLD SHUTDOWN condition, a qualified shift technical advisor shall be within 10 minutes of the control room. L03 5.2.1.a c. Plant-Specific Titles The plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be maintained in appropriate administrative documents.

Add proposed ITS 5.2.2.d M02 Amendment No. 193 TS 6.2-1 10/31/2007 Page 2 of 4 Attachment 1, Volume 16, Rev. 2, Page 16 of 171

Attachment 1, Volume 16, Rev. 2, Page 17 of 171 A01 ITS 5.2

d. Organizational Changes Changes not affecting safety may be made to the off-site and facility staff organizations. A03 Such changes that are described in the Technical Specifications shall be reported to the Commission in the form of an application for license amendment within 60 days of the implementation of the change.

Amendment No. 193 TS 6.2-2 10/31/2007 Page 3 of 4 Attachment 1, Volume 16, Rev. 2, Page 17 of 171

Attachment 1, Volume 16, Rev. 2, Page 18 of 171 ITS A01 ITS 5.2 6.3 PLANT STAFF QUALIFICATIONS

a. Qualification of each member of the Plant Staff shall meet or exceed the minimum acceptable levels of ANSI N18.1-1971 for comparable positions, except for:

1. The radiation protection manager who shall meet or exceed the recommendation of Regulatory Guide 1.8, Revision 1-R, September 1975, or their equivalent as further clarified in Attachment 1 to the Safety Evaluation See ITS Report enclosed with Amendment No. 46 to Facility Operating License DPR-43. 5.3

2. The education and experience eligibility requirements for operator license applicants, changes thereto, shall be those previously reviewed and approved by the NRC, specifically those referenced in NRC Safety Evaluation letter dated October 2, 2003 (K-03-140).

5.2.2.e b. The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in the design of the Kewaunee Plant and A04 plant transient and accident analysis.

Amendment No. 193 TS 6.3-1 10/31/2007 Page 4 of 4 Attachment 1, Volume 16, Rev. 2, Page 18 of 171

Attachment 1, Volume 16, Rev. 2, Page 19 of 171 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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

A02 CTS 6.2.b.1.A and B, in part, require one shift manager (SRO) and two licensed reactor operators on duty at all times. CTS 6.2.b.2 states "When above COLD SHUTDOWN, the on-duty shift complement shall consist of the personnel required by TS 6.2.b.1 and an additional SRO." CTS 6.2.b.4 states "At least one licensed operator shall be in the control room when fuel is in the reactor." CTS 6.2.b.5 states, "Two licensed operators, one of which shall be an SRO, shall be present in the control room when the unit is in an operational MODE other than COLD SHUTDOWN or REFUELING." CTS 6.2.b.6 states "REFUELING OPERATIONS shall be directed by a licensed SRO assigned to the REFUELING OPERATION who has no other current responsibilities during the REFUELING OPERATION." The ITS does not include these requirements. This changes the CTS by deleting these requirements. A change to the number of licensed Operators required on duty is discussed in DOC L01.

The purpose of CTS 6.2.b.1.A and B, 6.2.b.2, 6.2.b.4 6.2.b.5, and 6.2.b.6 is to ensure the proper number of licensed Senior Operators and Operators are on duty as required by NRC regulations. 10 CFR 50.54(m)(2)(i) provides the requirements for the number of required on duty licensed Senior Operators and Operators when the unit is operating (i.e., ITS MODES 1, 2, 3, and 4) and not operating (i.e., ITS MODES 5 and 6 and defueled). For a single unit, two licensed Senior Operators and two licensed Operators are required when in MODE 1, 2, 3, or 4 and one licensed Senior Operator and one licensed Operator are required on duty when in MODE 5 or 6 or defueled. 10 CFR 50.54(m)(2)(iii) states "When a nuclear power unit is in an 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 alterations of the core of a nuclear power unit (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." The KPS Technical Specifications are an attachment to the Operating License, and the Operating License requires that KPS operates in conformity with the NRC rules and Regulations, which include the above 10 CFR 50.54(m) regulations.

This change is acceptable because the requirements deleted from the Technical Specifications (with the exception of the number of licensed Operators required on duty when in MODES 4 and 5 and defueled - which is discussed in DOC L01) are already required by 10 CFR 50.54(m)(2)(i), 10 CFR 50.54(m)(2)(iii), and Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 16, Rev. 2, Page 19 of 171

Attachment 1, Volume 16, Rev. 2, Page 20 of 171 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION 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 6.2.d states that changes not affecting safety may be made to the off-site and facility organizations and that such changes that are described in the Technical Specifications shall be reported to the NRC in the form of an application for license amendment within 60 days of the implementation of the change. This allowance is not being maintained in the ITS. This changes the CTS by deleting an allowance to make certain changes to the CTS prior to actually receiving a Technical Specification change.

The purpose of CTS 6.2.d was to allow Dominion Energy Kewaunee (DEK) to make changes to the unit staff organization, such as title changes, without waiting for prior NRC approval. However, License Amendment 193 (ADAMS Accession No. ML072880065) changed the CTS by deleting all plant specific titles and replacing them with generic titles, and specifying that the plant specific titles are in administrative documents. Furthermore, as stated in CTS 6.2.a and b, the off-site and facility organization is described in the quality assurance program. Thus, with the approval of License Amendment 193, this CTS 6.2.d allowance is unnecessary - no changes that do not affect safety are necessary, since all plant-specific titles have been removed from the CTS. Therefore, this change is acceptable and is considered administrative since KPS does not believe the current CTS 6.2.d allowance can be used at this time.

A04 CTS 6.3.b states that "The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in the design of the Kewaunee Plant and plant transients. ITS 5.2.2.e states that "An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift." This changes the CTS by referencing the Commission Policy Statement on Engineering Expertise on Shift for qualification requirements instead of specific qualifications.

The purpose of 6.3.b requirements is to specify the minimum qualification requirements for the shift technical advisor. 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 shift technical advisor qualification requirements. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.1.a describes that the plant manager is responsible for overall plant operation. CTS 6.2.a states that the off-site organization for plant management and technical support shall be as described in the quality assurance manual.

CTS 6.2.b states that the plant organization shall be as described in the quality assurance program. ITS 5.2.1 states:

Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 16, Rev. 2, Page 20 of 171

Attachment 1, Volume 16, Rev. 2, Page 21 of 171 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION "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.

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 quality assurance program;

b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;

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

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.

This changes the CTS by providing more details concerning the onsite and offsite organizations in the Technical Specification.

This change is acceptable because the details of the onsite and offsite organizations need to be included in the Technical Specification as delineated in NUREG-1431, Rev. 3.0. This change is more restrictive because information that was not included in the CTS is being added to the ITS.

M02 ITS 5.2.2.d states that " The operations manager or assistant operations manager shall hold an SRO license." CTS 6.2 does not include these requirements. This changes the CTS by adding new requirements concerning unit staff qualification to the CTS.

The purpose of ITS 5.2.2.d is to ensure the person who exercises control over the on shift crew has a Senior Operator license. This change is acceptable because it provides specific details concerning these issues. This change is designated as more restrictive because it provides details of the unit staff organization that was not included in the CTS.

Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 16, Rev. 2, Page 21 of 171

Attachment 1, Volume 16, Rev. 2, Page 22 of 171 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 6.2.b.1.B requires two licensed reactor operators to be on duty at all times as part of the shift complement. This requirement is being deleted from the Technical Specification since it is duplicative of 10 CFR 50.54(m)(2)(i). This is discussed in DOC A02.

However, 10 CFR 50.54(m)(2)(i) only requires one licensed Operator to be on duty when in MODE 5 or 6 or defueled. This changes the CTS by only requiring one licensed Operator to be on duty when in MODE 5 or 6 or defueled.

The purpose of CTS 6.2.b.1.B is to ensure the proper number of licensed Operators are on duty as required by NRC regulations. This change is acceptable since the NRC has already approved, as documented in 10 CFR 50.54(m)(2)(i), the minimum number of Licensed Operators required to be on duty in MODES 5 and 6 and defuled. This change is designated as a less restrictive change since the LCO requirements are less restrictive than are currently required in the CTS.

L02 (Category 1 - Relaxation of LCO Requirements) CTS 6.2.b.1.E requires one radiation technologist to be on duty at all times as part of the shift complement.

ITS 5.2.2.c only requires one radiation technologist to be on duty when there is fuel in the reactor vessel. This changes the CTS by not requiring a radiation technologist to be on duty when the reactor is defueled.

The purpose of CTS 6.2.b.1.E is to ensure the proper number of radiation technologists are duty when required. This change is acceptable since the NRC has already approved, as documented in ISTS NUREG-1431, Rev. 3.0 (as well as the other four ISTS NUREGs), allowing no radiation technologists to be on duty when the reactor is defueled. This change is designated as a less restrictive change since the LCO requirements are less restrictive in the ITS than in the CTS.

L03 (Category 1 - Relaxation of LCO Requirement) CTS 6.2.b.7 states that "When the reactor is above the COLD SHUTDOWN condition, a qualified shift technical advisor shall be within 10 minutes of the control room." ITS 5.2.2.e requires the shift technical advisor to be on duty when the unit is in MODE 1, 2, 3, or 4, but does not require the individual to be within 10 minutes of the control room. This changes the CTS by deleting the requirement that the shift technical advisor be within 10 minutes of the control room.

Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 16, Rev. 2, Page 22 of 171

Attachment 1, Volume 16, Rev. 2, Page 23 of 171 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION The purpose of CTS 6.2.b.7 is to delineate the requirements of the shift technical advisor (STA). The change is acceptable since it is not necessary to prescribe the amount of time the STA has to reach the control room. The NRC has already approved, as documented in ISTS NUREG-1431, Rev. 3.0 (as well as the other four ISTS NUREGs), that it is not necessary to delineate the amount of time the STA has to reach the control room when needed. This change is designated as a less restrictive change since the LCO requirements are less restrictive in the ITS than in the CTS.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 16, Rev. 2, Page 23 of 171

Attachment 1, Volume 16, Rev. 2, Page 24 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 24 of 171

Attachment 1, Volume 16, Rev. 2, Page 25 of 171 CTS Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations 6.2.a, Onsite and offsite organizations shall be established for unit operation and 6.2.b corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

DOC M01, a. Lines of authority, responsibility, and communication shall be defined and 6.2.c 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 2 those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the [FSAR/QA Plan], 1 2 5 INSERT 1 6.1.a b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant, 5 DOC M01 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 5 DOC M01 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 Unit Staff The unit staff organization shall include the following: contains if the 6.2.b.1.c 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 3 if the room from which a reactor is operating in MODES 1, 2, 3, or 4.

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

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

WOG STS 5.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 25 of 171

Attachment 1, Volume 16, Rev. 2, Page 26 of 171 5.2 00 _INSERT 1 2 5 quality assurance program. The plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be maintained in appropriate plant documents.

Insert Page 5.2-1 Attachment 1, Volume 16, Rev. 2, Page 26 of 171

Attachment 1, Volume 16, Rev. 2, Page 27 of 171 CTS Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) 6.2.b.3 b. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.f for a period of time not to e

, except in severe 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 6 weather conditions, shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

technologist 6.2.b.1.E, c. A radiation protection technician shall be on site when fuel is in the reactor. 7 6.2.b.3 except in severe 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 6 weather conditions, 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 (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.

TSTF-Any deviation from the above guidelines shall be authorized in advance by 511 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.

DOC M02 d e. The operations manager or assistant operations manager shall hold an TSTF-511 SRO license. 7 Senior Operator TSTF-e 6.2.b.7, f. An individual shall provide advisory technical support to the unit operations 8 511 6.3.b When the unit is in shift crew in the areas of thermal hydraulics, reactor engineering, and plant MODE 1, 2, 3, or 4 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.

WOG STS 5.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 27 of 171

Attachment 1, Volume 16, Rev. 2, Page 28 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION

1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current Technical Specifications.

2. CTS 6.2.c allows the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications to be maintained in appropriate plant documents, in lieu of the ISTS requirement that they be in the FSAR/QA Plan. This allowance was approved by the NRC as part of License Amendment 193, dated 10/31/07 (ADAMS Accession No. ML072880065). Therefore, ISTS 5.2.1.a has been changed to reflect this allowance.

3. Kewaunee Power Station includes only one unit. Therefore, the words in ITS 5.2.2.a have been modified to reflect a single unit site.

4. 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.

5. Typographical error corrected.

6. ITS 5.2.2.b has been revised to allow additional time for the shift crew composition to be below the minimum requirements of 10 CFR 50.54 "when severe weather conditions exist." ITS 5.2.2.c has been revised to allow additional time for a vacant radiation protection technician when severe weather conditions exits. This change is consistent with current Kewaunee Power Station (KPS) licensing requirements.

7. The generic positions have been used. Also, 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."

8. ISTS 5.2.2.f has been modified to require the shift technical advisor only in MODES 1, 2, 3, and 4, consistent with the current Technical Specifications Requirements.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 28 of 171

Attachment 1, Volume 16, Rev. 2, Page 29 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 29 of 171

Attachment 1, Volume 16, Rev. 2, Page 30 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 30 of 171

, Volume 16, Rev. 2, Page 31 of 171 ATTACHMENT 3 ITS 5.3, UNIT STAFF QUALIFICATIONS , Volume 16, Rev. 2, Page 31 of 171

, Volume 16, Rev. 2, Page 32 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 32 of 171

Attachment 1, Volume 16, Rev. 2, Page 33 of 171 ITS A01 ITS 5.3 6.3 PLANT STAFF QUALIFICATIONS 5.3.1 a. Qualification of each member of the Plant Staff shall meet or exceed the minimum acceptable levels of ANSI N18.1-1971 for comparable positions, except for:

5.3.1 1. The radiation protection manager who shall meet or exceed the recommendation of Regulatory Guide 1.8, Revision 1-R, September 1975, or their equivalent as further clarified in Attachment 1 to the Safety Evaluation Report enclosed with Amendment No. 46 to Facility Operating License DPR-43.

5.3.1 2. The education and experience eligibility requirements for operator license applicants, changes thereto, shall be those previously reviewed and approved by the NRC, specifically those referenced in NRC Safety Evaluation letter dated October 2, 2003 (K-03-140).

b. The shift technical advisor shall have a bachelor's degree or equivalent in a scientific or See ITS engineering discipline with specific training in the design of the Kewaunee Plant and 5.2 plant transient and accident analysis.

Add ITS proposed 5.3.2 A02 Amendment No. 193 TS 6.3-1 10/31/2007 Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 33 of 171

Attachment 1, Volume 16, Rev. 2, Page 34 of 171 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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. 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 Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 34 of 171

Attachment 1, Volume 16, Rev. 2, Page 35 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 35 of 171

Attachment 1, Volume 16, Rev. 2, Page 36 of 171 CTS Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 6.3 5.3 Unit Staff Qualifications

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

Minimum qualifications for members of the unit staff shall be specified by use of an overall qualification statement referencing an ANSI Standard acceptable to the NRC staff or by 1 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.

6.3.a 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 INSERT 1 Standard acceptable to the NRC staff]. [The staff not covered by Regulatory 2 Guide 1.8 shall meet or exceed the minimum qualifications of Regulations, Regulatory Guides, or ANSI Standards acceptable to NRC staff].

DOC 5.3.2 For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator (SRO) and A02 a licensed Reactor Operator (RO) are those individuals who, in addition to 3 meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m).

WOG STS 5.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 36 of 171

Attachment 1, Volume 16, Rev. 2, Page 37 of 171 5.3 2

INSERT 1 ANSI N18.1-1971 for comparable positions, except for:

a. The radiation protection manager who shall meet or exceed the recommendation of Regulatory Guide 1.8, Revision 1-R, September 1975, or their equivalent as further clarified in Attachment 1 to the NRC Safety Evaluation Report enclosed with Amendment No. 46, dated July 12, 1982.

b. The education and experience eligibility requirements for operator license applicants, changes thereto, shall be those previously reviewed and approved by the NRC, specifically those referenced in NRC Safety Evaluation letter for Amendment 170, dated October 2, 2003.

Insert Page 5.3-1 Attachment 1, Volume 16, Rev. 2, Page 37 of 171

Attachment 1, Volume 16, Rev. 2, Page 38 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS

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 Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current Technical Specification requirements.

3. Grammatical/typographical error corrected. The terms in 10 CFR 55.4 and 10 CFR 50.54(m) are "Senior Operator" and "Operator."

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 38 of 171

Attachment 1, Volume 16, Rev. 2, Page 39 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 39 of 171

Attachment 1, Volume 16, Rev. 2, Page 40 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATION There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 40 of 171

, Volume 16, Rev. 2, Page 41 of 171 ATTACHMENT 4 ITS 5.4, PROCEDURES , Volume 16, Rev. 2, Page 41 of 171

, Volume 16, Rev. 2, Page 42 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 42 of 171

Attachment 1, Volume 16, Rev. 2, Page 43 of 171 ITS A01 ITS 5.4 6.8 PROCEDURES 5.4.1 a. Written procedures and administrative policies shall be established, implemented and maintained that meet the requirements and recommendations of the quality assurance program. Add proposed 5.4.1.a, 5.4.1.b, and 5.4.1.d M01

b. Changes to procedures are made in accordance with the provisions of the quality LA01 assurance program.

c. Procedures are reviewed in accordance with the provisions of the quality assurance LA01 program.

Add proposed 5.4.1.e M02 Amendment No. 193 TS 6.8-1 10/31/2007 Page 1 of 2 Attachment 1, Volume 16, Rev. 2, Page 43 of 171

Attachment 1, Volume 16, Rev. 2, Page 44 of 171 ITS A01 ITS 5.4 6.16 RADIOLOGICAL EFFLUENTS 5.4.1 a. Written procedures shall be established, implemented and maintained covering the activities referenced below:

See CTS

1. Process Control Program (PCP) implementation 6.0 5.4.1.e 2. OFF-SITE DOSE CALCULATION MANUAL (ODCM) implementation 5.4.1.c 3. Quality Assurance Program for effluent and environmental monitoring

b. The following programs shall be established, implemented, and maintained:

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

A. Limitations on the OPERABILITY 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 ten times the See ITS concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001- 5.5 20.2402.

C. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM.

D. Limitations on the annual and quarterly doses or dose commitment to a MEMBER(S) OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50.

E. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

Amendment No. 186 TS 6.16-1 10/04/2005 Page 2 of 2 Attachment 1, Volume 16, Rev. 2, Page 44 of 171

Attachment 1, Volume 16, Rev. 2, Page 45 of 171 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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

MORE RESTRICTIVE CHANGES M01 CTS 6.8.a requires that written procedures and administrative policies shall be implemented and maintained that meet the requirements and recommendation of the quality assurance program. ITS 5.4.1.a requires that procedures be established, implemented, and maintained for the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

ITS 5.4.1.b requires that procedures be established, implemented, and maintained for the emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. ITS 5.4.1.d requires that procedures be established, implemented, and maintained for Fire Protection Program implementation. This changes the CTS by specifying new requirements for procedures in the CTS.

The purpose of ITS 5.4.1.a is to ensure that written procedures are established, implemented, and maintained for the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. The purpose ITS 5.4.1.b is to ensure that procedures be established, implemented, and maintained for the emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. The purpose of ITS 5.4.1.d is to ensure that procedures be established, implemented, and maintained for Fire Protection Program implementation. This change is acceptable since it specifies the actual procedures covered by this requirement in lieu of referencing the quality assurance program as the location that specifies the procedures. This change is designated as more restrictive because it imposes new requirements for procedures within the Technical Specifications.

M02 ITS 5.4.1.e requires that written procedures shall be established, implemented, and maintained for all programs specified in Specification 5.5. The CTS does not include this requirement for any programs, other than the ODCM as specified in CTS 6.16.a.2. This changes the CTS by adopting a new requirement for procedures to address all programs described in ITS 5.5.

The purpose of ITS 5.4.1.e is to ensure that written procedures are established, implemented, and maintained covering all programs specified in the ITS 5.5.

This change is considered acceptable because it required written procedures, including proper procedure control to address programs required by ITS 5.5.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 16, Rev. 2, Page 45 of 171

Attachment 1, Volume 16, Rev. 2, Page 46 of 171 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES This change is designated as more restrictive because it imposes new requirements for procedures within the Technical Specifications.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, USAR, ODCM, NFQAPD, CLRT Program, IST Program, or ISI Program) CTS 6.8.b requires that changes to procedures are made in accordance with the provisions of the quality assurance program. CTS 6.8.c requires that procedures are reviewed in accordance with the provisions of the quality assurance program.

ITS 5.4 does not include these requirements. This changes the CTS by moving these details of procedure changes and reviews to the NFQAPD.

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.

ITS 5.4.1 still retains the requirement for written procedures required by the Technical Specifications to be established, implemented, and maintained.

Regulations provide an adequate level of control for the affected review requirement. The requirements for establishment, maintenance, and implementation of procedures related to activities affecting quality are contained in 10 CFR 50, Appendix B, Criterion II and Criterion V and ANSI N18.7-1976 (ANS 3.2-1976). In accordance with these requirements, the NFQAPD includes adequate detail with respect to administrative control of procedures related to activities affecting quality and nuclear safety, including the review requirements associated with maintenance of these procedures. Also, this change is acceptable because these types of procedural details will be adequately controlled in the NFQAPD. Any changes to the NFQAPD are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because Technical Specification requirements related to procedure review and approval are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 16, Rev. 2, Page 46 of 171

Attachment 1, Volume 16, Rev. 2, Page 47 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 47 of 171

Attachment 1, Volume 16, Rev. 2, Page 48 of 171 CTS Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 6.8.a 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

DOC M01 a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, 1 DOC M01 b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as 2 stated in [Generic Letter 82-33], 3 1 6.16.a.3 c. Quality assurance for effluent and environmental monitoring, 1 DOC M01 d. Fire Protection Program implementation, and 1 DOC M02, 6.16.a.2

e. All programs specified in Specification 5.5.

WOG STS 5.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 48 of 171

Attachment 1, Volume 16, Rev. 2, Page 49 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES

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

2. Grammatical/typographical error corrected.

3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current Technical Specifications.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 49 of 171

Attachment 1, Volume 16, Rev. 2, Page 50 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 50 of 171

Attachment 1, Volume 16, Rev. 2, Page 51 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 51 of 171

, Volume 16, Rev. 2, Page 52 of 171 ATTACHMENT 5 ITS 5.5, PROGRAMS AND MANUALS , Volume 16, Rev. 2, Page 52 of 171

, Volume 16, Rev. 2, Page 53 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 53 of 171

Attachment 1, Volume 16, Rev. 2, Page 54 of 171 ITS ITS 5.5 A01

o. RADIOLOGICAL EFFLUENTS

1. MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not See ITS occupationally associated with the plant. This category does not include 1.0 employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

2. OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The ODCM shall contain the current methodology and parameters used in: (1) 5.5.1.a the calculation of off-site doses due to radioactive gaseous and liquid effluents, (2) the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and (3) the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain: (1) the Radioactive Effluent Controls and Radiological 5.5.1.b Environmental Monitoring Programs required by TS 6.16.b, and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by TS 6.9.b.1 and TS 6.9.b.2.

3. PROCESS CONTROL PROGRAM (PCP)

The PCP shall contain the current formulae, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid See CTS radioactive wastes, based on demonstrated processing of actual or simulated wet 6.0 solid wastes, will be accomplished in such a way as to ensure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71, Federal and State regulations, burial ground requirements, and other requirements governing the disposal of the radioactive waste.

4. SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

See ITS 1.0

5. UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

Amendment No. 162 TS 1.0-5 9/19/2002 Page 1 of 17 Attachment 1, Volume 16, Rev. 2, Page 54 of 171

Attachment 1, Volume 16, Rev. 2, Page 55 of 171 ITS ITS 5.5 A01 6.18 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

a. The ODCM shall be approved by the Commission prior to implementation. A02 5.5.1.c b. Licensee initiated changes to the ODCM:

5.5.1.c.1 1. Shall be documented and records of reviews performed shall be retained as LA01 required by the quality assurance program. This documentation shall contain:

5.5.1.c.1.a) A. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change.

5.5.1.c.1.b) B. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. plant manager M01 5.5.1.c.2 2. Shall become effective after review and acceptance by the PORC. LA01 5.5.1.c.3 3. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

The date the changes were made shall be indicated. In addition, a method such as redlining should be used to clearly identify the changes.

Amendment No. 193 TS 6.18-1 10/31/2007 Page 2 of 17 Attachment 1, Volume 16, Rev. 2, Page 55 of 171

Attachment 1, Volume 16, Rev. 2, Page 56 of 171 ITS ITS 5.5 A01 6.12 SYSTEM INTEGRITY 5.5.2 The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

Add proposed systems M02 5.5.2.a a. Provisions establishing preventive maintenance and periodic visual inspection requirements.

5.5.2.b b. Integrated leak test requirements for each system at a frequency not to exceed REFUELING cycle intervals.

The provisions of SR 3.0.2 are applicable. A03 Amendment No. 162 TS 6.12-1 09/19/2002 Page 3 of 17 Attachment 1, Volume 16, Rev. 2, Page 56 of 171

Attachment 1, Volume 16, Rev. 2, Page 57 of 171 ITS A01 ITS 5.5 6.16 RADIOLOGICAL EFFLUENTS

a. Written procedures shall be established, implemented and maintained covering the See ITS 5.4 activities referenced below:

See CTS

1. Process Control Program (PCP) implementation 6.0

2. OFF-SITE DOSE CALCULATION MANUAL (ODCM) implementation See ITS 5.4

3. Quality Assurance Program for effluent and environmental monitoring

b. The following programs shall be established, implemented, and maintained:

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

5.5.3.a A. Limitations on the OPERABILITY of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM.

5.5.3.b B. Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402.

5.5.3.c C. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM.

5.5.3.d D. Limitations on the annual and quarterly doses or dose commitment to a MEMBER(S) OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50.

5.5.3.e E. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days.

Amendment No. 186 TS 6.16-1 10/04/2005 Page 4 of 17 Attachment 1, Volume 16, Rev. 2, Page 57 of 171

Attachment 1, Volume 16, Rev. 2, Page 58 of 171 ITS A01 ITS 5.5 5.5.3.f F. Limitations on the OPERABILITY and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31 day period would exceed 2% of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50.

5.5.3.g G. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be limited to the following:

1. For noble gases: a dose rate 500 mrem/yr to the total body and a dose rate of 3000 mrem/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ.

5.5.3.h 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 Appendix I to 10 CFR Part 50.

5.5.3.i I. Limitations on the annual and quarterly doses to MEMBER(S) OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50.

5.5.3.j J. Limitations on the annual dose or dose commitment to any MEMBER(S) OF THE PUBLIC, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

5.5.3 The provisions of TS 4.0.b and 4.0.c are applicable to the Radioactive Effluents Controls Program surveillance frequency.

2. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide: (1) representative measurement of radioactivity in the highest potential exposure pathways, and (2) verification of the See CTS accuracy of the effluent monitoring program and modeling of environmental 6.0 exposure pathways. The program shall: (1) be contained in the ODCM (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

A. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.

Amendment No. 186 TS 6.16-2 10/04/2005 Page 5 of 17 Attachment 1, Volume 16, Rev. 2, Page 58 of 171

Attachment 1, Volume 16, Rev. 2, Page 59 of 171 ITS A01 ITS 5.5 4.2 ASME CODE CLASS IN-SERVICE INSPECTION AND TESTING APPLICABILITY Applies to in-service structural surveillance of the ASME Code Class components and supports and functional testing of pumps and valves.

OBJECTIVE To assure the continued integrity and operational readiness of ASME Code Class 1, 2, 3, and MC components.

SPECIFICATION 5.5.6 a. ASME Code Class 1, 2, 3, and MC Components and Supports LA02

1. In-service inspection of ASME Code Class 1, Class 2, Class 3, and Class MC components and supports shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by LA02 10 CFR 50.55a(g), except where relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The testing and surveillance of shock suppressors (snubbers) is detailed in TS 3.14 and TS 4.14.

5.5.6.a 2. In-service testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda as required by 10 CFR 50.55a(f), except where relief has been granted by the Commission pursuant to 10 CFR 50.55a(f)(6)(i). Add proposed ITS 5.5.6.a, ITS 5.5.6.b, ITS 5.5.6.c and ITS 5.5.6.d A04

3. Surveillance testing of pressure isolation valves:

a. Periodic leakage testing 1 on each valve listed in Table TS 3.1-2 shall be accomplished prior to entering the OPERATING mode after every time the plant See ITS is placed in the COLD SHUTDOWN condition for refueling, after each time the 3.4.14 plant is placed in a COLD SHUTDOWN condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair, or replacement work is performed.

(1)

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators if accomplished in accordance with approved procedures See ITS and supported by computations showing that the method is capable of demonstrating valve 3.4.14 compliance with the leakage criteria.

Amendment No. 189 TS 4.2-1 12/14/2006 Page 6 of 17 Attachment 1, Volume 16, Rev. 2, Page 59 of 171

Attachment 1, Volume 16, Rev. 2, Page 60 of 171 ITS ITS 5.5 A01 5.5.7 6.22 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:

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

5.5.7.b b. 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 150 gpd per SG.

3. The operational LEAKAGE performance criterion is specified in TS 3.1.d, RCS Operational LEAKAGE.

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

Amendment No. 188 TS 6.22-1 Revised by letter dated August 29, 2006 Page 7 of 17 Attachment 1, Volume 16, Rev. 2, Page 60 of 171

Attachment 1, Volume 16, Rev. 2, Page 61 of 171 ITS A01 ITS 5.5 5.5.7.d d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be 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.

5.5.7.e e. Provisions for monitoring operational primary to secondary LEAKAGE.

Amendment No. 188 TS 6.22-2 Revised by letter dated August 29, 2006 Page 8 of 17 Attachment 1, Volume 16, Rev. 2, Page 61 of 171

Attachment 1, Volume 16, Rev. 2, Page 62 of 171 ITS A01 ITS 5.5 6.15 SECONDARY WATER CHEMISTRY 5.5.8 The licensee shall implement a secondary water chemistry monitoring program. The intent of this program will be to control corrosion thereby inhibiting steam generator tube degradation. The secondary water chemistry program shall act as a guide for the chemistry M06 group in their routine as well as non-routine activities.

and low pressure turbine disc stress corrosion cracking M05 Add proposed 5.5.8.a, 5.5.8.b, 5.5.8.c, 5.5.8.d, M06 5.5.8.e, and 5.5.8.f Amendment No. 162 TS 6.15-1 09/19/2002 Page 9 of 17 Attachment 1, Volume 16, Rev. 2, Page 62 of 171

Attachment 1, Volume 16, Rev. 2, Page 63 of 171 A01 ITS ITS 5.5

3. Performance Requirements Add proposed 5.5.9 generic program statement and SR 3.0.2 and SR 3.0.3 applicability statement A05 A. The results of the in-place cold DOP and halogenated hydrocarbon tests at 5.5.9.a, 5.5.9.a design flows on HEPA filters and charcoal adsorber banks shall show 99%

5.5.9.b DOP removal and 99% halogenated hydrocarbon removal.

Add proposed design flow values M13 5.5.9.c B. The results of laboratory carbon sample analysis from the Shield Building Ventilation System and the Auxiliary Building Special Ventilation System carbon shall show 97.5% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C, 95% RH for the Shield Building Ventilation System and 30°C, 95% RH for the Auxiliary Building Special Ventilation System.

5.5.9.a, C. Fans shall operate within +/- 10% of design flow when tested.

5.5.9.b

d. If the internal pressure of the reactor containment vessel exceeds 2 psi, the condition See ITS shall be corrected within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or the reactor shall be placed in a subcritical condition. 3.6.4

e. The reactor shall not be taken above the COLD SHUTDOWN condition unless the See ITS 3.6.5 containment ambient temperature is > 40°F.

Amendment No. 190 TS 3.6-4 03/08/2007 Page 10 of 17 Attachment 1, Volume 16, Rev. 2, Page 63 of 171

Attachment 1, Volume 16, Rev. 2, Page 64 of 171 A01 ITS ITS 5.5 3.12 CONTROL ROOM POST-ACCIDENT RECIRCULATION SYSTEM APPLICABILITY Applies to the OPERABILITY of the Control Room Post-Accident Recirculation System.

OBJECTIVE To specify OPERABILITY requirements for the Control Room Post-Accident Recirculation System.

SPECIFICATION

a. The reactor shall not be made critical unless both trains of the Control Room Post-Accident Recirculation System are OPERABLE.

b. Both trains of the Control Room Post-Accident Recirculation System, including filters, See ITS 3.3.7 and shall be OPERABLE or the reactor shall be shut down within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , except that when 3.7.10 one of the two trains of the Control Room Post-Accident Recirculation System is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding 7 days.

c. During testing the system shall meet the following performance requirements:

Add proposed 5.5.9 generic program statement and SR 3.0.2 and SR 3.0.3 applicability statement A05

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design 5.5.9.a, flows on HEPA filter and charcoal adsorber banks shall show 99% DOP removal 5.5.9.b and 99% halogenated hydrocarbon removal.

Add proposed design flow values M13 5.5.9.c 2. The results of the laboratory carbon sample analysis from the Control Room Post-Accident Recirculation System carbon shall show 95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C, and 95% RH.

5.5.9.a, 3. Fans shall operate within +/-10% of design flow when tested.

5.5.9.b Amendment No. 152 TS 3.12-1 02/28/2001 Page 11 of 17 Attachment 1, Volume 16, Rev. 2, Page 64 of 171

Attachment 1, Volume 16, Rev. 2, Page 65 of 171 A01 ITS ITS 5.5 4.4 CONTAINMENT TESTS APPLICABILITY Applies to integrity testing of the steel containment, shield building, auxiliary building special ventilation zone, and the associated systems including isolation valves.

OBJECTIVE To verify that leakage from the containment system is maintained within allowable limits in accordance with 10 CFR Part 50, Appendix J.

SPECIFICATION

a. Integrated Leak Rate Tests (Type A)

Perform required visual examinations and leakage rate testing in accordance with the See ITS 3.6.1 Containment Leakage Rate Testing Program.

As a one-time exception to the Containment Leakage Rate testing Program, the first Type A test following the Type A test performed in April 1994 shall be required no later than October 2009.

b. Local Leak Rate Tests (Type B and C) See ITS 3.6.1, 3.6.2, and Perform required air lock, penetration, and containment isolation valve leakage testing 3.6.3 in accordance with the Containment Leakage Rate Testing Program.

c. Shield Building Ventilation System A06 5.5.9 1. At least once per operating cycle or once every 18 months, whichever occurs first, the following conditions shall be demonstrated:

5.5.9.d a. Pressure drop across the combined HEPA filters and charcoal adsorber banks is L02

< 10 inches of water and the pressure drop across any HEPA filter bank is

< 4 inches of water at the system design flow rate (+/-10%). M13 Add proposed design flow values

b. Automatic initiation of each train of the system. See ITS 3.6.10

c. Deleted Amendment No. 204 TS 4.4-1 04/27/2009 Page 12 of 17 Attachment 1, Volume 16, Rev. 2, Page 65 of 171

Attachment 1, Volume 16, Rev. 2, Page 66 of 171 A01 ITS ITS 5.5

2. Shield Building Ventilation System Filter Testing 5.5.9 a. The in-place DOP test for HEPA filters shall be performed (1) at least once per 18 months and (2) after each complete or partial replacement of a HEPA filter bank or after any maintenance on the system that could affect the HEPA bank bypass leakage.

5.5.9 b. The laboratory tests for activated carbon in the charcoal filters shall be performed (1) at least once per 18 months for filters in a standby status or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of filter operation, and (2) following painting, fire, or chemical release in any ventilation zone communicating with the system.

5.5.9 c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any maintenance on the system that could affect the charcoal adsorber bank bypass leakage.

See ITS

d. Each train shall be operated at least 15 minutes every month. 3.6.10 5.5.9.e

3. An air distribution test on these HEPA filter banks will be performed after any 5.5.9 maintenance or testing that could affect the air distribution within the systems. The 5.5.9.e test shall be performed at design flow rate (+/-10%). The results of the test shall show the air distribution is uniform within +/-20%.(1) M13 Add proposed design flow values

4. Each train shall be determined to be operable at the time of its periodic test if it produces measurable indicated vacuum in the annulus within 2 minutes after See ITS initiation of a simulated safety injection signal and obtains equilibrium discharge 3.6.8 conditions that demonstrate the Shield Building leakage is within acceptable limits.

(1)

In WPS letter of August 25, 1976 to Mr. Al Schwencer (NRC) from Mr. E. W. James, we relayed test results for flow distribution for tests performed in accordance with ANSI N510-1975. This standard refers to flow distribution tests performed upstream of filter assemblies.

Since the test results upstream of filters were inconclusive due to high degree of turbulence, A07 tests for flow distribution were performed downstream of filter assemblies with acceptable results (within 20%). The safety evaluation attached to Amendment 12 references our letter of August 25, 1976 and acknowledges acceptance of the test results.

Amendment No. 201 TS 4.4-2 12/30/2008 Page 13 of 17 Attachment 1, Volume 16, Rev. 2, Page 66 of 171

Attachment 1, Volume 16, Rev. 2, Page 67 of 171 A01 ITS ITS 5.5

d. Auxiliary Building Special Ventilation System 5.5.9.a, 1. Periodic tests of the Auxiliary Building Special Ventilation System, including the door 5.5.9.b, 5.5.9.c. interlocks, shall be performed in accordance with TS 4.4.c.1 through TS 4.4.c.3, 5.5.9.d, except for TS 4.4.c.2.d. See ITS 5.5.9.e 3.7.12

2. Each train of Auxiliary Building Special Ventilation System shall be operated at least 15 minutes every month.

See ITS

3. Each system shall be determined to be operable at the time of periodic test if it starts 3.7.12 with coincident isolation of the normal ventilation ducts and produces a measurable vacuum throughout the special ventilation zone with respect to the outside atmosphere.

e. Containment Vacuum Breaker System The power-operated valve in each vent line shall be tested during each refueling outage to demonstrate that a simulated containment vacuum of 0.5 psig will open the valve and See ITS a simulated accident signal will close the valve. The check and butterfly valves will be 3.6.1, 3.6.3, and leak tested in accordance with TS 4.4.b during each refueling, except that the pressure 3.6.9 will be applied in a direction opposite to that which would occur post-LOCA.

f. Containment Isolation Device Position Verification

1. When the reactor is greater than Cold Shutdown condition, verify each 36 inch containment purge and vent isolation valve is sealed closed every 31 days.

2. When the reactor is critical, verify each 2 inch containment vent isolation valve is closed every 31 days, except when the 2 inch containment vent isolation valves are open for pressure control, ALARA, or air quality considerations for personnel entry, or Surveillances that require the valves to be open. See ITS 3.6.3

3. Containment isolation manual valves and blind flanges shall be verified closed as specified in TS 4.4.f.3.a and TS 4.4.f.3.b, except as allowed by TS 4.4.f.3.c.

a. When greater than COLD SHUTDOWN, verify each containment isolation manual valve and blind flange that is located outside containment and required to be closed during accident conditions is closed every 31 days, except for containment isolation valves that are locked, sealed, or otherwise secured closed or open as allowed by TS 3.6.b.2.

Amendment No. 206 TS 4.4-3 06/01/2009 Page 14 of 17 Attachment 1, Volume 16, Rev. 2, Page 67 of 171

Attachment 1, Volume 16, Rev. 2, Page 68 of 171 A01 ITS ITS 5.5 4.17 CONTROL ROOM POSTACCIDENT RECIRCULATION SYSTEM APPLICABILITY Applies to testing and surveillance requirements for the Control Room Postaccident Recirculation System in TS 3.12.

OBJECTIVE To verify the performance capability of the Control Room Postaccident Recirculation System.

SPECIFICATION A06 5.5.9 a. At least once per operating cycle or once every 18 months, whichever occurs first, the following conditions shall be demonstrated:

L02 5.5.9.d

1. Pressure drop across the combined HEPA filters and charcoal adsorber banks is

< 6 inches of water and the pressure drop across any HEPA bank is < 4 inches of water at the system design flow rate (+/- 10%). M13 Add proposed design flow values

2. Automatic initiation of the system on a high radiation signal and a safety injection See ITS signal. 3.7.10 5.5.9 b. 1. The in-place DOP test for HEPA filters shall be performed (1) at least once per 18 months and (2) after each complete or partial replacement of a HEPA filter bank or after any maintenance on the system that could affect the HEPA bank bypass leakage.

5.5.9 2. The laboratory tests for activated carbon in the charcoal filters shall be performed (1) at least once per 18 months for filters in a standby status or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of filter operation, and (2) following painting, fire, or chemical release in any ventilation zone communicating with the system.

5.5.9 3. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any maintenance on the system that could affect the charcoal adsorber bank bypass leakage.

See ITS

4. Each train shall be operated at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months each month. 3.7.10 Amendment No. 137 TS 4.17-1 06/09/98 Page 15 of 17 Attachment 1, Volume 16, Rev. 2, Page 68 of 171

Attachment 1, Volume 16, Rev. 2, Page 69 of 171 ITS A01 ITS 5.5 6.21 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM 5.5 The Bases Control Program shall be established, implemented and maintained. This program provides a means for processing changes to the bases of these Technical 5.5.12 Specifications.

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

5.5.12.b b. Changes to bases may be made without prior NRC approval provided the changes do not require either of the following:

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

5.5.12.b.1 2. A change to the USAR or bases that requires NRC approval pursuant to 10 CFR 50.59.

5.5.12.d c. Proposed changes that meet the criteria of 6.21.b.1 and 6.21.b.2 above shall be reviewed and approved by the NRC prior to implementation.

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

5.5.12.d e. Changes to the bases implemented without prior NRC approval shall be provided to the NRC on a frequency not to exceed that of 10 CFR 50.71(e).

Amendment No. 163 TS 6.21-1 9/24/2002 Page 16 of 17 Attachment 1, Volume 16, Rev. 2, Page 69 of 171

Attachment 1, Volume 16, Rev. 2, Page 70 of 171 ITS A01 ITS 5.5 6.20 CONTAINMENT LEAKAGE RATE TESTING PROGRAM 5.5.14.a A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. The provisions of TS 4.0.b do not apply to the test frequencies specified A08 in the Containment Leakage Rate Testing Program. The provisions of TS 4.0.c are 5.5.14.e applicable to the Containment Leakage Rate Testing Program.

Add proposed calculated 5.5.14.b The peak calculated containment internal pressure for the design basis loss-of-coolant containment internal pressure accident is less than the containment internal test pressure, Pa. The maximum allowable and containment design pressure 5.5.14.c leakage rate (La) is 0.2 weight percent of the contained air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at the peak test pressure (Pa) of 46 psig.

M12 For penetrations which extend into the auxiliary building special ventilation zone, the combined leak rate from these penetrations shall not exceed 0.10La. For penetrations See ITS which are exterior to both the shield building and the auxiliary building special ventilation 3.6.3 zone, the combined leak rate from these penetrations shall not exceed 0.01La. If leak rates are exceeded, repairs and retest shall be performed to demonstrate reduction of the combined leak rate to these values.

5.5.14.d Leakage rate acceptance criteria:

5.5.14.d.1 a. The containment leakage rate acceptance criterion is 1.0La.

5.5.14.d.1 b. Prior to unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.6La for Type B and C tests and < 0.75La for the Type A test. L01 5.5.14.d.1 c. The personnel and emergency air lock leakage rates, when combined with the A09 cumulative Type B and C leakage, shall be < 0.6La. For each air lock door seal, the 5.5.14.d.2 leakage rate shall be < 0.005La when tested to 10 psig.

Add Proposed ITS 5.5.4 M03 Add Proposed ITS 5.5.5 M04 Add Proposed ITS 5.5.10 M07 Add Proposed ITS 5.5.11 M08 Add proposed ITS 5.5.13 M09 Add proposed ITS 5.5.15 M10 Add proposed ITS 5.5.16 M11 Amendment No. 190 TS 6.20-1 03/08/2007 Page 17 of 17 Attachment 1, Volume 16, Rev. 2, Page 70 of 171

Attachment 1, Volume 16, Rev. 2, Page 71 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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

A02 CTS 6.18.a states that the ODCM shall be approved by the Commission prior to implementation. ITS 5.5.1 does not retain this requirement. This changes the CTS by not requiring the Commission's approval prior to implementation.

The purpose of CTS 6.18.a is for the Commission to approve of the ODCM prior to its use for the first time. Since the ODCM has already been issued prior to ITS and is currently at Revision 11, there is no need to retain this requirement.

Additionally, ITS 5.5.1.c.3 (CTS 6.18.b.3) requires that the license changes to the ODCM be submitted to the NRC as part of or concurrent with the Radioactive Effluent Release Report. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 6.12 specifies the requirements for System Integrity, however there is no specific statement as to whether or not the provisions of CTS 4.0.b are applicable. ITS 5.5.2 states that the provisions of SR 3.0.2 are applicable to the Primary Coolant Sources Outside Containment Program Surveillance Frequencies. This changes the CTS by adding the allowance of ITS SR 3.0.2 to the Primary Coolant Sources Outside Containment Program.

This statement is needed to maintain the allowance for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.2 is not normally applied to Frequencies identified in the Administrative Controls Chapter of the ITS. Since this change is a clarification required to maintain provisions that would be allowed in the LCO sections of the Technical Specifications, it is considered administrative in nature. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 4.2.a.2 specifies the testing requirements for the in-service testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves. As part of the Surveillance Requirement however, there is no statement whether the provisions of CTS 4.0.b and 4.0.c are applicable to CTS 4.2.a.2. Additionally, CTS 4.2.a.2 states, in part, that testing shall be performed in accordance with the ASME Code of Operation and Maintenance of Nuclear Power Plants, but no table providing a description of OM Code Frequency to Technical Specification Frequency is provided. ITS 5.5.6 adds specific reference to ITS SR 3.0.2 and ITS SR 3.0.3. Additionally, ITS 5.5.6 adds specific references for the Frequencies of the in-service testing activities. This changes the CTS by adding specific references and specific testing frequencies.

Kewaunee Power Station Page 1 of 10 Attachment 1, Volume 16, Rev. 2, Page 71 of 171

Attachment 1, Volume 16, Rev. 2, Page 72 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS The purpose of CTS 4.2.a.2 is to specify the in-service testing requirements for ASME Code Class 1, Class 2, and Class 3 pumps and valves. Since CTS 4.2.a.2 is a Surveillance Requirement, the requirements of SR 4.0.b and 4.0.c (ITS SR 3.0.2 and SR 3.0.3) apply even though there is not a direct statement in the CTS 4.2.a.2 section. In the ITS, since ITS 5.5.6 is in Chapter 5.0, the requirements of SR 3.0.2 and SR 3.0.3 do not apply unless they are specifically stated. Therefore, the statements have been included in ITS 5.5.6. Additionally, ITS 5.5.6 gives the specific testing frequencies that are given in the ASME Code of Operation and Maintenance of Nuclear Power Plants rather than just referencing them as the CTS does. This change is designated as administrative because it does not result in a technical change to the CTS.

A05 The Surveillances associated with the ventilation filter testing for the Shield Building Ventilation System (SBVS), the Auxiliary Building Special Ventilation (ASV) System, and the Control Room Post Accident Recirculation (CRPAR)

System have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.9). As such, a general program statement has been added as ITS 5.5.9. Also, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extensions do apply (as allowed in the CTS). This changes the CTS by moving the ventilation filter testing Surveillances associated with the SBVS, ASV System, and CRPAR System to a program in ITS 5.5 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program.

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

A06 CTS 4.4.c.1.a, CTS 4.4.c.2, and CTS 4.4.c.3 provide filter testing requirements for the SBVS. CTS 4.4.d, in part, requires the periodic test of CTS 4.4.c.1.a, 4.4.c.2.a, 4.4.c.2.b, 4.4.c.2.c, and 4.4.c.3 to be performed. CTS 4.4.c.1.a, 4.4.c.2.a, 4.4.c.2.b, 4.4.c.2.c, and 4.4.c.3 provide filter testing requirements for the ASV System. CTS 4.17.a.1, 4.17.b.1, 4.17.b.2, and 4.17.b.3 provide filter testing requirements for the CRPAR System. ITS 5.5.9.d requires demonstration that the pressure drop across the combined HEPA filters and charcoal adsorber banks for the SBVS, the ASV System, and the Control Room Post-Accident Recirculation System once every 18 months. This changes the CTS by deleting the "once per operating cycle" terminology.

This change is acceptable since the terms "operating cycle" and "18 months" are synonymous. The Surveillance Frequency remains essentially unchanged since the KPS refueling outage occurs every 18 months. The requirements in both the CTS and ITS remain unchanged in that a test is required to ensure the system will perform its intended function. This change is designated as administrative because it does not result in technical changes to the CTS.

Kewaunee Power Station Page 2 of 10 Attachment 1, Volume 16, Rev. 2, Page 72 of 171

Attachment 1, Volume 16, Rev. 2, Page 73 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS A07 CTS 4.4.c.3 contains a footnote which states:

"In WPS letter of August 25, 1976 to Mr. Al Schwencer (NRC) from Mr. E. W.

James, we relayed test results for flow distribution for tests performed in accordance with ANSI N510-1975. This standard refers to flow distribution tests performed upstream of filter assemblies. Since the test results upstream of filters were inconclusive due to high degree of turbulence, tests for flow distribution were performed downstream of filter assemblies with acceptable results (within 20%). The safety evaluation attached to Amendment 12 references our letter of August 25, 1976 and acknowledges acceptance of the test results."

ITS 5.5.9.e requires, in part, that demonstration for each of the specified safety related systems that when tested at the system design flowrate (+/- 10%), the air distribution is uniform within +/- 20%. This changes the CTS by not including the footnote.

The purpose of the footnote was to acknowledge that the test results for flow distribution performed in accordance with ANSI N510-1975 were inconclusive due to a high degree of turbulence. Therefore, the test is performed downstream of the filter assemblies. ITS 5.5.9.e continues to perform the testing downstream of the filter assemblies with results within +/- 20% of the design flow. ITS does not contain the reference to the WPS letter and Amendment 12. This change is designated as administrative because it does not result in technical changes to the CTS.

A08 CTS 6.20, Containment Leakage Rate Testing Program, requires the performance of containment leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as modified by NRC-approved exemptions, and Regulatory Guide 1.163, dated September 1995. CTS 6.20 states that the provisions of Specification 4.0.b do not apply to the test frequencies in the Containment Leakage Rate Testing Program. ITS 5.5.14 does not include this provision. This changes the CTS by deleting the statement that the provisions of Specification 4.0.b are not applicable.

This change is acceptable because no changes have been made to the existing requirements. The statement associated with CTS 4.0.b is not needed since the Frequency extension of ITS SR 3.0.2 is not applied to Frequencies identified in the Administrative Controls Section of the ITS, unless specifically identified. This change is designated as administrative because it does not result in technical changes to the CTS.

A09 The leakage rate acceptance criteria of CTS 6.20.c states, in part, that the personnel and emergency air lock leakage rates, when combined with the cumulative Type B and C leakage, shall be < 0.60 La. There is no separate leakage rate test for the airlock; only the individual door seals have a separate leakage limit. ITS 5.5.14 does not specify that the overall air lock leakage limit is part of the Types B and C limit. This changes the CTS by not specifically stating that the overall containment air locks leakage is part of the overall Types B and C limit.

Kewaunee Power Station Page 3 of 10 Attachment 1, Volume 16, Rev. 2, Page 73 of 171

Attachment 1, Volume 16, Rev. 2, Page 74 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS 10 CFR 50, Appendix J states overall air lock leakage tests are a Type B test.

Furthermore, ITS Bases 3.6.1, as well as ITS Bases 3.6.2 states that overall air lock leakage is part of the Types B and C limit. Therefore, since the overall air lock leakage is part of the combined Types B and C limit (<0.60 La), and this limit is specified in the ITS (ITS 5.5.14.d.1), there is no reason to restate this fact in the air lock section of the Program. This change is acceptable because no changes have been made to the existing requirements. The overall air lock leakage rate is considered to include the personnel and emergency air lock leakage rate in combination with the cumulative Type B and C leakage limit. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.18.b.2 states, in part, that the ODCM becomes effective after review and acceptance by the PORC. ITS 5.5.1.c.2 states, in part, that the ODCM becomes effective after review and acceptance by the plant manager. This changes the CTS by requiring the plant manager approval for the ODCM.

The purpose of CTS 6.18.b.2 is to ensure that the ODCM has been properly reviewed by the PORC. ITS 5.5.1 still requires that the PORC review is performed (see DOC LA01 for the exclusion of the PORC from the ITS 5.5.1), but also includes an additional acceptance that the plant manager must review and approve the ODCM. This change is designated as more restrictive since a higher level of approval is required in the ITS than was required in the CTS.

M02 CTS 6.12 specifies the requirements for System Integrity, however it does not list the systems that must be monitored. ITS 5.5.2 specifically lists the systems that must be monitored by the Primary Coolant Sources Outside Containment Program. This changes the CTS by adding the specific systems that are affected by the Primary Coolant Sources Outside Containment Program.

The purpose of the Primary Coolant Sources Outside Containment Program is to minimize leakage from those portions of systems that could contain highly radioactive fluids during serious transient or accidents to levels as low as reasonably practicable. The systems added to the Specification include the Safety Injection System, Chemical and Volume Control System, Containment Spray System, Reactor Building Ventilation System, Residual Heat Removal System, Miscellaneous Sumps and Drains System, and Primary Sampling System. This change is acceptable because these systems are currently monitored to satisfy the CTS requirement and is a complete list of those systems that could contain highly radioactive fluids during a serious transient or accident.

This change is designated as more restrictive because it adds an explicit list of systems to the Technical Specifications.

M03 The CTS does not include program requirements for a Component Cyclic or Transient Limit Program. The ITS includes a program for this activity. This changes the CTS by adding the Component Cyclic or Transient Limit Program.

Kewaunee Power Station Page 4 of 10 Attachment 1, Volume 16, Rev. 2, Page 74 of 171

Attachment 1, Volume 16, Rev. 2, Page 75 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS The Component Cyclic or Transient Limit Program is included to ensure controls are in place to track the USAR Section 4.1.5 and Table 4.1-8, cyclic and transient occurrences which ensures that components are maintained within the design limits. This change is acceptable because it supports implementation of the requirements of the USAR. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

M04 The CTS does not include program requirements for the Reactor Coolant Pump Flywheel Inspection Program. The ITS includes a program for this activity. This changes the CTS by adding the Reactor Coolant Pump Flywheel Inspection Program.

The Reactor Coolant Pump Flywheel Inspection Program provides a requirement to perform an inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. This change is acceptable because it supports implementation of the requirements of the ITS. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

M05 CTS 6.15, Secondary Water Chemistry Program, does not include the low pressure turbine disc stress corrosion cracking as part of the scope of the program. ITS 5.5.8 includes the low pressure turbine disc stress corrosion cracking as part of the monitoring program. This changes the CTS by adding the low pressure turbine disc stress corrosion cracking to the scope of the Secondary Water Chemistry Program.

The purpose of the Secondary Water Chemistry Program is to provide controls for monitoring secondary water chemistry to inhibit both steam generator tube degradation and low pressure turbine disc stress corrosion cracking. The contamination of steam generator secondary coolant is the fundamental cause of steam generator degradation, impairment of steam generator tube integrity, and low pressure turbine disc stress corrosion. In order to control deposits and corrosion within the secondary coolant, guidelines have been established to control the water chemistry of the secondary side. The addition of the low pressure turbine disc stress corrosion cracking to the Secondary Water Chemistry Program is acceptable because the careful control of the quality of the secondary water chemistry serves to inhibit the potential accumulation of corrosive impurities that could eventually contribute to component degradation and eventual failure. This change is more restrictive because a new requirement is being added to the Technical Specifications.

M06 CTS 6.15, Secondary Water Chemistry Program, does not contain specific control mechanisms for monitoring the secondary water chemistry. It only provides a general description that secondary water chemistry program shall act as a guide for the chemistry group in their routine as well as non-routine activities. ITS 5.5.8 includes specific control mechanisms that shall be included in the program. This changes the CTS by adding additional details regarding what is to be included in the Secondary Water Chemistry Program.

Kewaunee Power Station Page 5 of 10 Attachment 1, Volume 16, Rev. 2, Page 75 of 171

Attachment 1, Volume 16, Rev. 2, Page 76 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS The purpose of the additional steps in ITS 5.5.8 (i.e., 5.5.8.a through 5.5.8.f) is to provide specific details of what should be included in a program to monitor the secondary water chemistry. The additional details include identification of a sampling schedule for critical variables; identification of the procedures used to measure values of critical variables; identification of process sampling points to monitor for evidence of condenser in-leakage; procedures for recording and management of data; procedures defining corrective actions for off control point chemistry conditions; and a procedure identifying the authority responsible for interpretation of the data and the sequence and timing of administrative events required to initiate corrective actions. This change is acceptable because additional details are provided to assist in developing a program that will ensure the intent of the program is met and will contribute to an effective means of monitoring the chemistry of the secondary water. This change is designated more restrictive because it adds new requirements to the Technical Specifications.

M07 The CTS does not have a program for Explosive Gas and Storage Tank Radioactivity Monitoring. ITS 5.5.10 requires a program to provide controls for potentially explosive gas mixtures contained in the Gaseous Radioactive Waste Disposal System, the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. This changes the CTS by incorporating the requirements of ITS 5.5.10.

The purpose of the program is to control potentially explosive gas mixtures contained in the Gaseous Radioactive Waste Disposal System, the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. Additionally, this change is consistent with the requirements in ODCM sections 3/4.3 and 3/4.4. This change is designated as more restrictive because it imposes new programmatic requirements in the Technical Specifications.

M08 The CTS does not have a program for Diesel Fuel Oil Testing. ITS 5.5.11 requires a program to implement testing of both new fuel oil and stored fuel oil.

This changes the CTS by incorporating the requirements of ITS 5.5.11.

The purpose of the program is to provide sampling and testing requirements, along with acceptance criteria for testing of new and stored fuel oil. This change is designated as more restrictive because it imposes new programmatic requirements in the Technical Specifications.

M09 The CTS does not include program requirements for the Safety Function Determination Program. The ITS includes a program for the Safety Function Determination Program. This change the CTS by adding the Safety Function Determination Program (SFDP).

The Safety Function Determination Program is included to support implementation of the support system OPERABILITY characteristics of the Technical Specifications. The specific wording associated with this program is Kewaunee Power Station Page 6 of 10 Attachment 1, Volume 16, Rev. 2, Page 76 of 171

Attachment 1, Volume 16, Rev. 2, Page 77 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS found in ITS 5.5.13. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications.

M10 The CTS does not include a requirement for Battery Monitoring and Maintenance Program. The ITS includes a requirement for this program. This changes the CTS by adding the ITS 5.5.15, "Battery Monitoring and Maintenance Program."

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

M11 The CTS does not have a program for Setpoint Control. ISTS 5.5.18 (ITS 5.5.16) requires a program to satisfy the regulatory requirement of 10 CFR 50.36(c)(1)(ii)(A) that Technical Specifications will include items in the category of limiting safety system settings (LSSS), which are settings for automatic protective devices related to those variables having significant safety functions. This changes the CTS by incorporating the requirements of ISTS 5.5.18 (ITS 5.5.16).

The purpose of the program is to establish, implement, and maintain instrument setpoint controls for automatic protective devices related to those variables having significant safety functions. This change is designated as more restrictive because it imposes new programmatic requirements in the Technical Specifications.

M12 CTS 6.20 states that the peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal test pressure. The containment internal test pressure is defined as Pa in the CTS.

ITS 5.5.14.b contains a specific value for the calculated peak containment internal pressure for the design basis loss of coolant accident and the containment design pressure. The calculated peak containment internal pressure for the design basis loss of coolant accident is defined as Pa in the ITS.

This changes the CTS by adding a specific value for the calculated peak containment internal pressure for the design basis loss of coolant accident and a value for the containment design pressure.

The peak calculated containment internal pressure for the design basis loss of coolant accident is derived from the maximum containment pressure which is given as 44.6 psig at 19.9 seconds in USAR Table 14.3.5-8. The same maximum containment pressure was also reviewed and approved by the NRC as documented in the NRC Safety Evaluation for License Amendment 172 (the KPS Stretch Power Uprate), section 3.8.2.1.2.2, dated February 27, 2004 (ADAMS Accession No. ML040430633). The containment design pressure of 46 psig is also documented in the USAR and was reviewed and approved in the original USAR approval. This change is designated as more restrictive because it imposes new values that were not included in the CTS.

Kewaunee Power Station Page 7 of 10 Attachment 1, Volume 16, Rev. 2, Page 77 of 171

Attachment 1, Volume 16, Rev. 2, Page 78 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS M13 CTS 3.6.c.3.A, 3.6.c.3.C, 3.8.a.9.b.1, 3.8.a.9.b.3, 3.12.c.1, 3.12.c.3, 4.4.c.1.a, 4.4.c.3, 4.4.d.1, 4.12.a.1, and 4.17.a.1 require certain Ventilation System filter tests to be performed at the system design flow rate. However, no specific value of the various Ventilation System design flow rates is provided. ITS 5.5.9.a, b, d, and e require performance of similar Ventilation System filter tests, and include the specific value of the design flow rate for each required Ventilation System.

This changes the CTS by adding the specific design flow rate values into the filter test requirements of the ITS.

The various Ventilation System filter tests are performed at the system design flow rate. However, the specific values are currently controlled in plant procedures and the USAR. This change will add the values into the Technical Specifications, and thus any change will require NRC approval in lieu of changing it by the 10 CFR 50.59 process. Therefore, this change is considered acceptable. This change is designated as more restrictive because new Ventilation System design flow rate values are being included in the ITS that are not required in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.18.b.1 requires changes to the ODCM to be documented and records of reviews performed to be retained as required by the quality assurance program. CTS 6.18.b.2 requires changes to the ODCM to be effective after review and acceptance by the PORC. ITS 5.5.1.c.1 requires changes to the ODCM to be documented and records of reviews performed to be retained. ITS 5.5.1.c.2 requires changes to the ODCM to become effective after the approval of the plant manager. This changes the CTS by moving the record retention requirements reference and the PORC review and approval requirements to the Nuclear Facility Quality Assurance Program Description (NFQAPD). DOC M01 describes the addition of the plant manager approval.

The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety.

ITS 5.5.1 still retains the requirement for changes to the ODCM to be documented and retained. Also, this change is acceptable because these types of procedural details will be adequately controlled in the NFQAPD. Any changes to the NFQAPD are made under 10 CFR 50.54(a), which ensure changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specifications requirements are being removed from the Technical Specifications.

Kewaunee Power Station Page 8 of 10 Attachment 1, Volume 16, Rev. 2, Page 78 of 171

Attachment 1, Volume 16, Rev. 2, Page 79 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS LA02 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, USAR, ODCM, NFQAPD, CLRT Program, IST Program, ISI Program, or Setpoint Control Program) CTS 4.2.a and 4.2.a.1 provide requirements for the In-Service Inspection Program. The ITS does not include In-Service Inspection Program requirements. This changes the CTS by moving these requirements from the Technical Specifications to the In-Service Inspection (ISI) Program.

The removal of these requirements is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The Technical Specifications still retain requirements for the affected components to be OPERABLE. Also, this change is acceptable because these requirements will be adequately controlled by the ISI Program, which is required by 10 CFR 50.55a.

Compliance with 10 CFR 50.55a is required by the Kewaunee Operating License. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 6.20.b states, in part, that the leakage rate acceptance criteria prior to unit startup for the Type A test is

< 0.75 La. ITS 5.5.14.d.1 states, in part, that the leakage rate acceptance criteria prior to unit startup for the Type A test is 0.75 La. This changes the CTS by allowing the leakage rate to be exactly equal to 0.75 La in lieu of being < 0.75 La.

The purpose of ITS 5.5.14.d.1 is to ensure that prior to a unit startup the overall containment leakage rate with a certain amount of margin, does not exceed the value assumed in the accident analysis. This change is acceptable because the acceptance criteria limit in ITS 5.5.14.d.1 (the 1.0 La limit) continues to ensure the containment leakage is within the value assumed in the accident analysis. The 10CFR50 Appendix J Option B states that for Type A tests the leakage rate must not exceed the allowable leakage rate (La) with margin, as specified in the Technical Specifications. The ITS now provides a margin that includes allowing the limit to be exactly 0.75 La. This change is designated as less restrictive because the acceptance criteria applicable prior to a unit startup for the Type A leakage tests is now allowed to be exactly equal to 0.75 La in lieu of being

< 0.75 La.

L02 (Category 1 - Relaxation of LCO Requirements) CTS 4.4.c.1.a provides the pressure drop test acceptance criteria for the Shield Building Ventilation System (SBVS) and the Auxiliary Building Special Ventilation (ASV) System. The acceptance criteria are a pressure drop across the combined HEPA filters and charcoal adsorber banks < 10 inches of water and a pressure drop across any HEPA filter bank < 4 inches of water. CTS 4.17.a.1 provides the pressure drop test acceptance criteria for the Control Room Post Accident Recirculation (CRPAR) System. The acceptance criteria are a pressure drop across the combined HEPA filters and charcoal adsorber banks < 6 inches of water and a pressure drop across any HEPA filter bank < 4 inches of water. ITS 5.5.9.d provides the pressure drop test acceptance criteria for all three systems, and Kewaunee Power Station Page 9 of 10 Attachment 1, Volume 16, Rev. 2, Page 79 of 171

Attachment 1, Volume 16, Rev. 2, Page 80 of 171 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS requires the pressure drop across the combined prefilters, HEPA filters, and charcoal adsorber banks to be < 6.3 inches of water for the SBVS and the ASV System, and < 2.4 inches of water for the CRPAR System. This changes the CTS by reducing the acceptance criteria for the combined dP, including the prefilters as part of the combined dp test, and deleting the HEPA filter only dP criteria.

Performance testing demonstrating that dP is within prescribed limits assures that fans operate within +/- 10% of design flow, and thus ensures that filters perform adequately to satisfy the accident analyses. That is, if the performances are as specified, the calculated doses would be less than the guidelines stated in 10CFR Part 100 for the accidents analyzed.

During the development of Design Basis Documents for these ventilation systems it was identified that the analytical basis for the CTS dP acceptance criteria could not be verified. KPS has recently performed calculations, based on a newly developed analytical basis. These calculations have determined that the current dP acceptance criteria are non-conservative. KPS has implemented, in procedures, these more restrictive acceptance criteria. This change to the CTS replaces the non-conservative dp acceptance criteria with the current acceptance criteria from the recent calculation. This change is acceptable because it will incorporate newly calculated, and more conservative, dP acceptance criteria.

Combined dP testing across the prefilter, HEPA filter and charcoal adsorber banks, per the ISTS format, in lieu of CTS combined HEPA filter and charcoal adsorber bank and HEPA-only dP testing, will still adequately demonstrate that the filters are not clogged and capable of performing their design function.

The change related to the reduction of the combined dP limit is more restrictive, since it is reducing the limit and adding another component (prefilter) to the testing. However, the change related to removing the HEPA filter only dP limit is less restrictive. Therefore, this change is designated as less restrictive because the acceptance criteria applicable to the HEPA filter only dP test is being deleted.

Kewaunee Power Station Page 10 of 10 Attachment 1, Volume 16, Rev. 2, Page 80 of 171

Attachment 1, Volume 16, Rev. 2, Page 81 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 81 of 171

Attachment 1, Volume 16, Rev. 2, Page 82 of 171 CTS Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM) 1.0.o.2 a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program, and 3 1.0.o.2

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification [5.6.1] and Specification [5.6.2]. 1 6.18.b Licensee initiated changes to the ODCM: 2 c.

6.18.b.1 a. Shall be documented and records of reviews performed shall be retained. 2 1 This documentation shall contain:

6.18.b.1.A

1. Sufficient information to support the change(s) together with the 2 a) appropriate analyses or evaluations justifying the change(s) and 3 6.18.b.1.B 2. A determination that the change(s) maintain the levels of radioactive 2 b) effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations, 3 6.18.b.2

b. Shall become effective after the approval of the plant manager, and 2 3 2 ;

6.18.b.3 c. Shall be submitted to the NRC in the form of a complete, legible copy of the 2 3 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.

WOG STS 5.5-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 82 of 171

Attachment 1, Volume 16, Rev. 2, Page 83 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 Primary Coolant Sources Outside Containment The System Integrity (SIP) 4 6.12 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 System

[Recirculation Spray, Safety Injection, Chemical and Volume Control, gas 5 stripper, and Hydrogen Recombiner]. The program shall include the following:

INSERT 1 6.12.a a. Preventive maintenance and periodic visual inspection requirements and 3 6.12.b b. Integrated leak test requirements for each system at least once per

[18] months. 1 DOC A03 The provisions of SR 3.0.2 are applicable.

5.5.3 [ Post Accident Sampling

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

This program may be eliminated based on the implementation of WCAP-14986, Rev. 1, "Post Accident Sampling System Requirements: A Technical Basis," and the associated NRC Safety Evaluation dated June 14, 2000.

This program provides controls that ensure the capability to obtain and analyze 6 reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

a. Training of personnel,

b. Procedures for sampling and analysis, and

c. Provisions for maintenance of sampling and analysis equipment. ]

6 6.16.b.1 5.5.4 Radioactive Effluent Controls Program 3

This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

WOG STS 5.5-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 83 of 171

Attachment 1, Volume 16, Rev. 2, Page 84 of 171 5.5.2 o5 INSERT 1 Containment Spray System, Miscellaneous Sumps and Drains System, Reactor Building Ventilation System, Residual Heat Removal System, and Primary Sampling System Insert Page 5.5-2 Attachment 1, Volume 16, Rev. 2, Page 84 of 171

Attachment 1, Volume 16, Rev. 2, Page 85 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued) 6 6.16.b.1.A 3 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, ; 3 6.16.b.1.B b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, ; 3 6.16.b.1.C c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM, ; 3 6.16.b.1.D 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, 3 6.16.b.1.E e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days, ; 3 6.16.b.1.F f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I, ; 3 6.16.b.1.G g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin and

3

2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ, ; 3 6.16.b.1.H 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, ; 3 WOG STS 5.5-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 85 of 171

Attachment 1, Volume 16, Rev. 2, Page 86 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued) 6 3

6.16.b.1.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 > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, and 3

6.16.b.1.J j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

6.16.b.1 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency. 7 ies 6 7 DOC 5.5.5 Component Cyclic or Transient Limit Program M03 U 4 4

This program provides controls to track the FSAR, Section [ ], cyclic and 18 1 transient occurrences to ensure that components are maintained within the design limits.

5.5.6 [ Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in 6 accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC.

The provisions of SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies. ]

DOC M04 5.5.7 Reactor Coolant Pump Flywheel Inspection Program 6

5 This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.

WOG STS 5.5-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 86 of 171

Attachment 1, Volume 16, Rev. 2, Page 87 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Reactor Coolant Pump Flywheel Inspection Program (continued) 6 5

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

The inspection interval and scope for RCP flywheels stated above can be applied 8

to plants that satisfy the requirements in WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination."

4.2.a.2 5.5.8 Inservice Testing Program 6 6

This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following: 9 pumps and valves 4.2.a.2 a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

DOC A04 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 9 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 9

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 DOC A04 to

b. The provisions of SR 3.0.2 are applicable to the above required as 2 years or less TSTF-Frequencies and other normal and accelerated Frequencies specified in the 497-A Inservice Testing Program for performing inservice testing activities, 9 DOC A04 c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and 9 DOC A04 d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

WOG STS 5.5-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 87 of 171

Attachment 1, Volume 16, Rev. 2, Page 88 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.22 5.5.9 Steam Generator (SG) Program 6 7

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

6.22.a a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice 12 inspection results or by other means, prior to the plugging [or repair] of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, [or repaired] to 12 confirm that the performance criteria are being met. or 6.22.b b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

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 150 gallons per day exceed [1 gpm] per SG [, except for specific types of degradation at 1 12 specific locations as described in paragraph c of the Steam Generator Program.

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

6.22.c c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding [40%] of the nominal 1 tube wall thickness shall be plugged [or repaired]. 12

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

Alternate tube repair criteria currently permitted by plant technical specifications are 12 listed here. The description of these alternate tube repair criteria should be equivalent to the descriptions in current technical specifications and should also include any allowed accident induced leakage rates for specific types of degradation at specific locations associated with tube repair criteria.

[The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria: 12

1. . . .]

6.22.d d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the 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.

WOG STS 5.5-7 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 89 of 171

Attachment 1, Volume 16, Rev. 2, Page 90 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.22 5.5.9 Steam Generator (SG) Program (continued) 6 7

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

Plants are to include the appropriate Frequency (e.g., select the appropriate Item 2.) for their SG design. The first Item 2 is applicable to SGs with Alloy 600 13 mill annealed tubing. The second Item 2 is applicable to SGs with Alloy 600 thermally treated tubing. The third Item 2 is applicable to SGs with Alloy 690 thermally treated tubing.

6.22.d 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 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.]

13

[2. Inspect 100% of the tubes at sequential periods of 120, 90, 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 48 effective full power months or two refueling outages (whichever is less) without being inspected.]

13

[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 13 (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.

6.22.e e. Provisions for monitoring operational primary to secondary LEAKAGE.

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Attachment 1, Volume 16, Rev. 2, Page 91 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.22 5.5.9 Steam Generator (SG) Program (continued) 6 7

[f. Provisions for SG tube repair methods. Steam generator tube repair 12 methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

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

Tube repair methods currently permitted by plant technical specifications are to be listed here. The description of these tube repair methods should be 12 equivalent to the descriptions in current technical specifications. If there are no approved tube repair methods, this section should not be used.

1. . . .] 12 5.5.10 Secondary Water Chemistry Program 6 8

6.15 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 variables and control points for these variables, ; 3

b. Identification of the procedures used to measure the values of the critical variables, ; 3

c. Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage, ; 3

d. Procedures for the recording and management of data, ; 3

e. Procedures defining corrective actions for all off control point chemistry conditions, and 3

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

WOG STS 5.5-9 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 91 of 171

Attachment 1, Volume 16, Rev. 2, Page 92 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) 6 safety related Regulatory Positions C.5.c and C.5.d of 9

DOC A05, A program shall be established to implement the following required testing of 14 4.4.c.1.a, Engineered Safety Feature (ESF) filter ventilation systems at the frequencies 11 4.12.a.1, 4.17.a.1 specified in [Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, 1 Revision 2, ASME N510-1989, and AG-1]. ASTM D3803-1989, 14 INSERT 2 ANSI 1975, Regulatory Position C.5.c of listed below 3.6.c.3.A, 3.6.c.3.C, safety a. Demonstrate for each of the ESF systems that an inplace test of the high 11 3.8.a.9.b.1, related 3.8.a.9.b.3, < 1.0 efficiency particulate air (HEPA) filters shows a penetration and system 3.12.c.1, 3.12.c.3 bypass < [0.05]% when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below 1

[+/- 10%]. ANSI 1975 ESF Ventilation System Flowrate INSERT 3 14 safety [ ] [ ]

related listed below 11 3.6.c.3.A, 3.6.c.3.C, b. Demonstrate for each of the ESF systems that an inplace test of the < 1.0 3.8.a.9.b.1, 3.8.a.9.b.3, Regulatory charcoal adsorber shows a penetration and system bypass < [0.05]% when 3.12.c.1, 3.12.c.3 Position C.5.d of tested in accordance with [Regulatory Guide 1.52, Revision 2, and 3 ASME N510-1989] at the system flowrate specified below [+/- 10%].

ANSI 1975 ESF Ventilation System Flowrate 14

[ ] [ ] INSERT 4 safety listed below related 11 3.6.c.3.B, c. Demonstrate for each of the ESF systems that a laboratory test of a sample 3.8.a.9.b.2, 3.12.c.2 of the charcoal adsorber, when obtained as described in [Regulatory and AG-1 1 Guide 1.52, Revision 2], shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity specified below. of 95% 14 ESF Ventilation System Penetration RH Face Velocity (fps)

[ ] [See Reviewer's [See [See Reviewer's INSERT 5 14 Note] Reviewer's Note]

Note]

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

The use of any standard other than ASTM D3803-1989 to test the charcoal sample may result in an overestimation of the capability of the charcoal to adsorb 15 radioiodine. As a result, the ability of the charcoal filters to perform in a manner consistent with the licensing basis for the facility is indeterminate.

WOG STS 5.5-10 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 92 of 171

Attachment 1, Volume 16, Rev. 2, Page 93 of 171 CTS 5.5 14 INSERT 2 4.4.c.2.a, The test described in Specification 5.5.9.a shall be performed once per 18 months and 4.4.d.1, after each complete or partial replacement of the high efficiency particulate air (HEPA) 4.17.b.1 filter bank and any maintenance on the system that could affect the HEPA bank bypass leakage.

4.4.c.2.c, The test described in Specification 5.5.9.b shall be performed after each complete or 4.4.d.1, 4.17.b.3 partial replacement of a charcoal adsorber bank or maintenance on the system that could affect the charcoal adsorber bank bypass leakage.

4.4.c.2.b, The test described in Specification 5.5.9.c shall be performed once per 18 months for 4.4.d.1, filters in a standby status or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of filter operation, and following painting, 4.17.b.2 fire, or chemical release in any ventilation zone communicating with the system.

4.4.c.1.a, 4.4.d.1, The test described in Specification 5.5.9.d shall be performed once per 18 months.

4.17.a.1 4.4.c.3, The test described in Specification 5.5.9.e shall be performed after any maintenance or 4.4.d.1 testing that could affect the air distribution within the systems.

14 INSERT 3 Safety Related System Flow Rate (cfm)

Shield Building Ventilation System (SBVS) 5700 Auxiliary Building Special Ventilation (ASV) System 9000 Control Room Post Accident Recirculation (CRPAR) 2500 System 14 INSERT 4 Safety Related System Flow Rate (cfm)

SBVS 5700 ASV System 9000 CRPAR System 2500 Insert Page 5.5-10a Attachment 1, Volume 16, Rev. 2, Page 93 of 171

Attachment 1, Volume 16, Rev. 2, Page 94 of 171 5.5 14 INSERT 5 Safety Related System Penetration SBVS < 2.5%

ASV System < 2.5%

CRPAR System < 5%

Insert Page 5.5-10b Attachment 1, Volume 16, Rev. 2, Page 94 of 171

Attachment 1, Volume 16, Rev. 2, Page 95 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (continued) 6 9

ASTM D 3803-1989 is a more stringent testing standard because it does not differentiate between used and new charcoal, it has a longer equilibration period performed at a temperature of 30°C (86°F) and a relative humidity (RH) of 95%

(or 70% RH with humidity control), and it has more stringent tolerances that improve repeatability of the test.

Allowable Penetration = [(100% - Methyl Iodide Efficiency

for Charcoal Credited in Licensee's Accident Analysis) / Safety Factor]

When ASTM D3803-1989 is used with 30°C (86°F) and 95% RH (or 70% RH with humidity control) is used, the staff will accept the following:

Safety factor 2 for systems with or without humidity control. 15 Humidity control can be provided by heaters or an NRC-approved analysis that demonstrates that the air entering the charcoal will be maintained less than or equal to 70 percent RH under worst-case design-basis conditions.

If the system has a face velocity greater than 110 percent of 0.203 m/s (40 ft/min), the face velocity should be specified.

This value should be the efficiency that was incorporated in the licensee's accident analysis which was reviewed and approved by the staff in a safety evaluation.

safety listed below related 11 4.4.c.1.a, d. Demonstrate for each of the ESF systems that the pressure drop across the 4.4.d.1, combined HEPA filters, the prefilters, and the charcoal adsorbers is less 4.12.a.1, 4.17.a.1 than the value specified below when tested in accordance with [Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate 1 specified below [+/- 10%]. ANSI 1975 ESF Ventilation System Delta P Flowrate

[ ] [ ] [ ]

INSERT 6 14

[ e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below [+/- 10%] when tested in accordance with

[ASME N510-1989].

INSERT 7 14 ESF Ventilation System Wattage ]

[ ] [ ]

DOC A05 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

WOG STS 5.5-11 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 95 of 171

Attachment 1, Volume 16, Rev. 2, Page 96 of 171 5.5 14 INSERT 6 Safety Related System Combined Flow Rate Delta P (cfm)

(in. wc)

SBVS < 6.3 5700 ASV System < 6.3 9000 CRPAR System < 2.4 2500 14 INSERT 7 Demonstrate for each of the safety related systems listed below that when tested at the system flowrate specified below (+/- 10%) the air distribution is uniform within +/- 20%.

Safety Related System Flow Rate (cfm)

SBVS 5700 ASV System 9000 Insert Page 5.5-11 Attachment 1, Volume 16, Rev. 2, Page 96 of 171

Attachment 1, Volume 16, Rev. 2, Page 97 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program 6 M07 10 This program provides controls for potentially explosive gas mixtures contained Gaseous Radioactive in the [Waste Gas Holdup System], [the quantity of radioactivity contained in gas Waste Disposal storage tanks or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous radioactivity quantities shall be determined following the methodology in [Branch 1 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"].

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the [Waste Gas 1 Gaseous Radioactive Waste Disposal Holdup System] and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion), 3

b. A surveillance program to ensure that the quantity of radioactivity contained in [each gas storage tank and fed into the offgas treatment system] is less than the amount that would result in a whole body exposure of 0.5 rem to 1 any individual in an unrestricted area, in the event of [an uncontrolled release of the tanks' contents], and 7 3

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 Radwaste 1 Waste Disposal Treatment System] is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. 7 WOG STS 5.5-12 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 97 of 171

Attachment 1, Volume 16, Rev. 2, Page 98 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC M08 5.5.13 Diesel Fuel Oil Testing Program 6 11 A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for 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, 3

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, 3 and ;

3. A clear and bright appearance with proper color or a water and sediment content within limits. 3

b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a.,

above, are within limits for ASTM 2D fuel oil, and 3

c. Total particulate concentration of the fuel oil is 10 mg/l when tested every 31 days. 17 92 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

6.21 5.5.14 Technical Specifications (TS) Bases Control Program 6 12 6.21 This program provides a means for processing changes to the Bases of these Technical Specifications.

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

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

6.21.b.1 1. A change in the TS incorporated in the license or 3 U ;

6.21.b.2 2. A change to the updated FSAR or Bases that requires NRC approval 18 pursuant to 10 CFR 50.59.

WOG STS 5.5-13 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 98 of 171

Attachment 1, Volume 16, Rev. 2, Page 99 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.21 5.5.14 Technical Specifications (TS) Bases Control Program (continued) 6 12 6.21.d c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR. 18 U 12. 7 6.21.c, d. Proposed changes that meet the criteria of Specification 5.5.14b above shall 6 6.21.e be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

DOC 5.5.15 Safety Function Determination Program (SFDP) 6 M09 13 This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the 2 following: ¶ a.

a. Provisions for cross train checks to ensure a loss of the capability to perform 2 1 the safety function assumed in the accident analysis does not go undetected, 3

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of 2 2

function condition exists, 3

c. Provisions to ensure that an inoperable supported system's Completion 2 3

Time is not inappropriately extended as a result of multiple support system inoperabilities, and 3

d. Other appropriate limitations and remedial or compensatory actions. 2 4

b A loss of safety function exists when, assuming no concurrent single failure, no 2 concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: 3

a. A required system redundant to the system(s) supported by the inoperable 2 1

support system is also inoperable, or 3

b. A required system redundant to the system(s) in turn supported by the 2 2

inoperable supported system is also inoperable, or 3 WOG STS 5.5-14 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 99 of 171

Attachment 1, Volume 16, Rev. 2, Page 100 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC 5.5.15 Safety Function Determination Program (SFDP) (continued) 6 M09 13

c. A required system redundant to the support system(s) for the supported 2 3

systems (a) and (b) above is also inoperable.

described in Specifications 5.5.13.b.1 and 5.5.13.b.2

c. The SFDP identifies where a loss of safety function exists. If a loss of safety 2 function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

6.20 5.5.16 Containment Leakage Rate Testing Program 6 14

[OPTION A]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

b. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

c. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests. 19

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa.

b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

d. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

e. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

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Attachment 1, Volume 16, Rev. 2, Page 101 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.20 5.5.16 Containment Leakage Rate Testing Program (continued) 6 14 19

[OPTION B]

6.20 a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as .

modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, 20 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.

[ 3. ...]

6.20 b. The calculated peak containment internal pressure for the design basis loss 44.6 of coolant accident, Pa, is [45 psig]. The containment design pressure is 1

[50 psig].

46 0.2 6.20 c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of 14 1 containment air weight per day. 46 psig (Peak Test Pressure)

d. Leakage rate acceptance criteria are:

6.20.a, 1. Containment leakage rate acceptance criterion is 1.0 La. During the 6.20.b first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and 0.75 La for Type A tests.

door seal leakage 6.20.c 2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa. 22 seal is a b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig]. < 0.005 of WOG STS 5.5-16 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 101 of 171

Attachment 1, Volume 16, Rev. 2, Page 102 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) 6 14 6.20 e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the 23 testing Frequencies required by 10 CFR 50, Appendix J.

[OPTION A/B Combined]

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. [Type A][Type B and C] test requirements are in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. [Type B and C][Type A] test requirements are in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The 10 CFR 50, Appendix J, Option B test requirements shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, 19 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.

[ 3. ...]

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, [45 psig]. The containment design pressure is

[50 psig].

c. The maximum allowable containment leakage rate, La, at Pa, shall be [ ]% of containment air weight per day.

d. Leakage rate acceptance criteria are:

WOG STS 5.5-17 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 102 of 171

Attachment 1, Volume 16, Rev. 2, Page 103 of 171 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued) 6 14

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and [< 0.75 La for Option A Type A tests] [ 0.75 La for Option B Type A tests].

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is [0.05 La] when tested at Pa. 19 b) For each door, leakage rate is [0.01 La] when pressurized to

[ 10 psig].

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

DOC 5.5.17 15 Battery Monitoring and Maintenance Program 6 M10 This Program provides for battery restoration and maintenance, based on [the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer] including the following: 24 INSERT 8A

a. Actions to restore battery cells with float voltage < [2.13] V, and

b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

DOC 25 INSERT 8 M11 WOG STS 5.5-18 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 103 of 171

Attachment 1, Volume 16, Rev. 2, Page 104 of 171 CTS 5.5 24 INSERT 8A DOC M10 This Program provides controls for battery restoration and maintenance. The program shall be in accordance with IEEE Standard (Std) 450-2002, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," as endorsed by Regulatory Guide 1.129, Revision 2 (RG), with RG exceptions and program provisions as identified below:

a. The program allows the following RG 1.129, Revision 2 exceptions:

1. Battery temperature correction may be performed before or after conducting discharge tests.

2. RG 1.129, Regulatory Position 1, Subsection 2, "References," is not applicable to this program.

3. In lieu of RG 1.129, Regulatory Position 2, Subsection 5.2, "Inspections," the following shall be used: "Where reference is made to the pilot cell, pilot cell selection shall be based on the lowest voltage cell in the battery." after each performance of SR 3.8.6.5 4 In Regulatory Guide 1.129, Regulatory Position 3, Subsection 5.4.1, "State of Charge Indicator," the following statements in paragraph (d) may be omitted: "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage."

5. In lieu of RG 1.129, Regulatory Position 7, Subsection 7.6, "Restoration", the following may be used: "Following the test, record the float voltage of each cell of the string."

b. The program shall include the following provisions:

1. Actions to restore battery cells with float voltage < 2.13 V;

2. Actions to determine whether the float voltage of the remaining battery cells is 2.13 V when the float voltage of a battery cell has been found to be < 2.13 V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

Insert Page 5.5-18a Attachment 1, Volume 16, Rev. 2, Page 104 of 171

Attachment 1, Volume 16, Rev. 2, Page 105 of 171 5.5 CTS 1

INSERT 8 DOC M11 5.5.18 Setpoint Control Program 25 16 This program shall establish the requirements for ensuring that setpoints for automatic protective devices are initially within and remain within the assumptions of the applicable safety analysis provides a means for processing 27 changes to instrumentation setpoints and identifies setpoint methodologies to ensure instrumentation will function as required. The program shall ensure that testing of automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A) verify that instrumentation will function as required.

a. The program shall list the Functions in the following specifications to which it applies: Protection P

1. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation";

2. LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation Functions";

Offsite O

3. LCO 3.3.5, "Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation";

Vent 26

4. LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation";

Post Accident Recirculation (CRPAR) and

5. LCO 3.3.7, "Control Room Emergency Filtration System (CREFS)

Actuation Instrumentation;" ."

6. LCO 3.3.8, "Fuel Building Air Cleanup System (FBACS) Actuation Instrumentation;"

7. LCO 3.3.9, "Boron Dilution Protection System (BDPS)."

list the value of

b. The program shall require the Nominal Trip Setpoint (NTSP), Allowable Value for each (AV), As-Found Tolerance (AFT), and As-Left Tolerance (ALT) (as applicable) of the Functions described in Paragraph a. are calculated using 27 The NRC staff has not approved the NRC approved setpoint methodology, as listed below. In addition, the processing changes to Kewaunee Power Station instrumentation setpoints program shall list the value of the NTSP, AV, AFT, or ALT (as applicable) for under 10 CFR 50.59 using an approved setpoint methodology as described in each Function described in paragraph a. and shall identify the setpoint Option B of TSTF-493. NRC approval methodology used to calculate these values.

using 10 CFR 50.90 is required to change the value of the NTSP, AV, AFT, and ALT (as applicable) for each Function described in Paragraph a.

Insert Page 5.5-18b Attachment 1, Volume 16, Rev. 2, Page 105 of 171

Attachment 1, Volume 16, Rev. 2, Page 106 of 171 5.5 CTS 1 INSERT 8 (continued)

DOC M11 --------------------------------- Reviewer's Note ---------------------------------------

List the NRC safety evaluation report by letter, date, and ADAMS accession 15 number (if available) that approved the setpoint methodologies.

27

1. [Insert reference to NRC safety evaluation that approved the setpoint methodology.]

c. The program shall establish methods to ensure that Functions described in the list of values established Paragraph a. will function as required by verifying the as-left and as-found 27 by Paragraph b. settings are consistent with those established by the setpoint methodology. If the as-found value of the instrument channel trip setting is less conservative than the specified AV, then the SR is not met and the instrument channel shall be immediately declared inoperable.

d. ----------------------------------- REVIEWERS NOTE --------------------------------------

A license amendment request to implement a Setpoint Control Program must list the instrument functions to which the program requirements of paragraph

d. will be applied. Paragraph d shall apply to all Functions in the Reactor Trip System and Engineered Safety Feature Actuation System specifications unless one or more of the following exclusions apply:

1. Manual actuation circuits, automatic actuation logic circuits or to instrument functions that derive input from contacts which have no associated sensor or adjustable device, e.g., limit switches, breaker position switches, manual actuation switches, float switches, proximity detectors, etc. are excluded. In addition, those permissives and interlocks that derive input from a sensor or adjustable device that is tested as part of another TS function are excluded.

2. Settings associated with safety relief valves are excluded. The 15 performance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program.

3. Functions and Surveillance Requirements which test only digital components are excluded. There is no expected change in result between SR performances for these components. Where separate as-left and as-found tolerance is established for digital component SRs, the requirements would apply.

Insert Page 5.5-18c Attachment 1, Volume 16, Rev. 2, Page 106 of 171

Attachment 1, Volume 16, Rev. 2, Page 107 of 171 5.5 CTS 1

INSERT 8 (continued)

DOC M11 The program shall identify the Functions described in Paragraph a. that are automatic protective devices related to variables having significant safety functions as delineated by 10 CFR 50.36(c)(1)(ii)(A). The NTSP of these Functions are Limiting Safety System Settings. These Functions shall be demonstrated to be functioning as required by applying the following requirements during CHANNEL CALIBRATIONS, CHANNEL OPERATIONAL TESTS, and TRIP ACTUATING DEVICE OPERATIONAL TESTS that verify the NTSP.

1. The as-found value of the instrument channel trip setting shall be compared with the previous as-left value or the specified NTSP.

2. If the as-found value of the instrument channel trip setting differs from the previous as-left value or the specified NTSP by more than the pre-defined test acceptance criteria band (i.e., the specified AFT), then the instrument channel shall be evaluated before declaring the SR met and returning the instrument channel to service. This condition shall be entered in the plant corrective action program.

3. If the as-found value of the instrument channel trip setting is less conservative than the specified AV, then the SR is not met and the instrument channel shall be immediately declared inoperable.

4. The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NTSP at the completion of the surveillance test; otherwise, the channel is inoperable (setpoints may be more conservative than the NTSP provided that the as-found and as-left tolerances apply to the actual setpoint used to confirm channel performance).

e. The program shall be specified in [insert the facility FSAR reference or the name of any document incorporated into the facility FSAR by reference].

the Technical Requirements Manual Insert Page 5.5-18d Attachment 1, Volume 16, Rev. 2, Page 107 of 171

Attachment 1, Volume 16, Rev. 2, Page 108 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

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

2. This Specification has been renumbered to be consistent with the ITS Format and for clarity.

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

4. The Primary Containment Sources Outside Containment is currently called the System Integrity Program (SIP) at Kewaunee Power Station (KPS). To maintain consistency with ISTS 5.5.2 the title of the section has not been changed, but the site specific title for the program has been added to the description. This is acceptable since the ISTS has not been changed and the site can maintain the name of the program used in the CTS.

5. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. The systems provided are the systems KPS currently tests under the System Integrity Program (SIP) requirements of CTS 6.1.2. This is acceptable since the generic specific information/value is revised to reflect the current plant design.

6. The bracketed ISTS 5.5.3, "Post Accident Sampling," and the ISTS 5.5.6, "Pre-Stressed Concrete Containment Tendon Surveillance Program," are not included in the Kewaunee Power Station (KPS) ITS. The ISTS 5.5.3 requirement for Post Accident Sampling was deleted in License Amendment 160, dated January 16, 2002 (ADAMS Accession No. ML020170187) and the ISTS 5.5.6 requirement is not included since KPS does not have pre-stressed concrete tendons. Subsequent programs in ITS Section 5.5 have been renumbered as necessary.

7. Typographical/grammatical error corrected.

8. The ISTS 5.5.7 Reviewer's Note states that the inspection interval and scope for RCP flywheels stated above (i.e., the second paragraph of ISTS 5.5.7) can be applied to plants that satisfy the requirements of WCAP-15666. The current Kewaunee Power Station commitment with respect to the flywheel inspections is at a 20 year interval (consistent with WCAP-15666), as documented in the letter from Thomas Coutu (NMC) to the NRC, "Response to Request for Additional Information Related to Third and Fourth 10-Year In-Service Inspection Intervals Request for Relief for Nuclear Management Company, LLC," dated September 17, 2004 (ADAMS accession No. ML042720366). Therefore, the second paragraph allowances have been maintained in ITS 5.5.5. Furthermore, the Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This is not meant to be retained in the final version of the plant specific submittal.

Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 16, Rev. 2, Page 108 of 171

Attachment 1, Volume 16, Rev. 2, Page 109 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

9. The Inservice Testing Program (ISTS 5.5.8) has been modified to state that the IST Program provides control for ASME Code Class 1, 2, and 3 "pumps and valves" in place of the current "components." 10 CFR 50.55a(f) provides the regulatory requirements for the IST Program. It specifies that ASME Code Class 1, 2, and 3 pumps and valves are the only components covered by an IST Program. 10 CFR 50.55a(g) provides regulatory requirements for an Inservice Inspection (ISI) Program. It specifies that ASME Code Class 1, 2, and 3 components are covered by the ISI Program, and that pumps and valves are covered by the IST Program in 10 CFR 50.55a(f). The ISTS does not include ISI Program requirements as these requirements have been relocated to a plant specific document. Therefore, the components to which the IST Program applies (i.e., pumps and valves) have been added for clarity. In addition, the statement "The program shall include the following:" has been deleted because not all of the statements that follow are really part of the program requirements.

Furthermore, the terms weekly, semiannually, and every 9 months have been deleted since these terms are not used in the ASME OM Code.

10. Not Used.

11. ISTS 5.5.11 uses the term "Engineered Safety Feature (ESF)" to describe the ventilation systems tested as part of this Specification. The three ventilation systems covered in ITS 5.5.9 are the Shield Building Ventilation System (SBVS),

Auxiliary Building Special Ventilation (ASV) System, and Control Room Post Accident Recirculation (CRPAR) System. The KPS CRPAR System is not an ESF. Therefore, the term has been changed to "safety related," since all three of the ventilations systems are safety related.

12. The Reviewer's Note to ISTS 5.5.9.c and the subsequent wording states alternate tube repair criteria that are currently permitted by the plant technical specifications should be provided in the ITS. ISTS 5.5.9.f (ITS 5.5.7.f), including the Reviewer's Note, states, in part, the tube repair methods currently permitted by plant technical specifications should be provided in the ITS. The bracketed allowance to provide steam generator tube repair criteria and methods are not included since the current KPS Steam Generator Program does not allow repair; only plugging is allowed.

13. Kewaunee Power Station has steam generators with Alloy 690 thermally treated tubing. Therefore the third option is maintained, consistent with the current Technical Specifications.

14. Changes are made to the ISTS which reflect the current licensing bases for KPS.

15. The Reviewer's Note contains information 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. Therefore, the Reviewer's Note has been deleted.

16. Not Used.

17. ISTS 5.5.13.c requires the total particulate concentration of the fuel oil to be tested every 31 days. The current test frequency at KPS is 92 days (per plant Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 16, Rev. 2, Page 109 of 171

Attachment 1, Volume 16, Rev. 2, Page 110 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS procedures). ITS 5.5.11.c has been changed to be consistent with current KPS practices. KPS has reviewed the maintenance history of this test and determined that the proposed 92 day Frequency is sufficient to ensure total particulates stays within the new ITS 5.5.11.c limit of 10 mg/l. In addition, the KPS diesel storage tanks are outdoor tanks, subject to the weather. Thus, minimizing the number of times the tanks must be opened to obtain fuel oil samples will also benefit keeping snow, rain water, and other contaminants out of the storage tanks.

18. Changes are made to the ISTS which reflect the plant specific nomenclature.

19. Kewaunee Power Station (KPS) complies with Option B of 10 CFR 50, Appendix J. Therefore, the ISTS 5.5.16 Option A and combined Option A and B provisions have been deleted.

20. ISTS 5.5.16.a (ITS 5.5.14.a) contains exceptions to Regulatory Guide (RG) 1.163. The KPS Containment Leak Rate Testing Program does not take any exceptions to the RG 1.163 requirements. Therefore, these exceptions are deleted.

21. Not Used.

22. KPS does not include a separate overall air lock leakage limit; it is only included as part of the combined Types B and C leakage limit (0.60 La). Therefore, ISTS 5.5.16.d.2.a) has not been included. Due to this, there is no reason to include the requirements of ISTS 5.5.16.d.2.b) separate from ISTS 5.5.16.d.2. Thus it has been combined into ISTS 5.5.16.d.2. Furthermore, ISTS 5.5.16.d.2.b) states, in part, the air lock acceptance criteria for each door. The CTS 6.20.c states, in part, the air lock acceptance criteria for each air lock door seal. ITS 5.5.14.d.2) is written to address each air lock door seal. This is acceptable since the ITS is edited to reflect the text in the CTS and for clarification. Lastly, ISTS 5.5.16.d.2.b) (ITS 5.5.14.d) contains a bracketed value for the air lock door seal containment leakage rate acceptance criteria and the pressure to which each door seal is tested. The brackets have been removed for the pressure to which each door seal is tested and an acceptance criteria value of < 0.005 La has been provided consistent with CTS 6.20. This is acceptable since the generic specific information/value is revised to reflect the current plant design.

23. ISTS 5.5.16 (ITS 5.5.14) provides the requirements for the Containment Leakage Rate Testing Program. The statement in ISTS 5.5.16.f that "Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J" has been deleted. This phrase is not consistent with the allowances in ISTS 5.5.16.a (ITS 5.5.14.a), which states that "A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions." Thus the ISTS clearly allows exemptions to be made.

In addition, the ISTS clearly states (in the remainder of ITS 5.5.14.a) that "This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163," and RG 1.163 includes the appropriate Surveillance test Frequencies and how they can be adjusted.

Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 16, Rev. 2, Page 110 of 171

Attachment 1, Volume 16, Rev. 2, Page 111 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

24. ISTS 5.5.17, "Battery Monitoring and Maintenance Program," was added to NUREG-1431, Rev. 2 as part of TSTF-360, Rev. 1, which was approved by the NRC on December 18, 2000 (documented in a letter from W. D. Beckner, NRC to A. R. Pietrangelo, NEI). In a NRC subsequent NRC letter to the Technical Specification Task Force (TSTF), dated April 11, 2006, the NRC expressed concerns about certain aspects of the changes approved in TSTF-360, Rev.1.

Since that time, the NRC and TSTF have had ongoing discussions concerning this program, and a new TSTF, proposed TSTF-500, Rev. 2, has been generated to resolve the NRC concerns. KPS is proposing revisions to the program based this proposed TSTF. The proposed changes are discussed below.

The Battery Monitoring and Maintenance Program is revised to reference IEEE-450-2002 and Regulatory Guide 1.129, Revision 2 (with exceptions), to require actions to equalize and test battery cells when the electrolyte level drops below the top of plates instead of when the electrolyte level drops below the minimum established design limit, to require actions to verify the remaining cells are > 2.13 V when a cell or cells have been found to be < 2.13 V. The Program is also revised to state the license controlled program will contain limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and a requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

The exceptions to Regulatory Guide 1.129, Revision 2, are needed to make the Regulatory Guide requirements consistent with the proposed Technical Specification requirements, allow reasonable technical approaches, and be applicable to operating plants, as described below:

Exception 1: Regulatory Guide 1.129 states that temperature correction must be performed before and after the test. IEEE-450-2002 recommends performing temperature correction before or after the test and this is adequate to obtain accurate test results.

Exception 2: This change excludes the Regulatory Guide 1.129 referenced documents, as they are not relevant to the program.

Exception 3: Regulatory Guide 1.129, Regulatory Position 2, states, "Where reference is made to the pilot cell, pilot cell selection shall be based on finding an average cell that is representative of the entire batterys individual cell voltage and specific gravity readings." This position is inconsistent with the treatment of pilot cells in TSTF-500. As stated in the justification (above), "In the past, pilot cells were selected to represent average cells in the battery. The change to 2.07 V now requires pilot cells to be selected to represent the lowest voltage cells in the battery. This ensures that the other cells are above the pilot cell voltage which must remain above the TS limit."

Exception 4: The following statements are excluded from Regulatory Position 3, subsection 5.4.1, "When it has been recorded that the charging current has stabilized at the charging voltage for three consecutive hourly measurements, the battery is near full charge. These Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 16, Rev. 2, Page 111 of 171

Attachment 1, Volume 16, Rev. 2, Page 112 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS measurements shall be made after the initially high charging current decreases sharply and the battery voltage rises to approach the charger output voltage." This is inconsistent with the OPERABILITY requirements used in the KPS ITS, which state that verifying battery float current to be 2 amps while on float charge determines the battery is fully charged (See SR 3.8.6.1).

Exception 5: Regulatory Guide 1.129, Regulatory Position 7, recommends recording the specific gravity and float voltage of each cell in the string following the test. The Battery Monitoring and Maintenance Program requires obtaining specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations. The provision to follow the manufacturer's recommendations is a reasonable allowance given that the battery manufacturer is qualified to determine the benefit of the readings.

25. ISTS 5.5.18, "Setpoint Control Program (SCP)," has been added consistent with proposed TSTF-493, Revision 4. Any changes to the proposed program are discussed in other Justification for Deviations. In addition, the bracketed ISTS 5.5.3, "Post Accident Sampling," and the ISTS 5.5.6, "Pre-Stressed Concrete Containment Tendon Surveillance Program," are not included in the Kewaunee Power Station (KPS) ITS. Therefore, this Specification has been renumbered in the KPS ITS as 5.5.16.

26. Changes are made to be consistent with the LCO title in Section 3.3. In addition, ISTS 3.3.8 and 3.3.9 have not been adopted in the KPS ITS.

27. The NRC has not approved processing changes using the KPS setpoint methodology such that the Setpoint Control Program can be implemented as described in TSTF-493, Rev. 4 and shown in Insert 8. Therefore, based on discussions with the NRC staff, changes have been made to the Setpoint Control Program that essentially requires any changes to the NTSP, AV, AFT, and ALT be reviewed and approved by the NRC.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 16, Rev. 2, Page 112 of 171

Attachment 1, Volume 16, Rev. 2, Page 113 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 113 of 171

Attachment 1, Volume 16, Rev. 2, Page 114 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 114 of 171

, Volume 16, Rev. 2, Page 115 of 171 ATTACHMENT 6 ITS 5.6, REPORTING REQUIREMENTS , Volume 16, Rev. 2, Page 115 of 171

, Volume 16, Rev. 2, Page 116 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 116 of 171

Attachment 1, Volume 16, Rev. 2, Page 117 of 171 ITS A01 ITS 5.6

3. Deleted.

5.6.3 4. Core Operating Limits Report (COLR) 5.6.3.a A. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

5.6.3.a.1 (1) TS 2.1 Reactor Core Safety Limit 5.6.3.a.9 (2) TS 2.3.a.3.A Overtemperature T Setpoint 5.6.3.a.9 (3) TS 2.3.a.3.B Overpower T Setpoint 5.6.3.a.3 (4) TS 3.1.f.3 Moderator Temperature Coefficient (MTC) 5.6.3.a.11 (5) TS 3.8.a.5 Refueling Boron Concentration 5.6.3.a.2 (6) TS 3.10.a Shutdown Margin 5.6.3.a.6 (7) TS 3.10.b.1.A F Q N (Z) Limits 5.6.3.a.7 (8) TS 3.10.b.1.B F H N Limits 5.6.3.a.6 (9) TS 3.10.b.5 F Q N(Z) Limits 5.6.3.a.6 (10) TS 3.10.b.6.C.i F Q N(Z) penalty 5.6.3.a.8 (11) TS 3.10.b.8 Axial Flux Difference Target Band 5.6.3.a.8 (12) TS 3.10.b.8.A Axial Flux Difference Envelope 5.6.3.a.4 (13) TS 3.10.d.1 Shutdown Bank Insertion Limits 5.6.3.a.5 (14) TS 3.10.d.2 Control Bank Insertion Limits 5.6.3.a.10 (15) TS 3.10.k Core Average Temperature 5.6.3.a.10 (16) TS 3.10.l Reactor Coolant System Pressure 5.6.3.a.10 (17) TS 3.10.m.1 Reactor Coolant Flow 5.6.3.b B. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. When an initial assumed power level of 102% of the original rated power is specified in a previously approved method, 100.6% of uprated power may be used only when the main feedwater flow measurement (used as the input for reactor thermal output) is provided by the Crossflow ultrasonic flow measurement system (Crossflow system) as described in report (15) listed below. When main feedwater flow measurements from the Crossflow System are unavailable, a power measurement uncertainty consistent with the instrumentation used shall be applied.

Future revisions of approved analytical methods listed in this Technical Specification that currently reference the original Appendix K uncertainty of 102% of the original rated power should include the condition given above allowing use of 100.6% of uprated power in the safety analysis methodology when the Crossflow system is used for main feedwater flow measurement.

Amendment No. 196 TS 6.9-3 03/28/2008 Page 1 of 5 Attachment 1, Volume 16, Rev. 2, Page 117 of 171

Attachment 1, Volume 16, Rev. 2, Page 118 of 171 ITS A01 ITS 5.6 5.6.3.b The approved analytical methods are described in the following documents.

(1) Deleted (2) Kewaunee Nuclear Power Plant - Review For Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP.

(3) S.M. Bajorek, et al., WCAP-12945-P-A (Proprietary),

Westinghouse Code Qualification Document for Best-Estimate Loss-of -Coolant Accident Analysis, Volume I, and Volume II-V.

(4) N. Lee et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A.

(5) C.M. Thompson, et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection into the Broken Loop and COSI Condensation Model," WCAP-10054-P-A, Addendum 2.

(6) XN-NF-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company.

(7) ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation.

(8) EMF-92-116 (P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation.

(9) WCAP-10216-P-A, Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification.

(10) WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology.

(11) WCAP-8745-P-A, Design Bases for the Thermal Overtemperature T and Thermal Overpower T trip functions.

Amendment No. 196 TS 6.9-4 03/28/2008 Page 2 of 5 Attachment 1, Volume 16, Rev. 2, Page 118 of 171

Attachment 1, Volume 16, Rev. 2, Page 119 of 171 ITS A01 ITS 5.6 (12) S.I. Dederer, et al., WCAP-14449-P-A, Application of Best-Estimate Large-Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection.

(13) WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report.

(14) WCAP-11397-P-A, Revised Thermal Design Procedure.

(15) CENP-397-P-A, Improved Flow Measurement Accuracy Using Cross Flow Ultrasonic Flow Measurement Technology.

(16) Topical Report DOM-NAF-5-A, Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS).

5.6.3.c C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

5.6.3.d D. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.3.b E. The COLR will contain the complete identification of the TS approved analytical methods used to prepare the COLR (i.e. report number, title, revision, date, and any supplements).

Amendment No. 196 TS 6.9-5 03/28/2008 Page 3 of 5 Attachment 1, Volume 16, Rev. 2, Page 119 of 171

Attachment 1, Volume 16, Rev. 2, Page 120 of 171 ITS ITS 5.6 A01

b. Unique Reporting Requirements 5.6.1 1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 by May 15 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program L01 for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

Add proposed ITS 5.6.1 second paragraph M01 5.6.2 2. Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operation of the unit for the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications. See CTS 6.0 (1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

5.6.5 4. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into INTERMEDIATE SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.22, Steam Generator (SG)

Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, Amendment No. 188 TS 6.9-6 Revised by letter dated August 29, 2006 Page 4 of 5 Attachment 1, Volume 16, Rev. 2, Page 120 of 171

Attachment 1, Volume 16, Rev. 2, Page 121 of 171 ITS A01 ITS 5.6 5.6.5 d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging in each SG.

Add proposed ITS 5.6.4 M02 Amendment No. 188 TS 6.9-7 Revised by letter dated August 29, 2006 Page 5 of 5 Attachment 1, Volume 16, Rev. 2, Page 121 of 171

Attachment 1, Volume 16, Rev. 2, Page 122 of 171 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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

MORE RESTRICTIVE CHANGES M01 The second paragraph of ITS 5.6.1 includes details required to be included in the Annual Radiological Environmental Operating Report. CTS 6.9.b.1.A does not contain this level of detail. This changes the CTS by requiring additional detail to be included in the Annual Radiological Environmental Operating Report.

The purpose of the second paragraph of ITS 5.6.1 is to specify details to be included in the Annual Radiological Environmental Operating Report. This change is acceptable because the content requirements are consistent with the objectives outlined in the Offsite Dose Calculation Manual. This change is designated more restrictive because it adds new reporting requirements to the Technical Specifications.

M02 The CTS does not include program requirements for the Post Accident Monitoring Report. The ITS includes a requirement for this report. This changes the CTS by adding the Post Accident Monitoring Report requirement.

The purpose of ITS 5.6.4 is to specify details to be included in the Post Accident Monitoring Report. This change is acceptable because it supports implementation of the requirements of the ITS. This change is designated as more restrictive because it adds new reporting requirements to the Technical Specifications.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 -Relaxation of LCO Requirements) CTS 6.9.b.1.A requires the Annual Radiological Environmental Operating Report to be submitted before May Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 16, Rev. 2, Page 122 of 171

Attachment 1, Volume 16, Rev. 2, Page 123 of 171 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS 1 of the year. ITS 5.6.1 requires the Annual Radiological Environmental Operating Report to be submitted by May 15 of each year. This changes the CTS by allowing additional time to submit this report each year.

The purpose of the due date for submitting the Annual Radiological Environmental Operating Report is to ensure that the report is provided in a reasonable period of time to the NRC for review. This change is acceptable because the report is still required to be provided to the NRC on or before May 15 and cover the previous calendar year, report completion and submittal is clearly not necessary to assure operation in a safe manner for the interval between May 1 and May 15. Additionally, there is no requirement for the NRC to approve the report. This change is designated as less restrictive because it allows more time to prepare and submit the report to the NRC.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 16, Rev. 2, Page 123 of 171

Attachment 1, Volume 16, Rev. 2, Page 124 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 124 of 171

Attachment 1, Volume 16, Rev. 2, Page 125 of 171 Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Annual Radiological Environmental Operating Report

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

[ A single submittal may be made for a multiple unit station. The submittal should 1 combine sections common to all units at the station. ]

6.9.b.1 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

DOC M01 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements [in the format of the 2 table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979]. In the event that some individual results are not available for 2 inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.2 Radioactive Effluent Release Report

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

[ A single submittal may be made for a multiple unit station. The submittal shall combine sections common to all units at the station; however, for units with 1 separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. ]

6.9.b.2 The Radioactive Effluent Release Report covering the operation of the unit in the by previous year shall be submitted prior to May 1 of each year in accordance with 3 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

WOG STS 5.6-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 125 of 171

Attachment 1, Volume 16, Rev. 2, Page 126 of 171 Reporting Requirements 5.6 5.6 Reporting Requirements 6.9.a.4 5.6.3 CORE OPERATING LIMITS REPORT (COLR) 6.9.a.4.A a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: INSERT 1

[ The individual specifications that address core operating limits must be 2 referenced here. ]

6.9.a.4.B b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those INSERT 2 described in the following documents:

2

[ Identify the Topical Report(s) by number and title or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date.

6.9.a.4.E The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). ] 2 6.9.a.4.C c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.a.4.D d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]

4

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

[ Identify the Topical Report(s) by number and title or identify the NRC Safety Evaluation for a plant specific methodology by NRC letter and date.

The PTLR will contain the complete identification for each of the TS referenced Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements) .]

WOG STS 5.6-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 126 of 171

Attachment 1, Volume 16, Rev. 2, Page 127 of 171 CTS 5.6 2

1 INSERT 1 6.9.a.4.A 1. SL 2.1.1, "Reactor Core SLs";

2. LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

3. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";

4. LCO 3.1.5, "Shutdown Bank Insertion Limits";

5. LCO 3.1.6, "Control Bank Insertion Limits";

6. LCO 3.2.1, "Heat Flux Hot Channel Factor (F Q (Z))";

7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( FNH )";

8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

9. LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation," Functions 6 and 7 (Overtemperature T and Overpower T, respectively);

10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and

11. LCO 3.9.1, "Boron Concentration."

INSERT 2 2

6.9.a.4.B . When an initial assumed power level of 102% of the original RATED THERMAL POWER is specified in a previously approved method, 100.6% of uprated RATED THERMAL POWER may be used only when the main feedwater flow measurement (used as the input for reactor thermal output) is provided by the Crossflow Ultrasonic Flow Measurement System (Crossflow System) as described in the Reference listed in Specification 5.6.3.b.14 below. When main feedwater flow measurements from the Crossflow System are unavailable, a power measurement uncertainty consistent with the instrumentation used shall be applied.

Future revisions of approved analytical methods listed in this Technical Specification that currently reference the original Appendix K uncertainty of 102% of the original rated power should include the condition given above allowing use of 100.6% of uprated power in the safety analysis methodology when the Crossflow System is used for main feedwater flow measurement.

The approved analytical methods are described in the following documents:

Insert Page 5.6-2a Attachment 1, Volume 16, Rev. 2, Page 127 of 171

Attachment 1, Volume 16, Rev. 2, Page 128 of 171 5.6 CTS 2 INSERT 2 (continued) 6.9.a.4.B 1. Topical Report WPSRSEM-NP, "Kewaunee Nuclear Power Plant - Review For Kewaunee Reload Safety Evaluation Methods."

2. WCAP-12945-P-A (Proprietary), "Westinghouse Code Qualification Document for Best-Estimate Loss-of-Coolant Accident Analysis," Volume I and Volume II-V.

3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code."

4. WCAP-10054-P-A, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," Addendum 2.

5. XN-NF-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup."

6. ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU."

7. EMF-92-116 (P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs."

8. WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification."

9. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology."

10. WCAP-8745-P-A, "Design Bases for the Thermal Overtemperature T and Thermal Overpower T trip functions."

11. WCAP-14449-P-A, "Application of Best-Estimate Large-Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection."

12. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."

13. WCAP-11397-P-A, "Revised Thermal Design Procedure."

14. CENP-397-P-A, "Improved Flow Measurement Accuracy Using Cross Flow Ultrasonic Flow Measurement Technology."

15. Topical Report DOM-NAF-5-A, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station (KPS)."

Insert Page 5.6-2b Attachment 1, Volume 16, Rev. 2, Page 128 of 171

Attachment 1, Volume 16, Rev. 2, Page 129 of 171 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

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

The methodology for the calculation of the P-T limits for NRC approval should include the following provisions:

1. The methodology shall describe how the neutron fluence is calculated (reference new Regulatory Guide when issued).

2. The Reactor Vessel Material Surveillance Program shall comply with Appendix H to 10 CFR 50. The reactor vessel material irradiation surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update the PTLR curves.

3. Low Temperature Overpressure Protection (LTOP) System lift setting limits for the Power Operated Relief Valves (PORVs), developed using NRC-approved methodologies may be included in the PTLR.

4. The adjusted reference temperature (ART) for each reactor beltline material 4 shall be calculated, accounting for radiation embrittlement, in accordance with Regulatory Guide 1.99, Revision 2.

5. The limiting ART shall be incorporated into the calculation of the pressure and temperature limit curves in accordance with NUREG-0800 Standard Review Plan 5.3.2, Pressure-Temperature Limits.

6. LTOP arming temperature limit development methodology.

7. The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

8. Licensees who have removed two or more capsules should compare for each surveillance material the measured increase in reference temperature (RTNDT) to the predicted increase in RTNDT; where the predicted increase in RTNDT is based on the mean shift in RTNDT plus the two standard deviation value (2) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology.

WOG STS 5.6-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 129 of 171

Attachment 1, Volume 16, Rev. 2, Page 130 of 171 Reporting Requirements 5.6 5.6 Reporting Requirements DOC M02 5.6.5 Post Accident Monitoring Report 4 4

When a report is required by Condition B or F of LCO 3.3.[3], "Post Accident 8 2 Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 [ Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance 5 Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken. ]

6.9.b.4 5.6.7 Steam Generator Tube Inspection Report 5 5

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include: 7 7

a. The scope of inspections performed on each SG, 6

b. Active degradation mechanisms found,

6

c. Nondestructive examination techniques utilized for each degradation mechanism, 6

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, 6

e. Number of tubes plugged [or repaired] during the inspection outage for each 2 active degradation mechanism, 6 2

f. Total number and percentage of tubes plugged [or repaired] to date, 6

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, 6

and

[h. The effective plugging percentage for all plugging [and tube repairs] in each 2 SG, and] 6

[i. Repair method utilized and the number of tubes repaired by each repair 2 method.]

WOG STS 5.6-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 2, Page 130 of 171

Attachment 1, Volume 16, Rev. 2, Page 131 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.6, REPORTING REQUIREMENTS

1. Kewaunee Power Station (KPS) is a single unit site. Therefore, the allowance provided by this reviewers Note is not needed and has not been adopted in the KPS ITS.

2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.

3. Changed the ISTS 5.6.2 submittal date to be consistent with the current KPS submittal date in CTS 6.9.b.2.

4. ISTS 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," is not adopted in the ITS. CTS Figures TS 3.1-1 and TS 3.1-2, which provide Reactor Coolant System heatup and cooldown limitations, respectively, were adopted in ITS 3.4.3, "RCS Pressure and Temperature (P/T) Limits." Subsequent Specifications are renumbered accordingly.

5. The bracketed ISTS 5.6.6, "Tendon Surveillance Report," is not included in the Kewaunee Power Station (KPS) ITS since KPS does not have pre-stressed concrete tendons. Subsequent Specifications are renumbered accordingly.

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

7. ISTS 5.5.3, "Post Accident Sampling," and ISTS 5.5.6, "Pre-Stressed Concrete Containment Tendon Surveillance Program," are not included in the KPS ITS.

As a result, subsequent programs in ITS Section 5.5 have been renumbered and Specification 5.5.9 is now 5.5.7.

8. Changes have been made due to changes made to another Specification.

Condition F in ITS 3.3.3 has not been adopted, thus it is being deleted in ITS 5.6.4.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 131 of 171

Attachment 1, Volume 16, Rev. 2, Page 132 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 132 of 171

Attachment 1, Volume 16, Rev. 2, Page 133 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 133 of 171

, Volume 16, Rev. 2, Page 134 of 171 ATTACHMENT 7 ITS 5.7, HIGH RADIATION AREA , Volume 16, Rev. 2, Page 134 of 171

, Volume 16, Rev. 2, Page 135 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 135 of 171

Attachment 1, Volume 16, Rev. 2, Page 136 of 171 A01 ITS ITS 5.7 5.7 6.13 HIGH RADIATION AREA INSERT 1 5.7

a. In lieu of the "control device" or "alarm signal" required by Paragraph 20.1601(a) of 5.7.1 10 CFR Part 20,each high radiation area in which the intensity of radiation is > 100 5.7.1.a, 5.7.1.b, mrem/hr, but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high A02 5.7.2.a, 5.7.2.b radiation area and entrance thereto shall be controlled by requiring issuance of a radiation work permit (RWP).(1) Any individual or group of individuals permitted to enter L01 5.7.1.d such areas shall be provided with or accompanied by one or more of the following. INSERT 2 M01 5.7.1.d.1 1. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

INSERT 3 L02 5.7.1.d.2 2. A radiation monitoring device which continuously integrates the radiation dose in the 5.7.2.d.1 area and alarms when a preset integrated dose is received. Entry into such areas M02 5.7.1.e, with this monitoring device may be made after the dose rate level in the area has INSERT 4 5.7.2.e been established and personnel have been made knowledgeable of them.

Add proposed ITS 5.7.1.d.3

3. A health physics qualified individual (i.e., qualified in radiation protection L03 procedures) with a radiation dose rate monitoring device who is responsible for 5.7.1.d.4, 5.7.2.d.3 providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility health L04 physicist in the RWP. Add proposed ITS 5.7.1.d.4 and 5.7.2.d.3 A02

b. In addition to the requirements of 6.13.a., areas accessible to personnel with radiation 5.7.2 levels such that a major portion of the body could receive in one hour a dose > 1000 INSERT 5 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys L01 shall be maintained under the administrative control of the shift manager on duty and/or 5.7.2.a health physics supervision. Doors shall remain locked except during periods of access or equivalent by personnel under an approved RWP which shall specify the dose rate levels in the 5.7.2.b immediate work area and the maximum allowable stay time for individuals in that area. INSERT 6 For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose > 1000 mrem(2) that are located M01 5.7.2.f within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be L04 activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be 5.7.2.d.3 (ii) made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. L03 Add proposed ITS 5.7.2.d.2 L05 Add proposed ITS 5.7.2.d.4 Individuals qualified in radiation procedures LA01 such individuals (1)

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt 5.7.1.c from the RWP issuance requirement during the performance of their assigned radiation 5.7.2.c protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.

(2) 5.7.2 Measurement made at 30 centimeters from source of radioactivity.

Amendment No. 193 TS 6.13-1 10/31/2007 Page 1 of 2 Attachment 1, Volume 16, Rev. 2, Page 136 of 171

Attachment 1, Volume 16, Rev. 2, Page 137 of 171 ITS 5.7 A02 INSERT 1 at 30 centimeters from the radiation source or from any surface penetrated by the radiation M01 INSERT 2 L01 or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

L02 INSERT 3 Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, M02 INSERT 4 These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

A02 INSERT 5 at 30 centimeters and from the radiation source or from any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation M01 INSERT 6 other appropriate radiation protection equipment and measures.

Insert Page 6.13-1 Page 2 of 2 Attachment 1, Volume 16, Rev. 2, Page 137 of 171

Attachment 1, Volume 16, Rev. 2, Page 138 of 171 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

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

A02 CTS 6.13.a applies for control of entry into high radiation areas in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr.

CTS 6.13.b applies for control of entry into high radiation areas in which a major part of the body could receive a dose greater than 1000 mrem/hr and is clarified by Footnote (2) that measurement is made at 30 centimeters from the source of radioactivity. ITS 5.7.1 applies to controls for high radiation areas with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation. ITS 5.7.2 applies to controls for high radiation areas with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation, but less than 500 rads/hr at one meter from a radiation source or any surface through which radiation penetrates. This changes the CTS by deleting the reference to a high radiation area having radiation intensity in excess of 100 mrem/hr and adds the criteria of, "at 30 centimeters from the radiation source or from any surface penetrated by the radiation" to the parameter 1000 mrem/hr.

These changes are acceptable because the 100 mrem/hr to less than 500 rads/hr definition for a high radiation area is already addressed by 10 CFR 20.1003, and the method of measuring the 1000 mrem/hr is clarified in terms of being measured from a point source and from a surface. This change is designated administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 6.13.a, in reference to entrance into a high radiation area states "entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)." CTS 6.13.b states, in part, "access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area." ITS 5.7.1.b and 5.7.2.b state, "Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures." This changes the CTS by specifying certain information is required to be in the RWP or equivalent. The addition of the option to use a means equivalent to the RWP is addressed in DOC L01.

Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 16, Rev. 2, Page 138 of 171

Attachment 1, Volume 16, Rev. 2, Page 139 of 171 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA The purpose of the RWP requirement in CTS 6.13.a and CTS 6.13.b is to ensure personnel entering a high radiation area have the information necessary to work safely in those areas from a radiation standpoint. This change is acceptable because it states specific information to be included in the RWP to accomplish the same goal, and requiring issuance of the RWP with the required information makes the information available. These changes are designated as more restrictive because additional information to be included in the RWP is required.

M02 CTS 6.13.a.2 states that one of the optional criteria that allows entry into a high radiation area is a radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. It further states that entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them. ITS 5.7.1.e and ITS 5.7.2.e include a similar requirement, but also require that the continuously escorted personnel will receive a pre-job briefing prior to entry into such areas, and that the dose rate determination, knowledge, and pre-job briefing do not require documentation prior to initial entry. This changes the CTS by expanding the requirement to apply to all the options for conditions allowing entry into a high radiation area, and adding the criteria that the continuously escorted personnel will receive a pre-job briefing prior to entry into such areas and that the dose rate determination, knowledge, and pre-job briefing do not require documentation prior to initial entry.

The purpose of the second sentence in CTS 6.13.a.2 is to ensure personnel entering high radiation areas are aware of dose rates in the area. The proposed addition further ensures that a pre-job brief is conducted. This change is acceptable because it provides additional guidance to ensure personnel are aware of the relevant dose rates. This change is designated as more restrictive because additional criteria are added to the requirements for entering a high radiation area.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.13 Footnote (1) uses the title "Health Physics personnel." ITS 5.7.1.c and 5.7.2.c uses the generic title "Individuals qualified in radiation protection procedures." This changes the CTS by moving the specific KPS organizational title to the appropriate plant procedures and replacing them with generic titles.

The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 16, Rev. 2, Page 139 of 171

Attachment 1, Volume 16, Rev. 2, Page 140 of 171 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA allowance to relocate the specific KPS organizational titles is out of the Technical Specifications is consistent the allowance already provided in CTS 6.13. Also, this change is acceptable because the removed information will be adequately controlled in the appropriate procedures. Since ITS 5.4.1.a covers radiation protection procedures, any changes to the procedures are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 6.13.a states, for high radiation areas, "entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)." CTS 6.13.b also references use of an RWP when personnel access certain high radiation areas. ITS 5.7.1.b and ITS 5.7.2.b state, for high radiation areas, "Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent (emphasis added) that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures."

This changes the CTS by allowing an equivalent document to be used for access control. The addition of details required in the RWP is addressed by DOC M01.

The purpose of the specified phrase in CTS 6.13.a and 6.13.bis to designate the document through which access is controlled to the specified high radiation areas. This change is acceptable because a proper document is still required, but it may serve the same purpose as an RWP without having to be specifically called an RWP. This change is designated a less restrictive because an alternate document may be used for access control in lieu of an RWP.

L02 (Category 1 - Relaxation of LCO Requirements) CTS 6.13.a.2 states that entry a high radiation area when using a radiation monitoring device that continuously integrates the dose and alarms at a preset dose can be made only after the dose rate in the area has been established and personnel have been made knowledgeable of them. ITS 5.7.1.e and 5.7.2.e allows entry without this specific requirement provided the individuals are qualified in radiation protection procedures or the individuals are continuously escorted by an individual qualified in radiation protection procedures. This changes the CTS by allowing individuals qualified in radiation protection procedures or by individuals that are continuously escorted by an individual qualified in radiation protection procedures to enter a high radiation area using a radiation monitoring device that continuously integrates the dose and alarms at a preset dose prior to knowing the dose rate in the area.

The purpose of CTS 6.13.a.2 is to provide adequate protection such that individuals entering high radiation areas will not exceed predetermined dose limits. However, allowing entry prior to knowing the actual dose rates can reduce the overall dose received by personnel, since it reduces the total number of personnel that have to enter the area. The current requirements require an individual to first enter the area and determine the dose rates prior to the Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 16, Rev. 2, Page 140 of 171

Attachment 1, Volume 16, Rev. 2, Page 141 of 171 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA individual(s) that actually need to enter the area to perform work. Thus, an additional person receives a dose, above what is needed to actually perform the work. Allowing an individual qualified in radiation protection procedures to enter the area prior to knowing the dose would reduce the overall dose received by site personnel, while still ensuring that personnel are adequately protected (since at least one individual making the initial entry must be trained in radiation protection procedures). Therefore this change is considered acceptable. This change is designated as less restrictive because the ITS will allow entry into a high radiation area under certain conditions without knowing the specific dose rate in the area.

L03 (Category 1 - Relaxation of LCO Requirements) ITS 5.7.1.d.3 and 5.7.2.d.2 state that one of the options for devices an individual or group shall possess for radiation monitoring when entering a high radiation area is "A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area." ITS 5.7.2.d.2 also requires a means to communicate with and control every individual in the area. CTS 6.13.a and 6.13.b do not contain these options for an individual or group. This changes the CTS by providing an additional device an individual entering these high radiation areas must possess for radiation monitoring.

The purpose of ITS 5.7.1.d.3 and ITS 5.7.2.d.2 is to provide appropriate alternate means for monitoring the exposure of personnel in the respective high radiation areas. This change is acceptable because the means specified provide reliable means of monitoring personnel exposure. This change is designated as less restrictive because a new alternative for measuring personnel dose of personnel in high radiation areas has been provided.

L04 (Category 1 - Relaxation of LCO Requirements) CTS 6.13.a.3 states that one of the optional criteria that allow entry into a high radiation area is "An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit." CTS 6.13.b allows that in lieu of the stay time requirements of the RWP, direct or remote continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. ITS 5.7.1.d.4 states "A self reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance." ITS 5.7.2.d.3 reads the same as ITS 5.7.1.d.4, except the last phrase, "communicate with individuals in the area who are covered by such surveillance," is replaced with the phrase, Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 16, Rev. 2, Page 141 of 171

Attachment 1, Volume 16, Rev. 2, Page 142 of 171 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA "communicate with and control every individual in the area." This changes the CTS by deleting the discussion of positive controls over activities and performing radiation surveillances with a requirement for the monitoring device to have continuous dose rate displays and the responsibility to control dose rates in the area and adds an option to perform the monitoring of personnel remotely using the specified equipment and processes for a high radiation area < 1000 mrem/hr.

It further changes the CTS by specifying the individual has a self-reading pocket dosimeter when using this option.

The purpose of 6.13.a.3 is to provide the option of monitoring the exposure of individuals in high radiation areas by a separate individual qualified in radiation procedures. This change is acceptable because it provides adequate means of monitoring the personnel in the high radiation areas, but provides added flexibility for how to do it. This change is designated as less restrictive because additional methods for monitoring personnel exposure are provided.

L05 (Category 1 - Relaxation of LCO Requirements) ITS 5.7.2.d.4 permits the use of a radiation monitoring device that continuously displays radiation dose rates in the area when ITS 5.7.2.d.2 and ITS 5.7.2.d.3 are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle.

CTS 6.13.a and 6.13.b do not contain this option. This changes the CTS by providing an additional option for devices an individual entering these high radiation areas must use to control radiation dose.

The purpose of ITS 5.7.2.d.4 is to provide appropriate alternate means for monitoring the exposure of personnel in the respective high radiation areas. This change is acceptable because the means specified provide reliable means of monitoring personnel exposure. This change is designated as less restrictive because a new alternative for measuring personnel dose of personnel in high radiation areas has been provided.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 16, Rev. 2, Page 142 of 171

Attachment 1, Volume 16, Rev. 2, Page 143 of 171 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 16, Rev. 2, Page 143 of 171

Attachment 1, Volume 16, Rev. 2, Page 144 of 171 CTS All changes are 1 High Radiation Area 5.7 unless otherwise noted 5.0 ADMINISTRATIVE CONTROLS

[ 5.7 High Radiation Area ]

As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation 6.13.a

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

6.13 c. Individuals qualified in radiation protection procedures and personnel Footnote (1) continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

one of the following 6.13.a d. Each individual or group entering such an area shall possess: 2 6.13.a.1 1. A radiation monitoring device that continuously displays radiation dose 3 4 rates in the area, or ;

6.13.a.2 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or ; 3 DOC L03 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel 3

radiation exposure within the area, or ;

6.13.a.3 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or WOG STS 5.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 144 of 171

Attachment 1, Volume 16, Rev. 2, Page 145 of 171 CTS All changes are 1 High Radiation Area 5.7 unless otherwise noted 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.13.a.3 (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

6.13.a.2 e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 or 4 Centimeters from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

6.13.b,

1. All such door and gate keys shall be maintained under the manager 5

6.13.a administrative control of the shift supervisor, radiation protection manager, or his or her designees, and ; 3

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

6.13 c. Individuals qualified in radiation protection procedures may be exempted Footnote (1) from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

WOG STS 5.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 145 of 171

Attachment 1, Volume 16, Rev. 2, Page 146 of 171 CTS All changes are 1 High Radiation Area 5.7 unless otherwise noted 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters or from the Radiation Source of from any Surface Penetrated by the Radiation, but less 4 than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) one of the following

d. Each individual group entering such an area shall possess: 2 6.13.a.2 1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is 3

reached, with an appropriate alarm setpoint, or ;

DOC L03 2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or ; 3

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that 6.13.a.3, continuously displays radiation dose rates in the area; who is 6.13.b responsible for controlling personnel exposure within the area, or ;

3 (ii) Be under surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area. Specifications 5.7.2.d.2 and 5.7.2.d.3 5

DOC L05 4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displaces radiation dose rates in the area. 4 displays 6.13.a.2 e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

WOG STS 5.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 146 of 171

Attachment 1, Volume 16, Rev. 2, Page 147 of 171 CTS All changes are 1 High Radiation Area 5.7 unless otherwise noted 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters or 4 from the Radiation Source of from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.13.b f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

WOG STS 5.7-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 2, Page 147 of 171

Attachment 1, Volume 16, Rev. 2, Page 148 of 171 JUSTIFICATION FOR DEVIATIONS ITS 5.7, HIGH RADIATION AREA

1. ISTS 5.7 provides requirements for High Radiation Areas and contains bracketed information and/or values that is generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided.

2. Change made for added clarity.

3. These punctuation corrections have been made consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

4. Typographical/grammatical error corrected.

5. The proper Specification numbers have been provided.

6. 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.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 148 of 171

Attachment 1, Volume 16, Rev. 2, Page 149 of 171 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 16, Rev. 2, Page 149 of 171

Attachment 1, Volume 16, Rev. 2, Page 150 of 171 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.7, HIGH RADIATION AREA There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 16, Rev. 2, Page 150 of 171

, Volume 16, Rev. 2, Page 151 of 171 ATTACHMENT 8 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 16, Rev. 2, Page 151 of 171

, Volume 16, Rev. 2, Page 152 of 171 CTS 6.0, ADMINISTRATIVE CONTROLS , Volume 16, Rev. 2, Page 152 of 171

, Volume 16, Rev. 2, Page 153 of 171 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 2, Page 153 of 171

Attachment 1, Volume 16, Rev. 2, Page 154 of 171 CTS 6.0 6.4 TRAINING A retraining and replacement training program for the Plant Staff shall be maintained and LA01 shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI-N18.1-1971 and 10 CFR Part 55.

Amendment No. 162 TS 6.4-1 09/19/2002 Page 1 of 11 Attachment 1, Volume 16, Rev. 2, Page 154 of 171

Attachment 1, Volume 16, Rev. 2, Page 155 of 171 CTS 6.0 6.9 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the A01 appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

a. Routine Reports

1. Startup Report A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an OPERATING license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the USAR and shall in general include a description of the measured values of the OPERATING conditions or characteristics obtained during the test L01 program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within: (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) nine months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

2. Annual Reporting Requirements Routine OPERATING reports covering the operation of the unit during the previous calendar year shall be submitted prior to March 1 of each year. Items reported in L02 this category include:

A. Deleted Amendment No. 162 TS 6.9-1 09/19/2002 Page 2 of 11 Attachment 1, Volume 16, Rev. 2, Page 155 of 171

Attachment 1, Volume 16, Rev. 2, Page 156 of 171 CTS 6.0 B. Deleted.

C. Deleted.

D. This report shall document the results of specific activity analysis in which the reactor coolant exceeded the limits of TS 3.1.c.1.A during the past year. The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded.

(2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results L02 of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radioiodine concentrations.

(3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded.

(4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level.

(5) The time duration when the specific activity of the reactor coolant exceeded the radioiodine limit.

Amendment No. 179 TS 6.9-2 12/22/2004 Page 3 of 11 Attachment 1, Volume 16, Rev. 2, Page 156 of 171

Attachment 1, Volume 16, Rev. 2, Page 157 of 171 CTS 6.0

b. Unique Reporting Requirements

1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. See ITS 5.6

2. Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operation of the unit for the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

A01 (1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

4. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into INTERMEDIATE SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.22, Steam Generator (SG)

Program. The report shall include:

See ITS 5.6

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, Amendment No. 188 TS 6.9-6 Revised by letter dated August 29, 2006 Page 4 of 11 Attachment 1, Volume 16, Rev. 2, Page 157 of 171

Attachment 1, Volume 16, Rev. 2, Page 158 of 171 CTS 6.0 6.11 RADIATION PROTECTION PROGRAM

a. Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for LA02 all operations involving personnel radiation exposure.

b. Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine the airborne in-plant iodine concentrations under accident conditions. This program shall include the following:

LA03

1. Training of personnel

2. Procedures for monitoring

3. Provisions for maintenance of sampling and analysis equipment Amendment No. 162 TS 6.11-1 09/19/2002 Page 5 of 11 Attachment 1, Volume 16, Rev. 2, Page 158 of 171

Attachment 1, Volume 16, Rev. 2, Page 159 of 171 CTS 6.0 6.16 RADIOLOGICAL EFFLUENTS

a. Written procedures shall be established, implemented and maintained covering the See ITS activities referenced below: 5.4

1. Process Control Program (PCP) implementation LA05