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Volume 16, Revision 0, Davis-Besse, Unit 1 - Improved Technical Specifications Conversion, ITS Chapter 5.0 Administrative Controls.
	ML072200485
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Site:	Davis Besse
Issue date:	08/03/2007
From:	; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
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{{#Wiki_filter:Attachment 1, Volume 16, Rev. 0, Page 1 of 168 ATTACHMENT 1 VOLUME 16 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Revision 0 Attachment 1, Volume 16, Rev. 0, Page 1 of 168

Attachment 1, Volume 16, Rev. 0, Page 2 of 168 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. Deleted Current Technical Specifications Attachment 1, Volume 16, Rev. 0, Page 2 of 168

, Volume 16, Rev. 0, Page 3 of 168 ATTACHMENT 1 ITS 5.1, RESPONSIBILITY , Volume 16, Rev. 0, Page 3 of 168
, Volume 16, Rev. 0, Page 4 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 4 of 168

Attachment 1, Volume 16, Rev. 0, Page 5 of 168 ITS 5.1 ITS Add proposed approval requirement for proposed tests, experiments,] or modifications to systems or equipment that affects nuclear safety. M01 6.0 ADMINISTRATIVE CONTROLS I 6.1 RESPONSIBILITY 5.1.1 6.1.1 The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his/her absence. e fAdd proposed ITS 5.1.2 M01 6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for facility operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels up to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental See ITS]

5.2 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 Updated Safety Analysis Report.

b. 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.

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

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

6.2.2 FACILITY STAFF

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Operator shall be in the control panel area when fuel is in the reactor.

DAVIS-BESSE, UNIT 1 6-1 Amendment No. 9 42r2j-71,469g,--1 5 ,-i47-,-2-7-27 276 Page 1 of I Attachment 1, Volume 16, Rev. 0, Page 5 of 168

Attachment 1, Volume 16, Rev. 0, Page 6 of 168 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 5.1.1 requires that the plant manager or his designee approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affects nuclear safety. ITS 5.1.2 provides the requirement that a designated individual assume the responsibility for the control room command function. In MODES 1, 2, 3, and 4, ITS 5.1.2 requires the designated individual hold an active Senior Operator license. In MODE 5 or 6, ITS 5.1.2 requires the designated individual hold an active Senior Operator license or Operator license. This changes the CTS by adding an approval requirement for the plant manager or his designee and by adding requirements for the designated individual that assumes the control room command function. The purpose of the ITS 5.1.1 requirement is to provide additional assurance that the plant manager has direct responsibility for overall unit operation. The purpose of the ITS 5.1.2 requirement is to ensure that the control room command function is maintained. The change to ITS 5.1.1 is acceptable because the additional requirements ensure that the plant manager will be responsible for overall unit safe operation and shall have control over those activities necessary for safe operation and maintenance of the plant. The change to ITS 5.1.2 is acceptable because the requirement ensures that the designated individual assuming control room functions meets the appropriate qualification requirements. These changes are designated as more restrictive because they add additional requirements for the plant manager or his designee and they add control room command requirements. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Davis-Besse Page 1 of 2 Attachment 1, Volume 16, Rev. 0, Page 6 of 168

Attachment 1, Volume 16, Rev. 0, Page 7 of 168 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 16, Rev. 0, Page 7 of 168

Attachment 1, Volume 16, Rev. 0, Page 8 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 8 of 168

Attachment 1, Volume 16, Rev. 0, Page 9 of 168 CTS Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

                    ---------- -.-.------ -- -- I........RE\VEWER'S NOTE ----------

Titles for members of t1e unit staff shall be specified by use of an overall statement referencing n ANSI Standard acceptable to the N C staff from which the titles were obtaine , or an alternative title may be design ted for this position. Generally, the first thod is preferable; however, the seco method is adaptable to those u it staffs requiring special titles becaus of unique organizational struct res. The ANSI Standar shall be the same ANSI Standard refe enced in Section 5.3, 0 Unit Staff Qualifica*ons. If alternative titles are used, all r quirements of these Technical Specifi tions apply to the position with the alt rnative title as apply with the specified itle. Unit staff titles shall be specified n the Final Safety Analysis Report r Quality Assurance Plan. Unit staff tit es shall be maintained and revised usin those procedures approved for mod' ing/revising the Final Safety Analysis eport or Quality Assurance Plan. 6.1.1 5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect*_-...i-n[UcleIr saIey. m 0 ft...ge DOCM01 5.1.2 The R$hiftoSupervior (SS) shall be responsible for the control room command ri manag room command function. During any absence of the[from to assume thethecontrol control room while the unit is' in MODE 1, 2, shall 3, or 4, designated be an individual with an active Senior Operator FS Mlicense f n. During any absence of the[Jrom the control room while the unit is shi m in MODE 5 or 6. an individual with an active FS--F)F-license orR Operator T--__------ license shall be designated to assume the control room command tunction. eo Operator BVWOG STS 5.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 9 of 168

Attachment 1, Volume 16, Rev. 0, Page 10 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.1, RESPONSIBILITY

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed in to 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 brackets have been removed and the proper plant specific information/value has been provided.

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

4. Grammatical error corrected.

Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 10 of 168

Attachment 1, Volume 16, Rev. 0, Page 11 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 11 of 168

Attachment 1, Volume 16, Rev. 0, Page 12 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 12 of 168

, Volume 16, Rev. 0, Page 13 of 168 ATTACHMENT 2 ITS 5.2, ORGANIZATION , Volume 16, Rev. 0, Page 13 of 168
, Volume 16, Rev. 0, Page 14 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 14 of 168

Attachment 1, Volume 16, Rev. 0, Page 15 of 168 ITS 5.2 ITS 0 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY

                                                                                                                              -4See ITS]

5.1) 6.1.1 The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his/her absence. 5.2 6.2 ORGANIZATION 5.2.1 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for facility operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant. 5.2.1 .a a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels up to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These 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 Updated Safety Analysis Report. 5.2.1 .c b. 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 opertating, maintaining, and providing technical support to the plant to ensure nuclear safety. 5.2.1.b c. The plant manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant. 5.2.1 .d d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. 5.2.2 6.2.2 FACILITY STAFF

a. Each on duty shift shaI e composed of at least the minimum crew composition shown in Table 6,2

b. At least one licens. ator shall be in the control panel arehen fuel is in the A02 reactor.

DAVIS-BESSE, UNIT I 6-1 Amendment No. 9 -,4 2 r2 7 -,-7 6&, 8 r-l- 5-,- 5I4 ,47-2-r 276 Page 1 of 3 Attachment 1, Volume 16, Rev. 0, Page 15 of 168

Attachment 1, Volume 16, Rev. 0, Page 16 of 168 ITS 5.2 ITS 6.0 ADMINISTRATIVE CONTROLS 6.2.2 (Continued)

c. At least two licensed 9pemtors, one of which has a Senior Reacto erator license, shall be present in thec ol room while in M ODES 1, 2, 3 r4 /

5.2.2.c d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor?.

e. All CORE ALTERATI99S shall be directly supervised by either icensed Senior Reactor Operator or §ýdnior Reactor Operator Limited to Fuel Faidling who has no other * -A2

[concurrent responfsibilities during this operation./

f. Deleted 5.2.2.e g. The operations manager shall either hold or have held a senior reactor operator's license on a pressurized water reactor. The assistant operations manager shall hold a senior reactor operator license for the Davis-Besse Nuclear Power Station.

5.2.2.d 6.2.3 FACILITY STAFF OVERTIME Administrative controls shall be developed and implemented to limit the working hours of personnel who perform safety-related functions (e.g., senior reactor operators, reactor operators, auxiliary operators, health physicists, and key maintenance personnel). The controls shall include guidelines on working hours that ensure that adequate shift coverage is maintained without routine heavy use of overtime for individuals. Any deviation from the working hour guidelines shall be authorized in advance by the plant manager or his/her designees, in accordance with approved administrative procedures, and with documentation of the basis for granting the deviation. Routine deviation from the above guidelines shall not be authorized. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the plant manager or his/her designee(s) to ensure that excessive hours have not been assigned. 5.2.2.c " The individual qualified in radiation protection procedures may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence, provided immediate action is taken to fill the required position. DAVIS-BESSE, UNIT 1 6-2 Amendment No. 9,MU,-S99,-1--1-r,T., 4 4 2 72

                                                                                 -P- W   , 4 -- r 1                                                                                                       - -r 27 6 Page 2 of 3 Attachment 1, Volume 16, Rev. 0, Page 16 of 168

Attachment 1, Volume 16, Rev. 0, Page 17 of 168 ITS 5.2 ITS 6.0 ADMINISTRATIVE CONTROLS TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION# LICENSE APPLICABLE MO ES CATEGORY 1, 2,3 & 4 5&6 LO SOL 2** 1* OL 2 1 5.2.2.a Non-Licensed 2 1 5.2.2.f Shift Technical Advisor -** I None Required 5.2.2.b # Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

Does not include the sed Senior Reactor Operator or Seni r Operator Limited to Fuel Handling = 'ising CORE ALTERATIONS.

One of the two requir individuals filling the SOL positions may a o assume the STA function provided thindividual meets the qualifications for the bined SRO/STA position specified fdr Option 1 of the Commission's Policy Statient on Engineering A03 Expertise on Shi If this option is used for a shift, then the s arate STA position may be eliminatedi fnr n~tshift_ eliminated for tfiat shift

6.3 FACILITY STAFF QUALIFICATIONS 6311Ea~ch member of the facility staff shall meet or exceed the minimum qualifications of-- See ITS1 18.1-1971 for comparable positions, except for (1) the radiation protection manager who 5.3 I~h et or exceed the qualifications of Regulatory Guide 1.8, September 1975J fie- I . 7 A0 Technical Advisr who shall have a bachelor's egree or equivalent in a sceti c or engineering discipline with s eccific training in plant design/and response and analysis of/te plant for--- transients and a idents an e opera ons manager whose requirement tor a senior reac or See ITS 5.3 1 1 operator license is as s a ed in Specification 6.2.2.g.1; 6.4 Deleted 6.5 REVIEW AND AUDIT 6.5.1 Deleted 6.5.2 Deleted DAVIS-BESSE, UNIT I 6-3 Amendment No. 9,L222,3274,126_

                                                                         -86r g93,98,99,41.6gLO9T3S-1-3-,37, 4- -3 8 r t-39-,4l 4 2 r1 -69 4 74 , 4 7- 57 1-84 4 89
                                                                                                                         - r 1,25-23,6-,272-,- 276 Page 3 of 3 Attachment 1, Volume 16, Rev. 0, Page 17 of 168

Attachment 1, Volume 16, Rev. 0, Page 18 of 168 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 6.2.2.b states "At least one licensed Operator shall be in the control panel area when fuel is in the reactor." CTS 6.2.2.c states "At least two licensed Operators, one of which has a Senior Operator license, shall be present in the control room while in MODES 1, 2, 3, or 4." CTS 6.2.2.e requires all CORE ALTERATIONS to be directly supervised by a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. The ITS does not include these requirements. This changes the CTS by deleting these requirements. 10 CFR 50.54(m)(2)(iii) states "When a nuclear power unit is in 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 alteration 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." This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.54(m)(2)(iii) and 10 CFR 50.54(m)(2)(iv). This change is designated as administrative because it does not result in technical changes to the CTS. A03 CTS Table 6.2-1 footnote **states "One of the two required individuals filling the SOL positions may also assume the STA function provided the individual meets the qualifications for the combined SRO/STA position specified for Option 1 of the Commission Policy Statement on Engineering Expertise on Shift. Ifthis option is used for a shift, then the separate STA position may be eliminated for that shift." ITS 5.2.2, in part, requires the STA to meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift; it does not include this specific information. This changes the CTS by deleting this specific information. The purpose of CTS Table 6.2-1 footnote **is to provide the allowance for one of the personnel meeting the Senior Operator License requirement to also meet the requirements of the STA, as allowed in Option 1 of the Commission's Policy Statement on Engineering Expertise on Shift. The ITS already requires the STA to meet this policy statement (ITS 5.2.2.f), as described in DOC M01. Furthermore, this allowance is adequately addressed in the Commission Policy Davis-Besse Page 1 of 3 Attachment 1, Volume 16, Rev. 0, Page 18 of 168

Attachment 1, Volume 16, Rev. 0, Page 19 of 168 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Statement on Engineering Expertise on Shift, published in Generic Letter 86-04, dated February 13, 1986, and need not be specifically retained in the ITS. This change is considered acceptable since it is removing redundant requirements. This change is designated as administrative because it does not result in technical changes to the CTS. A04 CTS 6.3.1 provides, in part, qualification requirements for the Shift Technical Advisor (STA), and requires the STA to have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transient and accidents. ITS 5.2.2.f requires this individual to meet the qualification requirements of the Commission Policy Statement on Engineering Expertise on Shift for qualification requirements instead of listing the specific qualification requirements. The purpose of the CTS 6.3.1 STA requirements is to specify the minimum qualification requirements for the STA.. This change is acceptable because the qualification requirements included in the Commission Policy Statement on Engineering Expertise on Shift encompass the current STA qualification requirements. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M01 CTS Table 6.2-1 requires the minimum shift crew to include one STA when the unit is in MODE 1, 2, 3, or 4. ITS 5.2.2 requires that an individual 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, when the unit is in MODE 1, 2, 3, or 4. It furthermore states that the individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. This changes the CTS by detailing the specific responsibilities of the STA. The purpose of the CTS Table 6.2-1 and Footnote **STA requirements is to ensure that appropriate engineering expertise is available on shift. This change is acceptable because it clarifies STA requirements consistent with the Commission Policy Statement on Engineering Expertise on Shift. This change is designated as more restrictive because it provides specific details of the responsibilities of the STA. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, QAPM, IST Program,or liP) CTS 6.2.2.a and Table 6.2-1, including footnote *, provide minimum shift crew composition requirements. ITS 5.2.2 only Davis-Besse Page 2 of 3 Attachment 1, Volume 16, Rev. 0, Page 19 of 168

Attachment 1, Volume 16, Rev. 0, Page 20 of 168 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION includes the minimum shift crew composition requirements that are not already included in 10 CFR 50.54. This changes the CTS by moving the minimum shift crew composition requirements addressed by 10 CFR 50.54 to the Technical Requirements Manual (TRM). The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The minimum shift crew composition requirements for licensed operators and senior operators are also contained in 10 CFR 50.54(k), (I), and (m) and do not need to be repeated in the Technical Specifications. The minimum shift crew composition requirements for non-licensed operators are transferred from CTS Table 6.2-1 to ITS 5.2.2.a and the minimum shift crew composition requirements for the STA are transferred from CTS Table 6.2-1 to ITS 5.2.2.f. The relocation of the details of the minimum shift crew composition requirements to the TRM is acceptable considering the controls provided by regulations and the remaining requirements in the Technical Specifications. The TRM is currently incorporated by reference into the UFSAR, thus any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None Davis-Besse Page 3 of 3 Attachment 1, Volume 16, Rev. 0, Page 20 of 168

Attachment 1, Volume 16, Rev. 0, Page 21 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 21 of 168

Attachment 1, Volume 16, Rev. 0, Page 22 of 168 CTS Organization 5.2 5.2 ADMINISTRATIVE CONTROLS 6.2 5.2 Organization 6.2.1 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant. 6.2.1.a a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requireentsýincluding the plant-specific titles of U (P' those personnel fulfilling the responsibilities of the positions delineated in U/-SA. '- these Technical Specificationssallbe documented in the [FSR *Z 1 6.2.1.c b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plantL 6.2.1.b c. A specified corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safetW[_0( t----.: 6.2.1.d d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 5.2.2 Unit Staff The unit staff organization shall include the following Table 6.2-1 a. A non-licensed operator shall be assigned to ch reactor cont ninfuel and an additional non-licensed operator shall be assigned for eac control reactor is operating in MODES 1, 2, 3, or 0t. Two unit sites with

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

th units shutdown or defueled require total of three non-licensed operator or the two units. BWDOG STS 5.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 22 of 168

Attachment 1, Volume 16, Rev. 0, Page 23 of 168 CTS Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) Table 6.2-1 b. Shift crew composition may be less than the minimum requirement of Specifications and Note # 10 CFR 50.54(m)(2)(i) andý5.2.2.a and 5.2.2.f for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements,____, 6.2.2.d C. A radiation protection technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required positiorn o0 6.2.3 d. Administrative pshall be developed and implemented to limit the (7) working hours of personnel who perform safety related functions (e.g., icensed Senior ReorOperators S , licensed .e-.Operators

                                             )health physicists, auxiliary operators, and key maintenance 0

personnel. The controls shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime. Any deviation from the above guidelines shall be authorized in advance by the plant manager or rhs 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 toruieperiodicI lindependent reyieg* be conductedlto ensure that excessive hours have not been assignedý 4 2 either heldc

                                                                                                                      ýae 6.2.2.g                          e. The operations manager or assistant op        tions manager shalkoldia*

S iorOperator S 0license. - - 2 Table 6.2-1 f. +n individual shall provide advisory technical support to the unit operations (1) When the reactor is shift crew in the areas of thermal hydraulics, reactor engineering, and plant operating in MODE 1, analysis with regard to the safe operation of the unit. This individual shall 2, 3, or4 meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. BWOG STS 5.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 23 of 168

Attachment 1, Volume 16, Rev. 0, Page 24 of 168 5.2 O) INSERT I such that the individual overtime shall be reviewed monthly by the plant manager or designee O INSERT 2 The assistant operations manager shall hold a Senior Operator license for the Davis-Besse Nuclear Power Station; and Insert Page 5.2-2 Attachment 1, Volume 16, Rev. 0, Page 24 of 168

Attachment 1, Volume 16, Rev. 0, Page 25 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION

1. The brackets have been removed and the proper plant specific information/value has been provided.

2. Typographical/grammatical error corrected.

3. Davis-Besse Nuclear 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. These 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.

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

6. The referenced requirements are Specifications, not Code of Federal Regulations (CFR) requirements. Therefore, the word "Specifications" has been added to clearly state that 5.2.2.a and 5.2.2.f are Specifications.

7. The term "procedures" has been changed to "controls" in the first paragraph of ITS 5.2.2.d to be consistent with the usage of the term in the second paragraph. This is also consistent with the current licensing basis.

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

9. ISTS 5.2.2.d provides requirements for working hour limitations. These requirements are revised in ITS 5.2.2.d to reflect the Davis-Besse CTS 6.2.3 requirements which were approved by the NRC in License Amendment 212, dated November 8, 1996.

10. ISTS 5.2.2.e provides license requirements for the operations manager and assistant operations manager. These requirements are revised in ITS 5.2.2.e to reflect the Davis-Besse CTS 6.2.3 requirements, which were approved by the NRC in License Amendment 272, dated February 7, 2006.

11. ISTS 5.5.2.f provides requirements for the Shift Technical Advisor (STA). These requirements are revised in ITS 5.2.2.f to reflect the Davis-Besse CTS Table 6.2-1 MODE requirements for the STA.

Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 25 of 168

Attachment 1, Volume 16, Rev. 0, Page 26 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 26 of 168

Attachment 1, Volume 16, Rev. 0, Page 27 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 27 of 168

, Volume 16, Rev. 0, Page 28 of 168 ATTACHMENT 3 ITS 5.3, UNIT STAFF QUALIFICATIONS , Volume 16, Rev. 0, Page 28 of 168
, Volume 16, Rev. 0, Page 29 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 29 of 168

Attachment 1, Volume 16, Rev. 0, Page 30 of 168 ITS G6 ITS 5.3 0 6.0 ADMINISTRATIVE CONTROLS TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION# LICENSE APPLICABLE MODES CATEGORY l, 2,3 & 4 5&6 SOL 2** 1* OL 2 1 Non-Licensed 2 1 Shift Technical Advisor 1** None Required See ITS] 5.2 ]

       #    Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling supervising CORE ALTERATIONS.

       ** One of the two required individuals filling the SOL positions may also assume the STA function provided the individual meets the qualifications for the combined SRO/STA position specified for Option 1 of the Commission's Policy Statement on Engineering Expertise on Shift. If this option is used for a shift, then the separate STA position may be eliminated for that shift.

5.3 6.3 FACILITY STAFF QUALIFICATIONS 5.3.1 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N 18.l-1971 for comparable positions, except for (1) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, F(2) -theShft I r Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering - See ITS] 5.2 ] Sdiscipline with specific training in plant design, and response and analysis of the plant for " transients and accidentsiand (3) the operations manager whose requirement for a senior reactor operator license is as stated in Specification 6.2.2.g. 6 .Add proposed Specification 5.3.2R 6.4 Deleted 6.5 REVIEW AND AUDIT 6.5.1 Deleted 6.5.2 Deleted DAVIS-BESSE, UNIT I 6-3 Amendment No.-9y 12 7,-7, -r

                                                                      ; g 9 r 9 ,9 8 r 9 9,-1l)r6 S-1 l5 7 3'"         4"2;-"9,-1" -         -,     26, h,-2-35.-2-36--232-;- 276 Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 30 of 168

Attachment 1, Volume 16, Rev. 0, Page 31 of 168 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 that these individuals must meet all of the qualification requirements referenced in 10 CFR 55.4, ITS 5.3.1, and 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 Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 31 of 168

Attachment 1, Volume 16, Rev. 0, Page 32 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 32 of 168

Attachment 1, Volume 16, Rev. 0, Page 33 of 168 CTS Unit Staff Qualifications 53 5.0 ADMINISTRATIVE CONTROLS 6.3 5.3 Unit Staff Qualifications

       ...............................                   Z     /---octICI~AICID*0       Mdin            - -

Minimum qualifications for members of the unit staff shall be specified b use of an overall qualification statement rferencing an ANSI Standard acceptable to th -N RC staff or by specifying individual p ition qualifications. Generally, the first meth d is preferable; however, the second method is daptable to those unit staffs requiring speci qualification statements 0 because of unique ganizational structures. 6.3.1 5.3.1 Each member of the unit staff shall meet or exceed the minimum aualifications of [Regulatory Guide 1.X Revision 2, 1987, or more recent r'visions, or ANSI Standard acceptab to the NRC staff]. [The staff not co ered by Regulatory Guide 1.8 shall ret or exceed the minimum qualific ions of Regulations, Regulatory Guies, or ANSI Standards acceptable N RC staff]. DOC A02 5.3.2 For the purpse of 10 CFR 55.4, a licensed Senior R Operator (50l) and a licensed R Operator -k--Rare those individuals who, in addition to meeting the requirements of rM5.3.1, perform the functions described in 10 10 CFR 50.54(m). t BWOG STS 5.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 33 of 168

Attachment 1, Volume 16, Rev. 0, Page 34 of 168 5.3 (O INSERT 1 ANSI N18.1-1971 for comparable positions, except for the radiation protection manager and the operations manager. The radiation protection manager shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The operations manager shall be qualified as required by Specification 5.2.2.e. Insert Page 5.3-1 Attachment 1, Volume 16, Rev. 0, Page 34 of 168

Attachment 1, Volume 16, Rev. 0, Page 35 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS

1. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed in to 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 brackets are removed and the proper plant specific information/value is provided.

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

4. Change made to be consistent with the terminology used in other Specifications.

Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 35 of 168

Attachment 1, Volume 16, Rev. 0, Page 36 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 36 of 168

Attachment 1, Volume 16, Rev. 0, Page 37 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 37 of 168

, Volume 16, Rev. 0, Page 38 of 168 ATTACHMENT 4 ITS 5.4, PROCEDURES , Volume 16, Rev. 0, Page 38 of 168
, Volume 16, Rev. 0, Page 39 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 39 of 168

Attachment 1, Volume 16, Rev. 0, Page 40 of 168 ITS 5.4 ITS 6.0 ADMINISTRATIVE CONTROLS 6.6 Deleted 6.7 Deleted 5.4 6.8 PROCEDURES AND PROGRAMS 5.4.1 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below: 5.4.1.a a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, February, 1978. 1c. Surveillance and test actgfiieof safety related equipment. [0 Id. Physical Securi an implementation. A03 le. Davis-Bee EmerA"ydPlan implementation. p Si on4 5.4.1.b 4 Add proposed Specification _]( -Mo-1 5.4.1.d f. Fire Protection Program implementation. 0 5.4.1.c g. The radiological environmental monitoring program. 4l

h. Deleted.

{ Add proposed Specification 5.4.1.e } M02) Ii. Offsite Dose Calcul anual implementation, S0 6.8.2 Each procedure o I..1 above, and changes thereto, shall e ed and approved prioL to implementation et forth in 6.5.3 above. LAO1 6.8.3 Deleted 6.8.4 The following programs shall be established, implemented and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that See ITS]

could contain highly radioactive fluids during a serious transient or accident to as low as 5.5] practical levels. The systems include makeup, letdown, seal injection, seal return, low pressure injection, containment spray, high pressure injection, waste gas, primary sampling and reactor coolant drain systems. The program shall include the following: (i) Preventive maintenance and/or periodic visual inspection requirements, and DAVIS-BESSE, UNIT 1 6-5 Amendment No. 9,-7; 9&-86-,9;98, t 09,-9-89, "235,-2-4260,-2" 276 0 Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 40 of 168

Attachment 1, Volume 16, Rev. 0, Page 41 of 168 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 6.8.1 .b requires written procedures be established, implemented, and maintained covering refueling operations. CTS 6.8.1.c requires written procedures be established, implemented, and maintained covering surveillance and test activities of safety related equipment. ITS 5.4.1.a requires written procedures to be established, implemented and maintained to the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. This changes the CTS by deleting the specific wording of CTS 6.1.8.b and 6.8.1.c. This change is acceptable because the recommendations of Regulatory Guide 1.33, Appendix A, February 1978 already require procedures for refueling operations and for surveillance tests for safety related activities. This change is designated as administrative because it does not result in technical changes to the CTS. A03 CTS 6.8.1 .d and CTS 6.8.1 .e require written procedures be established, implemented, and maintained for the Physical Security Plan and the Davis-Besse Emergency Plan. The ITS does not contain these requirements. This changes the CTS by deleting the specific reference to the Security Plan and the Emergency Plan. This change is acceptable because the requirements for implementation of the Security and Emergency Plans are contained in 10 CFR 50.54(p) and 10 CFR 50.54(q), respectively. This change is designated as administrative because it does not result in technical changes to the CTS. A04 CTS 6.8.1 .i requires written procedures be established, implemented and maintained for the Offsite Dose Calculation Manual (ODCM). ITS 5.4.1 requires procedures for various activities, but does not specifically list the ODCM. This changes the CTS by removing the explicit requirements for written procedures for implementation of the ODCM. This change is acceptable because implementing procedures for ODCM are required by ITS 5.4.1.e. ITS 5.4.1.e (added as described in DOC M02) requires that written procedures be established, implemented, and maintained for all programs in ITS 5.5. ITS 5.5.1 covers the programmatic requirements for the ODCM. Therefore, it is not necessary to specifically identify each program in ITS 5.4.1. This change is designated as administrative because it does not result in technical changes to the CTS. Davis-Besse Page 1 of 3 Attachment 1, Volume 16, Rev. 0, Page 41 of 168

Attachment 1, Volume 16, Rev. 0, Page 42 of 168 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES MORE RESTRICTIVE CHANGES M01 ITS 5.4.1 .b requires written 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 CTS does not include this requirement. This changes the CTS by adopting a new requirement for emergency operating procedures. The purpose of ITS 5.4.1 .b is to ensure that written procedures are established, implemented, and maintained covering the emergency operating procedures to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. This change is acceptable because it is consistent with an existing requirement to comply with NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33, for emergency operating procedures. This change is designated as more restrictive because it imposes a new requirement for procedures within the Technical Specifications. M02 ITS 5.4.1 .e requires written procedures be established, implemented, and maintained for all programs specified in Specification 5.5. The CTS does not include this requirement for any program except the ODCM. This changes the CTS by adopting a new requirement for procedures to address all programs described in ITS 5.5. The purpose of ITS 5.4.1 .e is to ensure that written procedures are established, implemented, and maintained covering all programs specified in ITS 5.5. This change is considered acceptable because it requires written procedures, including proper procedure control to address programs required by ITS 5.5. This change is designated as more restrictive because it imposes new requirements for procedures within the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, QAPM, IST Program, or liP) CTS 6.8.2 requires that each procedure of CTS 6.8.1, and changes to these documents, be reviewed and approved prior to implementation as set forth in CTS 6.5.3. ITS 5.4 does not include this requirement. This changes the CTS by moving these details of procedure and administrative policy reviews to the QAPM. 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. Davis-Besse Page 2 of 3 Attachment 1, Volume 16, Rev. 0, Page 42 of 168

Attachment 1, Volume 16, Rev. 0, Page 43 of 168 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES 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 Regulatory Guide 1.33, Revision 2, February 1978. In accordance with these requirements, the QAPM 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. Furthermore, CTS 6.5.3 is being moved to the QAPM, as described in CTS 6.0 DOC LA01 (in this chapter). Also, this change is acceptable because these types of procedural details will be adequately controlled in the QAPM. Any changes to the QAPM are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because references for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None Davis-Besse Page 3 of 3 Attachment 1, Volume 16, Rev. 0, Page 43 of 168

Attachment 1, Volume 16, Rev. 0, Page 44 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 44 of 168

Attachment 1, Volume 16, Rev. 0, Page 45 of 168 CTS Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 6.8 5.4 Procedures 6.8.1 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities: 6.8.1.a a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1976L LW Q DOC M01 b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as 00 stated in Generic Letter 82-33k 6.8.1.g c. Quality assurance for effluent and environmental monitoring _D (D 6.8.1 .f d. Fire Protection Program implementation and 0 DOC M02 e. All programs specified in Specification 5.5. BV\OG STS 5.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 45 of 168

Attachment 1, Volume 16, Rev. 0, Page 46 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES

1. The brackets have been removed and the proper plant specific information/value has been provided.

2. These 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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 46 of 168

Attachment 1, Volume 16, Rev. 0, Page 47 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 47 of 168

Attachment 1, Volume 16, Rev. 0, Page 48 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 48 of 168

, Volume 16, Rev. 0, Page 49 of 168 ATTACHMENT 5 ITS 5.5, PROGRAMS AND MANUALS , Volume 16, Rev. 0, Page 49 of 168
, Volume 16, Rev. 0, Page 50 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 50 of 168

Attachment 1, Volume 16, Rev. 0, Page 51 of 168 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS 6.6 Deleted 6.7 Deleted 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

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

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Physical Security Plan implementation. See ITS 5.4 )

e. Davis-Besse Emergency Plan implementation.

f. Fire Protection Program implementation.

0s g, The radiological environmental monitoring program.

h. Deleted.

i. Offsite Dose Calculation Manual implementation.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed and approved prior to implementation as set forth in 6.5.3 above. 6.8.3 Deleted 5.5 6.8.4 The following programs shall be established, implemented and maintained: 5.5.2 a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include makeup, letdown, seal injection, seal return, low pressure injection, containment spray, high pressure injection, waste gas, primary sampling and reactor coolant drain systems. The program shall include the following: 5.5.2.a (i) Preventive maintenance and/or periodic visual inspection requirements, and DAVIS-BESSE, UNIT I 6-5 Amendment No. 9j-2-7; 86,--9-3-98, 419-l3 f 89-,-2351-24 260;--7-2-, 27 6 Page 1 of 20 Attachment 1, Volume 16, Rev. 0, Page 51 of 168

Attachment 1, Volume 16, Rev. 0, Page 52 of 168 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS 6.8.4.a (Continued) 6The provisions of SR 3.0.2 are applica" )le. 5.5.2.b (ii) Integrated leak test requirements for each system at refueling cycle intervals or less. / A02

b. In-Plant Radiation Moditorint A program which wil ensure the capability to accurately determ e the airborne iodine concentration in vi areas under accident conditions. This p shall include the following:

LAO1

1) Training of onnel,

2) Procedures r monitoring, and

3) Provisions for maintenance of sampling and analysis uipment.

c. Deleted 5.5.3 d. 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 MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

5.5.3.a 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM. 5.5.3.b 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2, 5.5.3.c 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM. 5.5.3.d 4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, DAVIS-BESSE, UNIT I 6-6 Amendment No. 5*,-84-l 0,-2-3-.64 276 Page 2 of 20 Attachment 1, Volume 16, Rev. 0, Page 52 of 168

Attachment 1, Volume 16, Rev. 0, Page 53 of 168 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS 6.8.4.d (Continued) 5.5.3.e 5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. 5.5.3.f 6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31 -day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 5.5.3-g 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table I1,Column 1, 5.53. h 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 5.5,3.i 9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-13 1, Iodine-] 33, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 5.5,3] 10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190. [The provisions of SR 3.0.2 and SIR 3.0.3 are applicable to the." A02 i IRadioactive Effluent Controls Program Surveillance Frequenc~ies. _/

e. Radiological Environinhtal Monitoring Program A program shall be p ided to monitor the radiation and radionucl es in the environs of the plant. The pro shall provide (]) representative measurem. ts of radioactivity in the highest potential xposure pathways, and (2) verification of th accuracy of the effluent monitoring rogram and modeling of environmental ecx ure pathways. The program shall (I) b contained in the ODCM, (2) conform to the idance of Appendix I to 10 CFR Part 50, d (3) include the following: LA02 I) Monitoring, pling, analysis, and reporting of radiation +d radionuclides in the environment' accordance with the methodology and par ineters in the ODCM,

2) A Land Use Census to ensure that changes in the use of 4eas at and beyond the SITE BOUND Y are identified and that modifications to th* monitoring program are made if r uired by the results of this census, and DAVIS-BESSE, UNIT I 6-7 Amendment No. t70; 276 Page 3 of 20 Attachment 1, Volume 16, Rev. 0, Page 53 of 168

Attachment 1, Volume 16, Rev. 0, Page 54 of 168 ) IT.S.SS IITS 5.5 6.0 ADMINISTRATIVE CONTROLS 6.8.4.e (Continued)

3) Participation' an Interlaboratory Comparison Program to sure that independent in checks on precision and accuracy of the measuremen of radioactive materials in environmu tal sample matrices are performed as part of e quality assurance pro or environmental monitoring.

5.5.10 f. Ventilation Filter Testing Program (VFTP): A program shall be established to implement the following required testing of safety related filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2", ANSI/ASME N510-1980, and ASTM D 3803-1989. 5.5.10.a 1) Demonstrate for each of the safety related systems that an in-place test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1% when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI/ASME N510-1980 at the system flowrate specified below, +/- 10%. Safety Related Ventilation System Flowrate Emergency Ventilation System 8000 cfm Control Room Emergency Ventilation System 3300 cfm 5.5.10.b 2) Demonstrate for each of the safety related systems that an in-place test of the charcoal adsorber shows a penetration and system bypass < 1% when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI/ASME N510-1980 at the system flowrate specified below, +/-10%. Safety Related Ventilation System Flowrate F1Sh1a

                   ~            uilij    Emergency Ventilation System                  8000 cfm                        CAO2 Control Room Emergency Ventilation System                        3300 cfm The periodic testing for        Shield Building Emergency Ventilation Sys    and the Control Room Emergency Venti tion System are performed once each REFU ING INTERVAL.

The need for testing fo owing painting, a fire, or a chemical release i any ventilation zone communicating with e Shield Building Emergency Ventilation S em or the Control Room Emergency Ventila

n System is as specified by the VFTP. The ethod of testing is based on Regulatory Guide .52, Revision 2, except for charcoal laborato testing which will be performed in a ce with ASTM D 3803-1989.

DAVIS-BESSE, UNIT 1 6-8 Amendment No. 170;M244,-265M 276 Page 4 of 20 Attachment 1, Volume 16, Rev. 0, Page 54 of 168

Attachment 1, Volume 16, Rev. 0, Page 55 of 168 ITS 5.5 ITS 0 6.0 ADMINISTRATIVE CONTROLS 6.8.4.f (Continued) 5.5.10..c 3) Demonstrate for each of the safety related systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D 3803-1989 at a temperature of 300 C and the relative humidity (RH) specified below. Safety Related Ventilation System Penetration RH hililding Emergency Ventilation System < 2.5% 95% Control Room Emergency Ventilation System < 2.5% 70% 5.5.1O.d 4) Demonstrate for each. of the safety related systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI/ASME N510-1980 at the system flowrate specified below,

                       +I- 1000.

Safety Related Ventilation System Delta P Flowrate nEmergency Ventilation System 6 inches Wate Gauge 8000 cfm Control Room Emergency Ventilation System 4.4 inches Water Gauge 3300 cfm 5.5.10 The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies. DAVIS-BESSE, UNIT I 6-9 Amendment No. 2447,-26i5- 276 Page 5 of 20 Attachment 1, Volume 16, Rev. 0, Page 55 of 168

Attachment 1, Volume 16, Rev. 0, Page 56 of 168 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS 6.8.4 (Continued) 5.5.8 g. 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: 5.5.8.a I1) 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 a SG inspection outage, as determined from the inservice 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 confirm that the performance criteria are being met. 5.5.8.b 2) Performance criteria for SG tube integrity: SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage. 5.5.8.b.1 a. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. 5.5.8. b.2 b. 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 I gpm per SG, except during a SG tube rupture. 5.5.8.b.3 c. The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage." DAVIS-BESSE, UNIT 1 6-10 Amendment No. 276 Page 6 of 20 Attachment 1, Volume 16, Rev. 0, Page 56 of 168

Attachment 1, Volume 16, Rev. 0, Page 57 of 168 IT.SS ITS U0) ITS 5.5 6.0 ADMINISTRATIVE CONTROLS 6.8.4.g (Continued) 5.5.8.c 3) Provisions for SG tube repair criteria.: 5.5.8.c. 1 a. Tubes found by inservice inspection to contain flaws, in a region of the tube that contains no repair, with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired. 5.5.8.c.2 b. Sleeves found by inservice inspection to contain flaws, in a region of the sleeve that contains no sleeve joint, with a depth equal to or exceeding 40% of the nominal sleeve wall thickness shall be plugged. 5.5.8.c.3 c. Tubes with a flaw, in either the parent tube or the sleeve, within a sleeve-to-tube joint shall be plugged. 5.5.8.c.4 d. Tubes with a flaw in a repair roll shall be plugged. 5.5.8.d 4) 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. For tubes that have undergone repair rolling, the tube and tube roll, outboard of the new roll area in the tube sheet, can be excluded from inspections because it is no, longer part of the pressure boundary once the repair roll is installed. For tubes that have undergone sleeving repairs, the segment of the parent tube between the bottom of the upper-most sleeve roll and the top of the middle sleeve roll can be excluded from inspection because it is no longer part of the pressure boundary once the sleeve is installed. In addition to meeting the requirements of 4.a through 4.e below, the inspection scope, inspection methods and inspection intervals shall be such as to ensure that SQ tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations. 5.5.8.d.1 a. Inspect 100% of the tubes in each SQ during the first refueling outage following SG replacement. 5.5.8.d.2 b. Inspect 100% of the tubes at sequetitial 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 interval between r'efueling outages (whichever is less) without being inspected. DAVIS-BESSE, UNIT 1 6-11 Amendment No. 276 Page 7 of 20 Attachment 1, Volume 16, Rev. 0, Page 57 of 168

Attachment 1, Volume 16, Rev. 0, Page 58 of 168 ITS ITSS (U) ITS 5.5 0 6.0 ADMINISTRATIVE CONTROLS 6.8.4.g.4 (Continued) 5.5.8.d.3 c. 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 interval between refueling outages (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. ' 5.5.8.d.4 d. During each periodic SG tube inspection, inspect 100% of the tubes that have been repaired by the repair roll process. This special inspection shall be limited to the repair roll joint and the roll transitions of the repair roll. 5.5.8.d.5 e. Inspect peripheral tubes in the vicinity of the secured internal auxiliary feedwater header between the upper tube sheet and the 15th tube support plate during each periodic SG tube inspection. The tubes selected for inspection shall represent the entire circumference of the steam generator and shall total at least 150 peripheral tubes. 5.5.8.e 5) Provisions for monitoring operational primary to secondary leakage. 5.5.8. 6) Provisions for SG tube repair methods: Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below. 5.5.8.f. 1 a. Sleeving in accordance with TopiWl Report BAW-2120P. 5.5.8.f.2 b. Repair rolling in accordance with Topical Report BAW-2303P, Revision 4. The new roll area must be free of flaws in order for the repair to be considered acceptable. 5.5.8.g 7) Special visual inspections: Visual inspections of the secured internal auxiliary feedwater header, header to shroud attachment welds, and the external header thermal sleeves shall be performed on each SG through the auxiliary feedwater injection penetrations. These inspections shall be performed during the third period of each ten-year Inservice Inspection Interval (ISI). DAVIS-BESSE, UNIT 1 6-12 Amendment No. 276 Page 8 of 20 Attachment 1, Volume 16, Rev. 0, Page 58 of 168

Attachment 1, Volume 16, Rev. 0, Page 59 of 168 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS 5.5.1 6.15 OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.5.1.c Changes to the ODCM: 5.5.1.c.1 a. Shall be documented and records of reviews performed shall be retained las re gired by Ffthe USAR Chapter 17 OWity Assurance Progran*. This documentation shall contain: 5.5.1.c.l.a) I) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s), and 5.5.1.c.1.b) 2) 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. 5.5.1.c.2 b. Shall become effective after the approval of the plant manager. 5.5.1.c.3 c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as 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. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented." DAVIS-BESSE, UNIT I 6-20 Amendment No. S6,-1-70T84,23t,

                                                                                              -2O-2t7-2; 276 Page 9 of 20 Attachment 1, Volume 16, Rev. 0, Page 59 of 168

Attachment 1, Volume 16, Rev. 0, Page 60 of 168 ITS G IITS 5.5 6.0 ADMINISTRATIVE CONTROLS 6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM 5.5.15.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 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: 5.5.15.a.1 1) A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision I. 5.5.15.a.2 2) The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, As-found testing will not be required. 5.5.15.b b. The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 38 psig. 5.5.15.c c. The maximum allowable containment leakage rate, Ia, at Pa, shall be 0.50% of containment air weight per day. 5.5.15.d d. Leakage rate acceptance criteria are: 5.5.15.d.1 I) 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 < 0.75 La for Type A tests,< 0.60 La for all penetrations and valves subject to S IT Type B and Type C tes! and < 0.03 L. for all penetrations that are secondartSee TS.6. ontainment bypass leakage paths;3.6.3

12) A single penetration leakage rate of < 0.15 1, for each containment purge penetration; See ITS]

5.5.15.d.2 3) Air lock acceptance criteria are: 5.5.15.d.2.a) a) Overall air lock leakage rate is < 0.015 La when tested at > Pa, 5.5.15.d.2.b) b) For each door, seal leakage rate is < 0.01 La when the volume between the door seals is pressurized to> 10 psig.

e. The provisionsdf Specification 4.0.2 do n apply to the test frequ cies specified in the A Containmentt akae Rate Testing Pro - s 5.5.15.e f. The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

DAVIS-BESSE, UNIT 1 6-21 Amendment No. 24*, 276 Page 10 of 20 Attachment 1, Volume 16, Rev. 0, Page 60 of 168

Attachment 1, Volume 16, Rev. 0, Page 61 of 168 ITS E IITS 5.5 6.0 ADMINISTRATIVE CONTROLS 5.5.13 6.17 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM This program provides a means for processing changes to the Bases of these Technical Specifications. 5.5.13.a a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews. 5.5.13.b b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following: 5.5.13.b.1 1) A change in the TS incorporated in the license or 5.5.13.b.2 2) A change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59. 5.5.13.c c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR. 5.5.13.d d. Proposed changes that meet the criteria of 6.1T7.l and 6.17b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

                 ,4                            Add proposed ITS 5.5.5 and ITS 5.5.14                               M 1

[AdFdpropose d ITS 5.5 16 and ITS 5.5.17 }( DAVIS-BESSE, UNIT .1 6-22 Amendment No. 249; 276 Page 11 of 20 Attachment 1, Volume 16, Rev. 0, Page 61 of 168

Attachment 1, Volume 16, Rev. 0, Page 62 of 168 ITS 5.5 ITS DEFINITIONS 1.29 Deleted

          !.30 Deleted 1.31 Del-ted OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.5.1.a,   1.32 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology 5.5.1.b   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/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Ef*fluent Release Reports required by Speciffications 6.9. 1.10 and 6.9. 1.11.

1.33 Deleted 1.34 Deleted 1.35 Deleted 1.36 Deleted MEMBER(S) OF THE PUBLIC 1.37 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally See ITS Chapter 1.0J associated with the plant. This category does not include 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 recrea-tional, occupational or other purposes not associated with the plant. SITE B O UND.AJRY 1.39 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwi'se controlled by the licensee. [DAVIS-BESSE, UNIT I 1-6a Amendment No. 7 272 Page 12 of 20 Attachment 1, Volume 16, Rev. 0, Page 62 of 168

Attachment 1, Volume 16, Rev. 0, Page 63 of 168 ITS 5.5 ITS APFUCAEUII.,r. SURVEnIlANCE REQURBh~fETS 4.0.1 Surveillance Rcqulrens shall be applicable during the OPERATIONAL MODES or other conditlons specified for Individual UrnItIng Conditions for Operation unless otherwise stated in in individual Surveillance Require=L See ITS 3.0 4.0-2 Each Surveillance Requircment shall be performed within the specified time interval with a interval. maximum allowable exitcnsion not to exceed 25 percent of the specified surveillance 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Umildng Condition for Operation (LCO), except as noted below. If it is discovered that a Surveillance was not perfonxed within its specified frequency, then compliance with the requirement to declare the LC not met may be delayed, from the time of discovery, up to 24 hours or up to tbe limit of the specified frequcncy, whichever is greater. This delay period k peri*tled to allow perfoarmac of th Surveillance. A risk evaluation shall be performed for any Surveillance delaycd greater than 24 hcurs, and the risk Impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable AC-IONS must be entered. When the Surveillance is performed within' the delay period and the Surveillance is not met, the LCOO must immediatzly be declared not met, and the applicable ACTIONS must be entered. Surveillance requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATION.AL MODE or other specified applicability condition shall not be made unless .the Surveillance Requirement(s) associated Yith the Urmiting Condition for SOpt.ration have beer performed within the stated rtuveinlance.interval or as otherwise specified. 5.5.7 Indenffrvc Ttn. of ASMECodeClaad ou 3pumpsndalss be prfo Plants LA05 accordance with e ASME Code for Operati n and Maintenance of N 'clear Power (ASNtE OM c) and applicable Addenda reqUired by 10 CFR.5 ,Section SO.SSL DAVIS-BESSEZ UNIT I 3/40-2 Axwendmn~tNo. 71,140,145,197,250 ,254 Page 13 of 20 Attachment 1, Volume 16, Rev. 0, Page 63 of 168

Attachment 1, Volume 16, Rev. 0, Page 64 of 168 ITS 5.5 ITS APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 5.5.7a b. Surveillance intervals specified in ISection ,of the ASMvIEBoiler and Prg sure Vessel LA04 ICode and apqlkiAble Addenda and theJASMEE OM Code and applicable Addenda shalIb e applicable as follows in these Technical Specifications: ASMVEy]oiler and Pre~ure Vessel Required frequencies for [Code pgnd theJASME OM Code and performing inservice -{"... applicable Addenda terminology for linspeion and trsting inservice linspeion and testing activities criteria LAO4 lWeekly .ct once per 7 days LA04 Monthly At least once per 31 days ISemi-quarterly t once per 46 days [ Quarterly or every 3 months At least once per 92 days Sem-iannually or eve 6 months At. Jýronce per 184 days Every 9 months.. ,Weast once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days 5.5.7.b c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inseon and testing activities. LA04

d. Performance of the inservice nspe ion andtesting ac y5ve hall be in addition to A07 other specified rveillance Requirements.

5.5.7.d e. Nothing in the [ASME Boiler a n sdsure Vessel Code or the ASME OM Code shall be construed to supersede the requirements of any Technical Specification. Add proposed ITS 5.5.7.c } DAVIS-BESSE, UNIT 1 3/4 0-3 Amendment No. 250 Page 14 of 20 Attachment 1, Volume 16, Rev. 0, Page 64 of 168

Attachment 1, Volume 16, Rev. 0, Page 65 of 168 ITS 5.5 ITS REACTOR COOLANT SYSTEM 3.4.10 STRUCTUR.AL INTEGPJTY ASME CODE CLASS 1. 2 and 3_COMPONENTS MITING CONDITION FOR ,OPERATODN 3.4.10.1 The structural integrity of ASME Code Class 1. 2 and 3 components shall be maintained in accordance with Specification 4.4. 10. 1. APPLICABILTTY: All MODES. ACTION:

a. With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements. restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50 'F above the minimum temperature required by NDT See CTS considerations. 3/4.4.10.1

b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200 *F.

c. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the component(s) to within its limit or isolate the affected component(s) from ser.ice.

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REUIREMENTS 4.4.10.1 In addition to the requirements of Specification 4.0.5: 5.5.6 a. Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every 10 years. The inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or a surface examination of exposed surfaces of the disassembled flywheel. The recommendations delineated in Regulatory Guide 1. 14, Revision 1, August 1975, Positions 3, 4 and 5 of Section C.4b shall apply. DAVIS-BESSE, UNIT I 3/4 4-30 Amendment No. 232 Page 15 of 20 Attachment 1, Volume 16, Rev. 0, Page 65 of 168

Attachment 1, Volume 16, Rev. 0, Page 66 of 168 ITS 5.5 ITS RVEIACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 5.5.4.a b. Each internals vent valve shall be demonstrated OPERABLE at least once per 24 months* Iduring slutdownlby: A09 5.5.4. a.l1 I. Verifying through visual inspection that the valve body and valve disc exhibit no abnormal degradation, 5.5.4.a.2 2. Verifying the valve is not stuck in an open position, and 5.5.4.a.3 3. Verifying through manual actuation that the valve is fully open when a force of

                       < 400 lbs. is applied vertically upward.

___ýThe provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Al10 Vessel Internals Vent Valves Program Surveillance Frequencies.

                                         -    I An exception applies r the interval following the March 2003 ver cation completed during the Thirneei6 Refueling Outage. Under this exception, e next performance of               --      a this sttrveillanc/eequirement may be delayed until March 25/2006.

DAVIS-BESSE, UNIT I 3/4 4-31 Amendment No. 23. 95, 165, 268 Page 16 of 20 Attachment 1, Volume 16, Rev. 0, Page 66 of 168

Attachment 1, Volume 16, Rev. 0, Page 67 of 168 ITS 5.5 ITS ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continuedi I. Verifying the fuel level in the day fuel tank. [e3.8.1 C See ITS

2. Verifying the fuel level in the fuel storage tank. SeeITS

J3.8.3

3. Verifying the fuel transfer pump can be started and trans-fers fuel from the storage system to the day tank.

4. Verifying the diesel starts and accelerates up to 900 rpm, preceded by an engine prelube and/or appropriate other warmup procedures.

5. Verifying the generator is synchronized, loaded to >_ 1000 SeelTS kw, and operates for > 60 minutes. 3.8.1 J

6. Verifying the diesel generator is aligned to provide standby power to the associated essential busses.

7. Verifying that the automatic load sequence timer is OPERABLE with each load sequence time within +/- 10% of its required value. Add proposed ITS 5.5.11

b. At least once per 92 days by verifying that a sample ofthdesel generic program statement fuel from the fuel storage tank is within the acceptable limits @

5.5.12 specified in Table I of ASTM D975-68 when checked for viscosity,L0 water and sediment.e At least once per J84 days by: 7- (*SeeITS f3.8.3

11. Verifying the fuel level in the day fuel tank and SeelTS k.ITS 3.8.21)

12. Verifying the fuel lee nthe fulstorage tank. -3.1 levelin ful * "* See ITS1

3. Verifying the fuel transfer pump can be started and 3.8.3 transfers fuel from the storage system to the day tanK.

4. Verifying the diesel startsfrom ambientcondition and accelerates to at least 900 rpm in < 10 seconds.

5. Verifying the generator is synchronized, loaded to > 1000 kw, and operates for > 60 minutes. See IT

6. Verifying the diesel generator is aligned to provide standby power to the associated essential busses.

7. Verifying'that the automatic load sequence timer is OPERABLE with each load sequence time within +/- 10% of its required value.

DAVIS-BESSE, UNIT I 3/4 8-3 Amendment No. 75, 97, ,105,, 203 he provisions Iiesel of SR 3.0.2 Fuel Oil Testing and SPR Program 3.0.3 are applicable Surveillance to the_j Frequencies. Page 17 of 20 Attachment 1, Volume 16, Rev. 0, Page 67 of 168

Attachment 1, Volume 16, Rev. 0, Page 68 of 168 ITS UITS IT 5.5 RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKS* A13* L~~ODTO ~ ~ O P ~AIN/ ~I I N ~ _ ~~d proposed ITS 5.5.10 generic program statem~ent 5.5.11, . 3.11.1 The quantity of radioactive material contained in each of the 5.5.11.b following unprotected outdoor tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.

a. Outside temporary tank.

APPLICABILITY: At all times. ACTION: LA0

a. With the q antity of radioactive material in ny of the above listed ta s exceeding the above limit, imme ately suspend all additions of radioactive material to the tan and within 48 hours reduce t e tank contents to within the limi, and describe the event

ding to this condition in the next Radioactive Effluent Release Report.

b. The p visions of Specifications 3.0.3 an 3.0.4 are not appl f able.

SURVEILLANCE REQUIREMENTS 5.5.11.b 4.11.1 The quantity of radioactive material contained in each of the above listed.tanks shall be sample a representative of theto tank deteriined be within contents thelaiabove once by. leastlimit per L0 linalyzing the/tank.( 17 days when radioacti *e material's are being added to Gas

                       <              I The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Frequencies .          A1 and Storage Tank Radioactivity Monitoring Program Surveillance 5.5.11.b    *Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes.. or walls capable of holding the tank contents or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

DAVIS-BESSE, UNIT 1 3/4 11-1 Amendment No. 06, V7(J.184-Page 18 of 20 Attachment 1, Volume 16, Rev. 0, Page 68 of 168

Attachment 1, Volume 16, Rev. 0, Page 69 of 168 ITS 5.5 ITS RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE (Hydrogen rich systems not designed to withstand a hydrogen explosion) LIMITING CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement 5.5.11, 3.11.2 The concentration of oxygen in the waste gas system shall be limitedýol s than 5.5.11.a LA06 lor equal to 2% byvolume whenever the hyd7d,-,en concentration exceedsý4% b volume. APPLICABILITY: At I times. ACTION:

a. With the concentr tion of oxygen in the waste gas system great r than 2% by volume but less than or e uai to 4% by volume, reduce the oxygen con entration to the above LAO6 limits, within 48 ours.

b. With the concen ration of oxygen in the waste gas system gr ater than 4% by volume and the hydrog concentration greater than 4% by volume, immediately suspend all additions of w te gases io the system and reduce the conce tration of oxygen to less than or equal t 2% by volume without delay.

c. The provision of Specifications 3.0.3 and 3.0.4 are not ap licable.

SURVEILLANCE REQUIREMENTS 5.5.11.a 4.11.2 The concentrations of oxygen in the waste gas system shall be determined to be within the above limits by monitoring the waste gases in the waste gas system, 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.

          .1 DAVIS-BESSE, UNIT I                                   3/4 11-2                        Amendment No. 86, 170, 234 Page 19 of 20 Attachment 1, Volume 16, Rev. 0, Page 69 of 168

Attachment 1, Volume 16, Rev. 0, Page 70 of 168 ITS 5.5 ITS

                                   -6    -

2.C(4) Fire Protection FENOC shall implement and maintain in effect all provisions Not partof ITS of the approved Fire Protection Program as described in the conversion Updated Safety Analysis Report and as approved in the SERs dated July 26, 1979, and May 30, 1991, subject to the foilowing provision: FENOC may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. 5.5.9 (5) FENOC shall maintain in effect and implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. The program shall include: 5.5.9.a (a) Identification of a sampling schedule for the critical parameters and control points for these parameters; (b) Identification of the procedures used to quantify 5.5.9.b parameters that are critical to control points; 5.5.9.c (c) identification of process sampling points; 5.5.9.d (d) Procedure for the recording and management of data; 5.5.9.e (e) Procedures defining corrective actions for off control point chemistry conditions; and 5.5.9.f (f) A procedure identifying the authority responsible for the interpretation of the data, and the sequence and timing of administrative events required to initiate corrective action. (6) Antitrust Conditions patof ITS1 Nmot FENOC and FirstEnergy Nuclear Generation Corp. shall comply _ conversionI 2.E with the antitrust conditions delineated in Condition of this license as if named therein. FENOC shall not market or broker power or energy from the Davis-Besse Nuclear Power Station, Unit No. 1. FirstEnergy Nuclear Generation Corp. is responsible and accountable for the actions of FSNOC to the extent that said actions affect the marketing or brokering of power or energy from the Davis-Besse Nuclear Power Station, Unit No. 1, and in any way, contravene the antitrust license conditions contained in the license. L-6 Amendment No. iA,W,6 ý,.,270 Page 20 of 20 Attachment 1, Volume 16, Rev. 0, Page 70 of 168

Attachment 1, Volume 16, Rev. 0, Page 71 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 6.8.4.a specifies the requirements for the Primary Coolant Sources Outside Containment Program, however there is no statement as to whether or not the provisions of CTS 4.0.2 are applicable. CTS 6.8.4.d specifies the requirements for the Radioactive Effluent Controls Program, however there is no statement as to whether or not the provisions of CTS 4.0.2 and CTS 4.0.3 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 Frequency. ITS 5.5.3 states that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program Surveillance Frequencies. This changes the CTS by adding the allowances of ITS SR 3.0.2 to the Primary Coolant Sources Outside Containment Program and the allowances of ITS SR 3.0.2 and SR 3.0.3 to the Radioactive Effluent Controls Program. These statements are needed to maintain allowances for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.2 and SR 3.0.3 are 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. A03 CTS 6.8.4.f provides the requirements for the Ventilation Filter Testing Program. The Program uses the nomenclature "Shield Building Emergency Ventilation System." However, CTS 3.6.5.1 uses the nomenclature "emergency ventilation system" for the same ventilation system. ITS 5.5.10 uses the nomenclature "Station Emergency Ventilation System." This changes the CTS by using a common nomenclature for the Station Emergency Ventilation System. This change is acceptable since it is only providing a common nomenclature for this ventilation filter system. Both ITS 3.7.12 and ITS 5.5.10 use the nomenclature "Station Emergency Ventilation System" for this system. This change is designated administrative since it does not result in any technical changes to the CTS. A04 CTS 6.8.4.f footnote

states the periodic testing for the Shield Building Emergency Ventilation System and the Control Room Emergency Ventilation System are performed once each REFUELING INTERVAL. The need for testing following painting, a fire, or a chemical release in any ventilation zone communicating with the Shield Building Emergency Ventilation System or Control Room Emergency Ventilation System is as specified in the VFTP. The method Davis-Besse Page 1 of 9 Attachment 1, Volume 16, Rev. 0, Page 71 of 168

Attachment 1, Volume 16, Rev. 0, Page 72 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS for testing is based on Regulatory Guide 1.52, Revision 2, except for charcoal laboratory testing which will be performed in accordance with ASTM D 3803-1989. ITS 5.5.10 does not contain this footnote. This changes the CTS by deleting this footnote. This change is acceptable because no changes have been made to the existing requirements. The CTS 6.8.4.f footnote

restates what is already stated in the Ventilation Filter Testing Program description. This change is designated as administrative because it does not result in technical changes to the CTS.

A05 CTS 6.16, 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 1.1.63, dated September 1995. CTS 6.16.e states that the provisions of Specification 4.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Program. ITS 5.5.15 does not include this provision. This changes the CTS by deleting the statement that the provisions of Specification 4.0.2 are not applicable. This change is acceptable because no changes have been made to the existing requirements. The statement associated with CTS 4.0.2 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. A06 ITS 5.5.16 provides the requirements for the Battery Monitoring and Maintenance Program. ITS 5.5.17 provides the requirements for the Control Room Envelope Habitability Program. The CTS does not include these two programs. This changes the CTS by including these two new programs. ITS 5.5.16 has been added due to changes described in ITS 3.8.6, Discussion of Changes (DOC) L01, L02, L03, L07, and L08. ITS 5.5.17 has been added due to changes described in ITS 3.7.10, DOC L01. As such, the addition of these two programs in ITS 5.5 is acceptable and are designated as administrative because they do not result in technical changes to the CTS not already described in other ITS DOCs. A07 CTS 4.0.5.d states that the performance of the above testing activities shall be in addition to other specified Surveillance Requirements. ITS 5.5.7 does not include a similar statement. This changes the CTS by deleting the statement. CTS 4.0.5.d restates that all applicable requirements must be met. Repeating this overall requirement as a specific detail is redundant and unnecessary. Therefore, this detail can be omitted without any technical change in the requirements and is acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A08 CTS 4.0.5 specifies the requirements for the Inservice Testing Program, however there is no statement whether the provisions of CTS 4.0.3 are applicable. ITS 5.5.7.c states that the provisions of SR 3.0.3 are applicable to the inservice Davis-Besse Page 2 of 9 Attachment 1, Volume 16, Rev. 0, Page 72 of 168

Attachment 1, Volume 16, Rev. 0, Page 73 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS testing activities. This changes the CTS by adding the allowances of ITS SR 3.0.3 to the Technical Specification Inservice Testing Program requirements. This statement is needed to maintain allowances for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.3 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 a technical change to the CTS. A09 CTS 4.4.10.1 .b requires reactor vessel internals vent valves to be tested every 24 months "during shutdown." ITS 5.5.4 requires similar testing every 24 months, but does not include the "during shutdown" requirement. This changes the CTS by deleting the "during shutdown" testing requirement. This change is acceptable since the reactor vessel internals vent valves can only be tested when the unit is shutdown and the reactor vessel head removed. Therefore, stating that the unit must be shutdown is redundant and unnecessary. This change is designated as administrative because it does not result in any technical changes to the CTS. A10 The internal vent valves requirements in CTS 4.4.10.1 .b have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.4). As such a general program statement of applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify the allowances for Surveillance Frequency extensions do apply. This changes the CTS by specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3. The addition of ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. All CTS 4.4.10.b.1 requires reactor vessel internals vent valves to be tested every 24 months. CTS 4.4.10.1.b is modified by footnote *, which states that an exception applies for the interval following March 2003 verification completed during the Thirteenth Refueling Outage. Under this exception, the next performance of the surveillance requirement may be delayed until March 25, 2006. ITS 5.5.4 does not contain this footnote. This changes the CTS by deleting the footnote. CTS 4.4.10.b footnote

is an exception for the Thirteenth Refueling Outage.

Since this refueling outage has been completed and the test has been performed by March 25, 2006 as required by the footnote, there is no need to maintain the footnote. This change is designated as administrative because it does not result in technical changes to the CTS. A12 The Surveillance associated with diesel fuel oil testing (CTS 4.8.1.1.2.b) has been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.12). As such, a general program statement has been added as Davis-Besse Page 3 of 9 Attachment 1, Volume 16, Rev. 0, Page 73 of 168

Attachment 1, Volume 16, Rev. 0, Page 74 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ITS 5.5.12. 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 extension do apply. This changes the CTS by moving the diesel fuel oil testing Surveillance 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 Surveillance. The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A13 The liquid holdup tank requirements in CTS 3/4.11.1 and the explosive gas mixture requirements of CTS 3/4.11.2 have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.11). As such, a general program statement has been added. Also, a statement of applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify the allowances for Surveillance Frequency extensions do apply. This changes the CTS by moving liquid holdup tank requirements and the explosive gas mixture requirements to a program in ITS 5.5.11 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 Specification. The addition of ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M01 The CTS does not include program requirements for an Allowable Operating Transient Cycles Program or a Safety Function Determination Program. The ITS includes programs for these activities. This changes the CTS by adding the Allowable Operating Transient Cycles Program and Safety Function Determination Program (SFDP). The Allowable Operating Transient Cycles Program provides controls to track the UFSAR Section 5, cyclic and transient occurrences to ensure that components are maintained within the design limits. 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 these programs are found in ITS 5.5.5 and ITS 5.5.14. 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. M02 CTS 4.8.1.1.2.b requires verifying every 92 days that a sample of diesel fuel from the fuel oil storage tank is within the acceptable limits specified in Table 1 of Davis-Besse Page 4 of 9 Attachment 1, Volume 16, Rev. 0, Page 74 of 168

Attachment 1, Volume 16, Rev. 0, Page 75 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ASTM D975-68 when checked for viscosity, water, and sediment. In addition, no testing is currently required on new fuel oil prior to addition to the fuel oil storage tank. ITS 5.5.12.a restricts the acceptability of new fuel oil for use prior to addition to storage tanks by requiring the determination that the fuel oil has an API gravity or a absolute specific gravity within limits, a flash point and kinematic viscosity within limits, and either a clean and bright appearance with proper color or a water and sediment content within limits. ITS 5.5.12.b requires all other properties of new fuel to be verified within 31 days following addition of the new fuel oil to the storage tank. ITS 5.5.12.c requires the total particulate concentration of the stored fuel oil to be -<10 mg/I when tested every 31 days. This changes the CTS by providing restrictions on the acceptability of new fuel oil prior to addition to the fuel oil storage tank and after addition to the fuel oil storage tank, and providing a requirement that the total particulate concentration of the stored fuel oil be < 10 mg/I when tested every 31 days. The purpose of ITS 5.5.12.a and ITS 5.5.12.b are to ensure that only high quality fuel oil is added to the fuel oil storage tank. The purpose of ITS 5.5.12.c is to ensure that the quality of the stored fuel oil is satisfactory for long term operation of the EDGs. The change is acceptable because the proposed Surveillances are sufficient to ensure high quality fuel oil is placed and maintained in the storage tank. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or Other TS Requirement to the TRM, UFSAR, ODCM, QAPM, IST Program,or liP) CTS 6.8.4.b, "In-Plant Radiation Monitoring," describes a program to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. ITS 5.5 does not include this program. This changes the CTS by moving the requirements for the In-Plant Radiation Monitoring Program to the Technical Requirements Manual (TRM). The removal of this requirement from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The CTS 6.8.4.b program is designed to minimize radiation exposure to plant personnel in vital areas of the plant after an accident, and has no impact on nuclear safety or the health and safety of the public. This change is acceptable because the program requirements will be adequately controlled in the TRM. The TRM is currently incorporated by reference into the UFSAR, thus any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications. Davis-Besse Page 5 of 9 Attachment 1, Volume 16, Rev. 0, Page 75 of 168

Attachment 1, Volume 16, Rev. 0, Page 76 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS LA02 (Type 4 - Removal of LCO, SR, or Other TS Requirement to the TRM, UFSAR, ODCM, QAPM, IST Program, or liP) CTS 6.8.4.e, "Radiological Environmental Monitoring Program," describes a program to monitor the radiation and radionuclides in the environs of the plant. ITS 5.5 does not include this program. This changes the CTS by moving the requirements for the Radiological Environmental Monitoring Program to the Offsite Dose Calculation Manual (ODCM). The purpose of CTS 6.8.4.e is to provide representative measurements of radioactivity in the highest potential exposure pathways, and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The removal of the requirement for this program 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 5.6.1 still requires an annual report of the results of the "Radiological Environmental Monitoring Program." Also, this change is acceptable because these types of procedural details will be adequately controlled in the ODCM. Changes to the ODCM are controlled by the ODCM change control process in ITS 5.5.1, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirement change because the requirements for a program are being removed from the Technical Specifications. LA03 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 6.15.1.a requires changes to the ODCM to be documented and records of reviews performed to be retained as required by the UFSAR Chapter 17 Quality Assurance Program. ITS 5.5.1.c.1 requires changes to the ODCM to be documented and records of reviews performed to be retained. This changes the CTS by moving the record retention requirement reference to the Quality Assurance Program Manual (QAPM). 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. Also, this change is acceptable because these types of procedural details will be adequately controlled in the QAPM. Any changes to the QAPM 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 procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA04 (Type 4 - Removal of LCO, SR, or other TS requirementto the TRM, UFSAR, ODCM, QAPD, or liP) CTS 4.0.5 provides requirements for the Inservice Inspection Program. The ITS does not include Inservice Inspection Program requirements. In addition, since the Inservice Testing Program is the only requirement remaining, the reference to ASME Code Class 1, 2, and 3 "components" has been changed to "pumps and valves" for clarity. Pumps and valves are the only components related to the Inservice Testing Program (as Davis-Besse Page 6 of 9 Attachment 1, Volume 16, Rev. 0, Page 76 of 168

Attachment 1, Volume 16, Rev. 0, Page 77 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS described in CTS 4.0.5.a). This changes the CTS by moving these requirements from the Technical Specifications to the Inservice Inspection Program (liP). 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 lIP, which is required by 10 CFR 50.55a. Compliance with 10 CFR 50.55a is required by the Davis-Besse Operating License. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications. LA05 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.0.5.a specifies that the Inservice Testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50, Section 50.55a. ITS 5.5.7 states that the Inservice Testing Program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. This changes the CTS by moving these procedural details from the Technical Specifications to the Inservice Testing Program. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements for the control for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. Also, this change is acceptable because these types of details will be adequately controlled in the plant controlled Inservice Testing Program. Changes to the Inservice Testing Program will be controlled by the provisions of 10 CFR 50.55a. This change is designated as a less restrictive removal of detail change because the details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA06 (Type 3 - Removing ProceduralDetail for Meeting TS Requirements or Reporting Requirements) CTS 3/4.11.1 includes the details for implementing the requirements for the liquid holdup tanks. CTS 3/4.11.2 includes the details for implementing the requirements for the explosive gas mixture. The details for implementing these requirements, including the specific limits for the explosive gas mixture, are not included in the ITS. The ITS only includes a requirement to maintain a program for these requirements. This changes the CTS by moving these procedural details for implementing the requirements, including the specific limits, from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details for the specific explosive gas limits, Applicability, Actions, and Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Davis-Besse Page 7 of 9 Attachment 1, Volume 16, Rev. 0, Page 77 of 168

Attachment 1, Volume 16, Rev. 0, Page 78 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.11 still retains the requirement to include a program, which provides controls for potentially explosive gas mixtures contained in the Waste Gas System and the quantity of radioactivity contained in outdoor temporary liquid storage tanks. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. The TRM is currently incorporated by reference into the UFSAR, thus any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L01 (Category 5 - Deletion of Surveillance Requirement) CTS 4.8.1.1.2.b requires verifying every 92 days that a sample of diesel fuel from the fuel oil storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water, and sediment. ITS 5.5.12.c only requires total particulate concentration of the stored fuel oil to be tested every 31 days. This changes the CTS by deleting the quarterly viscosity, water, and sediment checks of stored fuel oil. The purpose of CTS 4.8.1.1.2.b is to ensure that the quality of the stored diesel fuel oil is acceptable so that the emergency diesel generators can perform their safety function. This change is acceptable because the new Surveillance Requirements (added as described in Discussion of Change M02) provide an acceptable level of equipment reliability. ITS 5.5.12.a restricts the acceptance of new fuel oil for use prior to addition to storage tank until the determination that the fuel oil has an API gravity or an absolute specific gravity within limits, a flash point and kinematic viscosity within limits, and either a clear and bright appearance with proper color or a water and sediment content within limits. ITS 5.5.12.b requires all other properties of new fuel to be verified within 31 days following addition of the new fuel oil to the storage tank. ITS 5.5.12.a and ITS 5.5.12.b will ensure that the new fuel oil is of high quality. Fuel oil degradation during long term storage shows up as an increase in particulate, mostly due to oxidation. Therefore, total particulate concentration of the fuel oil is determined and compared to an acceptable limit every 31 days as required by ITS 5.5.12.c. The presence of particulate does not mean that the fuel oil will not burn properly in a diesel engine but the particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure. This test is required to be performed every 31 days since fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between the 31 day Frequency interval. In addition, ITS SR 3.8.3.5 has been added (see Discussion of Change M04 for ITS 3.8.3) to ensure that microbiological fouling does not occur. Microbiological fouling is also a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. The new Surveillance has been added to ensure the removal of water from the fuel storage tank once every 31 days to eliminate the necessary environment for Davis-Besse Page 8 of 9 Attachment 1, Volume 16, Rev. 0, Page 78 of 168

Attachment 1, Volume 16, Rev. 0, Page 79 of 168 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS bacterial survival. This change is designated as less restrictive because a Surveillance required in the CTS will not be required in the ITS. Davis-Besse Page 9 of 9 Attachment 1, Volume 16, Rev. 0, Page 79 of 168

Attachment 1, Volume 16, Rev. 0, Page 80 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 80 of 168

Attachment 1, Volume 16, Rev. 0, Page 81 of 168 CTS Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 6.8.4 The following programs shall be established, -implemented, and maintained. Definition 1.32 5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring progra rk4a))

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 R5.6.11 and Specification 05.6.2g.

6.15 Licensee initiated changes to the ODCM: 6.15.a - . Shall be documented and records of reviews performed shall be retained. This documentation shall contain: 6.15.a.1) jj--. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s)and.i - - 6.15.a.2) - A determination effluent that the change(s) maintain the levels of radioactive 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 calculationsnjQ 6.15.b f2 Shall become effective after the approval of the plant managerM [ 0 6.15.c - Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented. BWOG STS 5.5-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 81 of 168

Attachment 1, Volume 16, Rev. 0, Page 82 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6.8.4.a 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems include NSrT 1 *[Low Pressuro injection, Reactor Building'Spray, Makeup and/Purificatioad

                     '   I~Hydrogen R/ecombinerql The program shall include the following-                                   0 6.8.4.a(i)                 a. Preventive maintenance and periodic visual inspection requirementsand 0

6.8.4.a(ii) b. Integrated leak test requirements for each system at least once per 0 months. The provisions of SR 3.0.2 are applicable. 5.5.3 Post Acciden Sampling

                                        -       .------.-----REVIEW ER'S NOTE ------ ---- ..................-------- --    -

This progr m may be eliminated based on the implemen ation of BAW-2387, "Justificati n for the Elimination of the Post Accident Sa pling System From the Licensing Bases of Babcock and Wilcox-Designed Pla s," and the associated NRC Saf ty Evaluation. This p gram provides controls that ensure the capa ility to obtain and analyze react r coolant, radioactive gases, and particulates i plant gaseous effluents 0 and ontainment atmosphere samples under accide t conditions. The program shal include the following:-

a. Training of personnel, b Procedures for sampling and analysis, and Provisions for maintenance of sampling a d analysis equipment.]

6.8.4.d 5.5.~ Radioactive Effluent Controls Program 0 This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: 6.8.4.d. 1) a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM Jo 0 BWAOG STS 5.5-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 82 of 168

Attachment 1, Volume 16, Rev. 0, Page 83 of 168 5.5 CTS O INSERT 1 makeup, letdown, seal injection, seal return, low pressure injection, containment spray, high pressure injection, waste gas, primary sampling, and reactor coolant drain systems Insert Page 5.5-2 Attachment 1, Volume 16, Rev. 0, Page 83 of 168

Attachment 1, Volume 16, Rev. 0, Page 84 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals Radioactive Effluent Controls Proqram (continued) 0 6.8.4.d.2) b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten es the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20t.i-2024..1e 0 6.8.4.d.3) c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCMIV J-0 0 6.8.4.d.4) d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released to 10 CFR 50, Appendix I and projected from each unit to unrestricted areas, conforming 6.8.4.d.5) e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days --- 0 0 6.8.4.d.6) f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix 0 6.8.4.d.7) g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas ;I-r~beyond the site boundary shall be in accordance witand uoaelying:r 0 eFor noble gases: a ose rate fo 500 earehnyr to the who] bodyand a Appendix Bi Table dose b2, rate o rem yr to the For io~dinje-l 1, iodine-133, tritium, and skiCr. and all radionuclide./in particulateJ 0 for wit h lf-i~ives greater thn8ays: a dsrae<_ 500 mrernyrI to any or a/n

h. Limitations on the annual and quarterly air doses resulting from noble 6.8.4. d.8) gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I,*'

0 6.8.4. d.9) i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I and

                                                                                                                    "=-0 BW\OG STS                                           5.5-3                                Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 84 of 168

Attachment 1, Volume 16, Rev. 0, Page 85 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals Radioactive Effluent Controls Program (continued) 0 6.8.4.d.10) j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program lurveillance Irequencp--j] 5.5.5 lCompohent Cycliq'or Transj6nt Limiti-* llwbje Opera~ting Transient Cycles Program This program provides controls to tra*k the FSAR, Section ,yclic and transient occurrences to ensure that components are maintained within the 00 design limits. 5.5.6 [ Pre-Strdssed Concrete Containment Tendon Surveillan e Pro ram This pro ram provides controls for monitoring any tend n degradation in pre-stresse concrete containments, including effectivenes of its corrosion prote on medium, to ensure containment structural i egrity. The program shall includ baseline measurements prior to initial operati ns. The Tendon Surve lance Program, inspection frequencies, and a ceptance criteria shall be in 0 acco dance with Section XI, Subsection IWL of the SME Boiler and Pressure Ves el Code and applicable addenda as required b 10 CFR50.55a, except whe e an alternative, exemption, or relief has been authorized by the NRC. T/ r Th provisions of SR 3.0.3 are applicable to the T ndon Surveillance Program in pection frequencies. ] 4 .4.1 0 .1.aV Reactor Coolant Pumo Flywheel Insoection Proaram 0 This program shall provide for the inspection of each reactor coolant pump Inse ervi e recommendation oframgulatory position ý4. of Regulatory I~ e1.1A, Revision 1, August 19/'.* I*E*- Inservice Testing Progqram 4.0.5 _ý-_5? 17 LJ This program provides controls for inservice testing of ASME Code Class 1, 2, 0 4.0.5 LW pumps l a.

                                 **eproram                        shall clude theoowing:

Testing frequencies applicable to the ASME Code for Operations and 0 4.0.5.b Maintenance of Nuclear Power Plants (ASME OM Codes) and applicable Addenda as follows: BWOG STS 5.5-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 85 of 168

Attachment 1, Volume 16, Rev. 0, Page 86 of 168 5.5 CTS O* INSERT 2 4.4.10.1.b 5.5.4 Reactor Vessel Internals Vent Valves Program A program shall be established to implement the testing of the reactor vessel internals vent valves every 24 months as follows: 4.4.10.1.b.1 a. Verify by visual inspection that the valve body and valve disc exhibit no abnormal degradation; 4.4.10.1.b.2 b. Verify the valve is not stuck in an open position; and 4.4.10.1.b.3 c. Verify by manual actuation that the valve is fully open when a force of

                      < 400 lbs is applied vertically upward.

DOC Al1 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Vessel Internals Vent Valves Program test Frequencies. INSERT 3 Inservice inspection of each reactor coolant pump flywheel shall be performed every 10 years. The inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or a surface examination of exposed surfaces of the disassembled flywheel. The recommendations delineated in Regulatory Positions C.4.b(3), (4), and (5) of Regulatory Guide 1.14, Revision 1, August 1975, shall apply. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. Insert Page 5.5-4 Attachment 1, Volume 16, Rev. 0, Page 86 of 168

Attachment 1, Volume 16, Rev. 0, Page 87 of 168 CTS Programs and Manuals 5:5 5.5 Programs and Manuals 7ý5.5 Inservice Testing Program (continued) 0 4.0.5.b ASME OM Code and applicable Required Frequencies for Addenda terminology for inservice performing inservice testing testing activities activities lWeekly Monthly At lea nce per 7 days At least once per 31 days 0 Quarterly or every 3 months At least once per 92 days Semiannually orery 6 months Every 9 montJ;,1 At least o9s'e per 184 days At lea once per 276 days 0 Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days as2years TT 4.0.5.c b. The provisions of SR 3.0.2 are applicable to the above required less -497, Frequencies and other normal and accelerated Frequencies specifie in the Inservice Testing Program for performing inservice testing activities* (r) DCO A09 c. The provisions of SR 3.0.3 are applicable to inservice testing activities ad 4.0.5.e d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. 6.8.4.g V5.5 Steam Generator (SG) Program 0 A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions: 6.8.4.g.1)

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging~or repairlof 0 tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, pluggedor repaired*]to 0 confirm that the performance criteria are being met.

BWOG STS 5.5-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 87 of 168

Attachment 1, Volume 16, Rev. 0, Page 88 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 6 Steam Generator (SG) Program (continued) 6.8.4.g.2) b. :Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE. 6.8.4.g.2)a 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. 6.8.4.g.2)b 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 P1 gpmO Pier SGI[, ex-cept for specific types of/degradatio-n at7/ LCO 3.4.13, "RCS Operational LEAKAGE." 6.8.4.g.3) c. Provisions for SG tube repair criteriaTervice inspectionTeo 6.8.4.g.3)a to contain flawsfwith a depth equal to or exceeding @40%Jof the nominal tube wall thickness shall be plugged Mr repairedý,

                          'in a region of the tube that contains no repair.

OINSERT 53A BWAOG STS 5.5-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 88 of 168

Attachment 1, Volume 16, Rev. 0, Page 89 of 168 5.5 CTS INSERT 3A 6.8.4.g.3)b 2. Sleeves found by inservice inspection to contain flaws, in a region of the sleeve that contains no sleeve joint, with a depth equal to or exceeding 40% of the nominal sleeve wall thickness shall be plugged; 6.8.4.g.3)c 3. Tubes with a flaw, in either parent tube or the sleeve, within a sleeve to tube joint shall be plugged; and 6.8.4.g.3)d 4. Tubes with a flaw in a repair roll shall be plugged. Insert Page 5.5-6 Attachment 1, Volume 16, Rev. 0, Page 89 of 168

Attachment 1, Volume 16, Rev. 0, Page 90 of 168 5.5 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5 r SteamaGenerator (SG) Program (continued) 0

                          -------------------- -- ----------- REVIEWE  V    'S NOTE----------

Alternate tube re air.criteria currently perr itted by plant technii al specifications are listed here. he description of these/alternate tube repair riteria should be equivalent to th descriptions in current technical specificatio and should also 0 include any allo ed accident induced lI_ kage rates for specifc types of degradation at pecific1locations asaoc'ted with tube repair riteria. [The following alornate tube repair crite ia may be applied a an alternative to the 40% depth I'ased criteria: 0 6.8.4.g.4) 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 INSERT 4 through d.5

                             ]part       of the tube.4 In addition to meeting the requirements of d.1d i below, the inspection scope, inspection methods, and inspection 0

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.

                                                  - - - - S--- -R    V E WE R/S NO TE------

Plants are to in uude the appropriate FTrequency (e.g., sele t the appropriate Ite 2.) for their SG desi n. The first Item 2 i applicable to SGs with Alloy 00 mill annealed tub'h g. The second Ite 2 is applicable 0 to SGs with Al oy 600 thermally treat d tubing. The third Item 2 is applicable to Gs with Alloy 690 the mally treated tubin 6.8.4.g.4)a 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. 6.8.4.g.4)b R2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to 0 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] 0 BWOG STS 5.5-7 Rev. 3.1, 12/01105 Insert Page 5.5-6 Attachment 1, Volume 16, Rev. 0, Page 90 of 168

Attachment 1, Volume 16, Rev. 0, Page 91 of 168 5.5 CTS INSERT 4 6.8.4.g.4) For tubes that have undergone repair rolling, the tube and tube roll, outboard of the new roll area in the tube sheet, can be excluded from inspections because it is no longer part of the pressure boundary once the repair roll is installed. For tubes that have undergone sleeving repairs, the segment of the parent tube between the upper-most sleeve roll and the top of the middle sleeve roll can be excluded from inspection because it is no longer part of the pressure boundary once the sleeve is installed. Insert Page 5.5-7 Attachment 1, Volume 16, Rev. 0, Page 91 of 168

Attachment 1, Volume 16, Rev. 0, Page 92 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.54 Steam Generator (SG) Program (continued) 0 [2. Inspect 10 A% of the tubes at seq ential periods of 12 , 90, and, thereafter, 0 effective full power months. The first s quential period shall be co sidered to begin afte the first inservice i spection of the SGs. In a dition, inspect 50% of the tubes by the ref eling outage nearest th midpoint of the perio and the remaining 50% by the refueling o tage nearest the end of the period. No shall operate for more tt an 48 effective full po r months or two r fueling outages (whicheve is less) without bein inspected.] 00@ [2. Inspect 1 0% of the tubes at se uential periods of 144, 108, 72, and, thereafte , 60 effective full pow r months. The first equential period shall be nsidered to begin aft r the first inservice nspection of the SGs. In ddition, inspect 50% f the tubes by the r fueling outage nearest e midpoint of the per od and the remainin 50% by the refueling outage nearest the e d of the period. No SG shall operate for more than 72 effective full ower months or thr e refueling outages (whichever is less) wi hout being inspecte .1 6.8.4.g.4)c 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. 0 6.8.4.g.5) e. Provisions for monitoring operational primary to secondary LEAKAGE. 6.8.4.g.6) jJf. Provisions for SG tube repair methods. Steam generator tube repair 0) 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.

                               ---------                       REVlEWE 'S NOTE-------

Tube repair met ds currently permitted y plant technical spe ifications are to be listed here. e description of these %uberepair methods should be 0 equivalent to th descriptions in current technical specificatio rs. Ifthere are no approved tube epair methods, this section should not be us.

1. . . .1 BWOG STS 5.5-8 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 92 of 168

Attachment 1, Volume 16, Rev. 0, Page 93 of 168 5.5 CTS S INSERT 4A 6.8.4.g.4)d 4. During each periodic SG tube inspection, inspect 100% of the tubes that have been repaired by the repair roll process. This special inspection shall be limited to the repair roll joint and the roll transitions of the roll repair. 6.8.4.g.4)e 5. Inspect peripheral tubes in the vicinity of the secured internal auxiliary feedwater header between the upper tube sheet and the 15th tube support plate during each periodic SG tube inspection. The tubes selected for inspection shall represent the entire circumference of the the steam generator and shall total at least 150 peripheral tubes. INSERT 5 6.8.4.g.6)a 1. Sleeving in accordance with Topical Report BAW-2120P. 6.8.4.g.6)b 2. Repair rolling in accordance with Topical Report BAW-2303P, Revision 4. The new roll area must be free of flaws in order for the repair to be considered acceptable. 6.8.4.g.7) g. Special visual inspections: Visual inspections of the secured internal auxiliary feedwater header, header to shroud attachment welds, and the external header thermal sleeves shall be performed on each SG through the auxiliary feedwater injection penetrations. These inspections shall be performed during the third period of each 10 year Inservice Inspection Interval (ISI). Insert Page 5.5-8 Attachment 1, Volume 16, Rev. 0, Page 93 of 168

Attachment 1, Volume 16, Rev. 0, Page 94 of 168 5.5 CTS Programs and Manuals 5.5 5.5 Programs and Manuals License 5.5.Dg Secondary Water Chemistry Program 0 Condition 2.C(5) This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradationjand Iovypressure turbine dig,stress corrosionl

                                        ý      . Te program shall include:

License a. Identification of a sampling schedule for the critical variables and control Condition 2.C(5)(a) points for these variables., 0 License Condition 2.C(5)(b) b. Identification of the procedures used to measure the values of the critical variables L0 License Condition 2.C(5)(c) c. Identification of process sampling point, vich shall include monito)'ng the License Idischarge of the condensate pumps for egidence of condenser in Iakag , 0 Condition 2.C(5)(d) License

d. Procedures for the recording and management of dataLD1 0 Condition 2.C(5)(e) e. Procedures defining corrective actions for all off control point chemistry conditions, d 0 License f. A procedure identifying the authority responsible for the interpretation of the Condition 2.C(5)(f) data and the sequence and timing of administrative events, which is required to initiate corrective action.

6.8.4.f 5ý.5. Ventilation Filter Testing Program (VFTP) 0 A program shall be established to implement the following required testing of (safety related Engine red Safety Fe ure (ESF filter ventilation systemslat t e frequoncies] jspecified in [Regulatory/Guide], and in accordance withýRegulatory Guide 1.52, Revision 2,4ASME N510-P198*), and 1AG-1]r- ý198o, ASTM D ý3803-1ý989

                                                                                                 ýand

[ ANs,,/ 6.8.4.f.1) lsafety related a. Demonstrate for each of ther-Eý systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system E bypass< 0 5] Yowhen tested in accordance withf*Regulatory Guide 1.52, Revision 2, and SME N510- at the system flowrate specified below

                                             ;     %.9 ESF Ve7ion System                    Flowrate INSERT 6 I[ I r

BWOG STS 5.5-9 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 94 of 168

Attachment 1, Volume 16, Rev. 0, Page 95 of 168 5.5 CTS 0 INSERT 6 Safety Related Ventilation System Flowrate (cfm) 6.8.4.f.1) Station Emergency Ventilation System (EVS) > 7200 and < 8800 Control Room Emergency Ventilation System (CREVS) > 2970 and < 3630 Insert 5.5-9 Attachment 1, Volume 16, Rev. 0, Page 95 of 168

Attachment 1, Volume 16, Rev. 0, Page 96 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5. Ventilation Filter Testing Program (continued)

b. Demonstrate for each of the s related systems that an inplace test of th 1.0 0

6.8.4.f.2) charcoal adsorber shows a penetration and system bypass < F[0_5]1% when __ tested in accordance with ýRegulatory Guide 1.52, Revision 2, and A ASME N510-9]at the system flowrate specified below[ . ESF Venltaeion System dFlowrate Ssafety re 6.8.4.f.3) c. Demonstrate for each of the"M systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described inrRegulatory Guide 1.52, Revision 2S 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 humidit specified below.

                                                                       -                            7..    .  .            (RH)

J

                                                                -  REVIEVEl 'S NOTE------------- -- ----

The use of any st ndard other than ASTI D3803-1989 to test e charcoal sample may resul in an overestimation oq the capability of the harcoal to adsorb radioiodine. As a result, the ability of the harcoal filters to pe rm in a manner consistent with th licensing basis for the facility is indetermina e. ASTM D 3803-1 89 is a more stringent t sting standard beca e it does not differentiate be en used and new char al, it has a longer e uilibration period performed at a t mperature of 30'C (860 ) and a relative hum ity (RH) of 95% (or 70% RH with humidity control), and i has more stringent t lerances that improve repeata ility of the test. Allowable Penet ation = [(100% - Methy Iodide Efficiently

fo Charcoal Credited 0

in Licensee's A ident Analysis) / Safe Factor] When ASTM D 803-1989 is used with 01C (860 F) and 95% H (or 70% RH with humidity ntrol) is used, the staff iii accept the followir: Safety fa or >_2 for systems with or without humidity ntrol. Humidity contr I can be provided by h aters or an NRC-app ved analysis that demonstrates hat the air entering the harcoal will be maint ined less than or equal to 70 pe cent RH under worst- se design-basis cond ions. BVWOG STS 5.5-10 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 96 of 168

Attachment 1, Volume 16, Rev. 0, Page 97 of 168 5.5 CTS 0 INSERT 7 Safety Related Ventilation System Flowrate (cfm) 6.8.4.f.2) Station EVS > 7200 and < 8800 CREVS > 2970 and < 3630 0 INSERT 8 Safety Related Ventilation System Penetration (%) RH (%) 6.8.4.f.3) Station EVS <2.5 95 CREVS <2.5 70 Insert Page 5.5-10 Attachment 1, Volume 16, Rev. 0, Page 97 of 168

Attachment 1, Volume 16, Rev. 0, Page 98 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5{ Ventilation Filter Testing Program (continued) 0 If the system has; face velocity greater t an 110 percent of 0. 03 m/s

                          ,(40 ftlmin), the fa e velocity should be sp cified.
                           *This value sho d be the efficiency that       s incorporated in t e licensee's 0

accident analys' which was reviewed d approved by the s If in a safety evaluation. safety related. Demonstrate for each of the systems that the pressure drop across 6.8.4.f.4) the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance withoRegulatory Guide 1.52, Revision 2, andIASME N510-F[7 at the system flowrate specified below [- %I ESF Ventilato ystem Delta P Flowrate [I: [ ]ý/ [ I/

e. Demonstra that the heaters for ach of the ESF systoms dissipate the value spec ied below [+/- 10%] wh n tested in accorda ce with

                                  '[ASME N5 0-1989].

ES Ventilation System Wattage]

                                            ,      [ I                     [ I                                     J The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

3.11.1, 3.11.2 5,5. MExplosive Gas and Storage Tank Radioactivity Monitoring Program 00 This program provides controls for potentially explosive gas mixtures contained in the *aste Gaspoýý Syster-Mlftheq-u-a-ntity of radioactivity cont~tned in gas 0 Istorage tanks or/fed into the off as treatment si'stem, and the quantity or radioactivity contained in unprotected outdoorfliquid storage tankk eegaseous radioactivity quantiteA shall be determined following the metho ology in [Branch Technical Position ( TP) ETSB 11-5, "Postulated Radioactive elease due to 0 Waste Gas System/Leak or Failure"]. The liquid radwaste qua tities shall be determined in acc 1dance with [Standard Review Plan, Secti 15.7.3, "Postulated RadioActive Release due to Tank Failures"]. BWOG STS 5.5-11 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 98 of 168

Attachment 1, Volume 16, Rev. 0, Page 99 of 168 5.5 CTS 0 INSERT 9 Safety Related Ventilation System Delta P (inches wcq) Flowrate (cfm) 6.8.4.f.4) Station EVS <6 7200 and < 8800 CREVS < 4.4 > 2970 and < 3630 Insert Page 5.5-11 Attachment 1, Volume 16, Rev. 0, Page 99 of 168

Attachment 1, Volume 16, Rev. 0, Page 100 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 3.11.1, 3.11.2 .5[ Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) 0 The program shall include: 3.11.2, a. The limits for concentrations of hydrogen and oxygen in the[jaste Gas 4.11.2 Ho u Systemrrand 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) L = 00

b. A surveillance pr ram to ensure that the quantity of ra activity contained in [each gas sto ge tank and fed into the offgas treat nt system] is less 00 than the amo t that would result in a whole body e osure of -e0.5 rem to any individu in an unrestricted area, in the event [an uncontrolled 3.1 4.11.1 temporary A surveillance program to ensure that the quantity of radioactivity contained in all out iqui tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank 0

overflows and surrounding area drains connected to theRLiquid Radwaste Treatment Systerlislless than the amo"uFt that Would result inale _<10ci, excluding Jconcentratiops less than the limits of 11 FR 20, Appendix B, Ta 2 tritium and dissolved --------- gretasnes.no~water Column 2, a/t the nearest potable water /upply and the nearest s ~fce supply in an unrestricted area, inlthel event of an uncentroje d release/ 0 gases.

of the tanl*' 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 Purveillance fequencies. M] 00 4.81.12.b 5L.5. Diesel Fuel Oil Testina Proaram 0 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 ___

0

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil and 0

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

BWOG STS 5.5-12 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 100 of 168

Attachment 1, Volume 16, Rev. 0, Page 101 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 4.8.1.1.2.b 55 Diesel Fuel Oil Testing Program (continued) 0

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 oila d 0

c. Total particulate concentration of the fuel oil is _*10 mg/I when tested every 31 days.

The provisions of SR 3.0; and SR 3.0.3 are applicable to the Diesel Fuel Oil 6.17 Testing Program testing frequencies. 00 5.5.~ Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications. 6.17.a a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews. 6.17.b b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following: 6.17.b.1) 1. A change in the TS incorporated in the license or 0 6.17.b.2) 2. A change to the updafed'FSAR or Bases that requires NRC approval 0 pursuant to 10 CFR 50.59. 6.17.c c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with .FSAR. Specification 0 6.17.d d. Proposed changes that meet the criteria oý .5.b Move shall be reviewed 0 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 MO0 5,.5 Safety Function Determination Program (SFDP) 0 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 limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following; 0 BWOG STS 5.5-13 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 101 of 168

Attachment 1, Volume 16, Rev. 0, Page 102 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC M01 5.5.ý Safety Function Determination Program (continued) 0 1414 Provisions for cross train checks to ensure a loss of the capability to 3 perform the safety function assumed in the accident analysis does undetected LO not go Provisions of fifunction condition for ensuring the plant is maintained in a safe condition if a loss exists Cl Provisions to ensure that an inoperable supported system's Completion 0 Iinoperabilities Time is not inappropriately extended as a result of multiple support system and 0

                               -         Other appropriate limitations and remedial or compensatory actions.                 (

s-. A loss of safety function exists when, assuming no concurrent single failure, no (D concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable and A required system redundant to the system(s) supported by the inoperable 3 support system is also inoperable or 0 A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable or 0

                                                                                               -IJ0 A required system redundant to the support system(s) for the supported                (Do described in                    systems                above is also inoperable.

Specifications 5.5.14.*b.'1 and 5.5..14.b.2and -. The SFDP identifies where a loss of safety function exists. If a loss of safety 0 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. 6.16 5.5.. Containment Leakage Rate Testing Program 0 [OPTION A]

a. A program hall establish the leak ge rate testing of the containment as required b 10 CFR 50.54(o) and C0CFR 50, Appendix J, Option A, as modified approved exemption

b. The maxi um allowable contain ent leakage rate, La at Pa, shall be [ ]%

of contai ment air weight per da BWG STS 5.5-14 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 102 of 168

Attachment 1, Volume 16, Rev. 0, Page 103 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.~ Containment Leakage Rate Testing Program (continued) 0

c. Leakage ra acceptance criteria ar :

1. Cont nment leakage rate a ptance criterion is :__ .0 La. During the first u it startup following testi g in accordance with his program, the leaka e rate acceptance crite ia are < 0.60 La for th Type B and C tests nd < 0.75 La for Type tests.

2. Air I ck testing acceptance c iteria are:

a) Overall air lock leakag rate is _ [0.05 La] wh n tested at _>Pa. b) For each door, leakag rate is _<[0.01 La] wh n pressurized to [> 10 psig].

d. The prov sions of SR 3.0.3 are a plicable to the Contair ment Leakage Rate Te ing Program.

e. Nothing n these Technical Spec' cations shall be cons rued to modify the testing requencies required by 0 CFR 50, Appendix I[OPT DN B]I 6.16.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 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 visua examination of contaoment concrete surf ces intended to fulfill the r quirements of 10 CF 50, Appendix J, 0 tion B testing, will be pe ormed in accordanc with the requireme ts of and frequen¢ specified by the AS E Section XI Code, ubsection IWL, except ere relief has been a thorized by the NR INSERT 10 2. The visu I examination of the teel liner plate insid containment intende to fulfill the require nts of 10 CFR50, pendix J, Option , will be performed in ccordance with the requirements of and fre uency specified by th ASME Section XI ode, Subse ion IVWE, except wher relief has been aut orized by the NRC.

[3I . -...] BWOG STS 5.5-15 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 103 of 168

Attachment 1, Volume 16, Rev. 0, Page 104 of 168 5.5 CTS (DINSERT 10 6.16.a. 1) 1. A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1. 6.16.a.2) 2. The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, as-found testing will not be required. Insert Page 5.5-15 Attachment 1, Volume 16, Rev. 0, Page 104 of 168

Attachment 1, Volume 16, Rev. 0, Page 105 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals Containment Leakage Rate Testing Program (continued) 0 6.16.b b. The calculated peak containment internal pressure for the design basis loss 38 of coolant accident, Pa, ij 4 .. IThe containmen sign pressure is

c. The maximum allowable containment leakage rate, L8, at Pa, shall be Ao 6.16.c of containment air weight per day. 0.5 l 0

6.16.d d. Leakage rate acceptance criteria are: 6.16.d.1) 1. Containment leakage rate acceptance criterion is er 5La. -- During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are M.60La for the Type B and C tests and < 0.75 La for Type A tests. 6.16.d.3) 2. Air lock testing acceptance criteria are:

6.16.d.3)a) a) Overall air lock leakage rate is < 1[005 L.1 when tested at >_Pa.

0 6.16.d.3)b) b) For each door, leakage rate is <_ri.Ol L,!Jwhenfressurized to

                                                   ~Ž 1 psg~Jth~evolumne between the door seals isý_

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

f. Nothing in thesl!e-echnical Specifications shall be coprued to modify the testing Fre dencies required by 10 CFR 50, Appodix J. 0

[OPTION A/B Co bined]

a. A program all establish the leaka rate testing of the ntainment as required by 0 CFR 50.54(o) and 1 CFR 50, Appendix J. [Type A][Type B and C] test equirements are in a rdance with 10 CFR 0, Appendix J, Option A, a modified by approved xemptions. [Type B nd C][Type A]

test require ents are in accordanc with 10 CFR 50, App ndix J, Option B, as modifie by approved exemptio s. The 10 CFR 50, A pendix J, Option B t st requirements shall b in accordance with th guidelines contained n Regulatory Guide 1.1 3, "Performance-Bas d Containment Leak-Test Program," dated Septe ber, 1995, as modifie by the following exception :

1. The visual examination of c tainment concrete s rfaces intended to fulfil the requirements of 10 FR 50, Appendix J, ption B testing, will e performed in accord nce with the requirem nts of and fre uency specified by the SME Section XI Cod , Subsection IWL, ex pt where relief has bee authorized by the N C.

BVWOG STS 5.5-16 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 105 of 168

Attachment 1, Volume 16, Rev. 0, Page 106 of 168 CTS Programs.and Manuals 5.5 5.5 Programs and Manuals 5.5.03 Containment Leakage Rate Testing Program (continued) 0

2. The viua xamination of the teel liner plate inside ntainment intendi d to fulfill the require nts of 10 CFR50, App ndix J, Optio B, will be performed in ccordance with the r quirements of and fr, quency specified by th ASME Section XI Co e, Subs ction IWE, except wher relief has been autho ized by the NRC.

[3. .... ]

b. The calcul ted peak containment i ernal pressure for the design basis loss of coolant ccident, Pa, is [45 psig]. The containment des Ign pressure is

[50 psig].

c. The maxi um allowable containm nt leakage rate, La, at Pa, shall be [ ]%

of contain nt air weight per day.

d. Leakage r te acceptance criteria a e:

1. Con inment leakage rate a eptance criterion is

1.0 La. During the first nit startup following tes ing in accordance wit this program, the leak ge rate acceptance crit ria are < 0.60 La for t e Type B and C test and [< 0.75 La for Optio A Type A tests] [_<0 5 La for Option B Typ A tests].

2- Air Ick testing acceptance riteria are: a) Overall air lock leakag rate is< [0.05 La] wh n tested at - Pa. b) For each door, leakag rate is !5 [0.01 La] wh n pressurized to [> 10 psig].

e. The prov sions of SR 3.0.3 are a plicable to the Contai ment Leakage Rate Te ing Program.

f. Nothing n these Technical Speci ications shall be cons rued to modify the testing equencies required by 0 CFR 50, Appendix BVWOG STS 5.5-17 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 106 of 168

Attachment 1, Volume 16, Rev. 0, Page 107 of 168 CTS Programs and Manuals 5.5 5.5 Programs and Manuals DOC A06 5ý.5. Battery Monitoring and Maintenance Program 0 160 This Program provides for battery restoration and maintenance, based on jthe recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice 0 for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturerM including the following: 0

a. Actions to restore battery cells with float voltage < M2.13E]V, and 0

b. Actions to equalize and test battery cells that had been discovered with al ,t-"rnl\u'o Ioval-, ha=lrn

                                                    .'-  '- fha= Inni,-  *=fhi*t='                lLZ22tn
                                                                                            ,.',".,n               f h22n~tJ    TSTF-I.-I    h.ln-   tKa lminiý"m actnWici;-1 rl.ci              1*----FtW of the  lates 451T TSTF-INSERT 10            448 BWOG STS                                                5.5-18                                Rev. 3.1, 12/01105 Attachment 1, Volume 16, Rev. 0, Page 107 of 168

Attachment 1, Volume 16, Rev. 0, Page 108 of 168 5.5 CTS INSERT 10 DOC A06 5.5. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of M5 rem whole body or its equivalent to any part of the body [5 rem tpal effective 06se equivalot (TEDE)]Ifor the duration of the accident. The program shall include the following elements:

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

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

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Section C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0 4-4*

[The following e exceptions to Sections .1 and C.2 of Regulatory G ide 1.197, Revision 0:

1. ;and] 7

d. Measurements, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a (7 Frequency of months on a STAGGERED TEST BASIS. The results shall e trended and used as part of the[m month assessment of the CRE boundary,.*_ El Insert Page 5.5-18a Attachment 1, Volume 16, Rev. 0, Page 108 of 168

Attachment 1, Volume 16, Rev. 0, Page 109 of 168 5.5 CTS 9INSERT 10 (continued) DOC \06 e. The quantitative limits on unfiltered air leakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the ecoatioo.5.17.cL testing described in ar[ 03 . The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis",-., 0

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by para , respectively.

                                                              '   Specifications 5.5.177.c and 5.5.17.d Insert Page 5.5-18b Attachment 1, Volume 16, Rev. 0, Page 109 of 168

Attachment 1, Volume 16, Rev. 0, Page 110 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

1. These 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. The brackets have been removed and the proper plant specific information/value has been provided.

3. This Specification has been renumbered to be consistent with the ITS format and for clarity.

4. Changes are made to be consistent with the Radioactive Effluent Controls Program limits in the Davis-Besse CTS.

5. The bracketed ISTS 5.5.3, "Post Accident Sampling," is not included in the Davis-Besse ITS. The requirement for Post Accident Sampling was deleted from the CTS in License Amendment 264, dated June 10, 2005. Subsequent programs have been renumbered, as necessary.

6. Changes made to be consistent with the current Program, which is controlled in the Davis-Besse Technical Requirements Manual.

7. ISTS 5.5.7 provides requirements for the Pre-Stressed Concrete Containment Tendon Surveillance Program. This bracketed requirement regarding Pre-Stressed Concrete Containment Tendon Surveillance Program is deleted because it is not applicable to Davis-Besse. The Davis-Besse containment does not utilize pre-stressed concrete containment tendons. Subsequent programs have been renumbered, as necessary.

8. ITS 5.5.4, "Reactor Vessel Internals Vent Valves Program," has been added to the ITS. The design of the Davis-Besse reactor vessel includes internal vent valves.

This requirement is currently a Surveillance Requirement in CTS 3/4.4.10.1. It has been included as a program similar to the Reactor Coolant Pump Flywheel Inspection Program, which is also a Surveillance Requirement in CTS 3/4.4.10.1.

9. The Inservice Testing (IST) Program (ISTS 5.5.7) has been modified to state that the IST Program provides control for ASME Code Class 1, 2, and 3 "pumps and valves" in place of the current "components." 10 CFR 50.55a(f) provides the regulatory requirements for an IST Program. It specifies that ASME Code Class 1, 2, and 3 pumps and valves are the only components covered by an IST Program.

10 CFR 50.55a(g) provides regulatory requirements for an Inservice Inspection (ISI) Program. It specifies that ASME Code Class 1, 2, and 3 components are covered by the ISI Program, and that pumps and valves are covered by the IST Program in 10 CFR 50.55a(f). The ISTS does not include ISI Program requirements as these requirements have been relocated to a plant specific document. Therefore, the components to which the IST Program applies (i.e., pumps and valves) have been added for clarity. 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. Davis-Besse Page 1 of 3 Attachment 1, Volume 16, Rev. 0, Page 110 of 168

Attachment 1, Volume 16, Rev. 0, Page 111 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

10. The Reviewer's Note has been deleted since it is not intended to be included in the ITS.

11. The program details of the Explosive Gas and Storage Tank Radioactivity Monitoring Program are described in ISTS 5.5.12.a and 5.5.12.b (ITS 5.5.11.a and 5.5.11 .b). Therefore, the sentence in the introductory paragraph that specifies a method to determine the explosive gas and storage tank radioactivity is not necessary. Additionally, the requirements specified in ISTS 5.5.12.b do not apply to Davis Besse. UFSAR Section 15 states that a waste gas decay tank release is not a credible accident and that the dose will remain within the 10 CFR 100 guidelines. This is also consistent with the CTS.

12. ISTS 5.5.16 (ITS 5.5.15) 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 5.5.15.a), which states that the 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." These exceptions stated in ITS 5.5.15.a are modifications to the testing Frequencies required by 10 CFR 50, Appendix J. In addition, there is no need to state any specific exception to any of the other requirements of the Specifications that discuss testing frequencies, because the convention of application of requirements in the sections of ISTS 5.5 is that no other Specification requirements apply unless otherwise stated. For example, ISTS SR 3.0.2 does not apply to any of the ITS 5.5 sections, unless specifically noted. Therefore, there is no need to include a statement that ITS SR 3.0.2 does not apply to the Frequencies of ITS 5.5.15.

13. Typographical/grammatical error corrected.

14. The Davis-Besse plant-specific reactor coolant pump flywheel inspection requirements have been provided. These requirements were approved by the NRC in License Amendment 232, dated June 8, 1999. An allowance to apply the provisions of ITS SR 3.0.2 and ITS SR 3.0.3 has been provided, consistent with the current licensing basis (the program is currently a Surveillance, thus these allowances apply).

15. Changes are made to be consistent with the current Steam Generator Program requirements in the Davis-Besse CTS.

16. Changes are made to be consistent with the current Secondary Water Chemistry Program requirements in Davis-Besse License Condition 2.C(5).

17. Changes are made to be consistent with the current Ventilation Filter Testing Program requirements in the Davis-Besse CTS.

18. Changes are made to ISTS 5.5.12.c (ITS 5.5.11.b) to be consistent with the first paragraph in ISTS 5.5.12 (ITS 5.5.11) and with the current licensing basis.

Davis-Besse Page 2 of 3 Attachment 1, Volume 16, Rev. 0, Page 111 of 168

Attachment 1, Volume 16, Rev. 0, Page 112 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS

19. Changes are made to be consistent with the current limits for the outdoor temporary liquid storage tanks in the Davis-Besse CTS.

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

21. Editorial changes for consistency. These items and paragraphs are Specifications.

22. Davis-Besse complies with Option B of 10 CFR 50, Appendix J. Therefore, the Option A and combined Option A and B provisions have been deleted.

23. The Davis-Besse exceptions have been provided, consistent with the current licensing basis. The containment design pressure limit specified in ISTS 5.5.16.b has not been included because it currently does not exist in the Davis-Besse CTS, and because this limit does not provide any useful input to the Containment Leakage Rate Testing Program. The Davis-Besse specific limits for Types B and C leakage (5 0.60 La) have been provided, consistent with current licensing basis.

Furthermore, the specific manner in which the air lock door seal test is performed has been included, consistent with current licensing basis. Davis-Besse Page 3 of 3 Attachment 1, Volume 16, Rev. 0, Page 112 of 168

Attachment 1, Volume 16, Rev. 0, Page 113 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 113 of 168

Attachment 1, Volume 16, Rev. 0, Page 114 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 114 of 168

, Volume 16, Rev. 0, Page 115 of 168 ATTACHMENT 6 ITS 5.6, REPORTING REQUIREMENTS , Volume 16, Rev. 0, Page 115 of 168
, Volume 16, Rev. 0, Page 116 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 116 of 168

Attachment 1, Volume 16, Rev. 0, Page 117 of 168 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 5.6 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted Ito the appr6priate Regional .fMice unlessi A lotherwie notedj in accordance with 10 CFR 50.4 A02 STARTUP REPORT 6.9.1.1 Deleted. 6.9.1.2 Deleted. 6.9.1.3 Deleted. ANNUAL OPERATING REPORT 6.9.1.4 Annual reporcovering the activities of ýe unit during the previo be submittedprior tMarch 31 of each year./ L--e 6.9.1.5 Reports ruired on an annual basis sl3Al include:

a. Deleted

b. Deleted

c. The results of speci c activity analysis in which the primary cool t exceeded the limits of Specification ca 3.4 8. 'Me following information shall be includ : (1) Reactor power history starting 48 ours prior to the first sample in which the Ii t was exceeded; (2)

Results of the lastt isotopic analysis for radioiodine performed p. or to exceeding the limit, al s while limit was exceeded and results of one alysis after the results of analysi radioiodine ac i s acti th su should include date and. was reduced to less than limit. Each resul time of sampli and the radioiodine concentrations; (3) Cl -up system flow history L--e starting 48 ho prnior to the first sample in which the liurni!i as exceeded; O'ý (4) Graph of mý c and one other radioiodine isotope the 1-131 con 9 tration centration in microcuries per gram asI fiinction u I of time for the duration of the speci c activity above the steady-state level;

d The time duration when the specific ac vity of the primary coolant exceeded je e radioiodine limit. re h MONTHLY OPERATING REPORT 6.9.1.6 Deleted DAVIS-BESSE, UNIT I 6-13 Amendment No.-% -4i;-5-2,-P 93,t1-O

                                                                                      -1*-~,       26~-
                                                                                                     -- - 276 Page 1 of 5 Attachment 1, Volume 16, Rev. 0, Page 117 of 168

Attachment 1, Volume 16, Rev. 0, Page 118 of 168 ITS IT)01 ITS 5.6 6.0 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 5.6.3.a 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle and any remaining part of a reload cycle for the following: 2.1.2 AXIAL POWER IMBALANCE Protective Limits for Reactor Core Specification 2.1.2 2.2.1 Trip Setpoint for Flux - AFlux/Flow for Reactor Protection System Setpoints Specification 2.2.1 3.1.1.3c Negative Moderator Temperature Coefficient Limit 3.1.3.6 Regulating Rod Insertion Limits

3. 1.3.7 Rodirogram A0 3.1.3.8 :7non Reactivit I

3.1.3.9 Axial Power Shaping Rod Insertion Limits (LCo 3.1.1, "SHUTDOWN MARGIN," LCO 3.1.7, "Position Indicator Channels," (SR 3.1.7.1 limits) 3.2.1 AXIAL POWER IMBALANCE LCO 3.1.8, "PHYSICS TEST Exceptions - MODE 1," A04 S3.9.1, LCO 3.1.9, "Boron"PHYSICS TEST Exceptions - MODE 2," LCO Concentration" 3.2.2 Nuclear Heat Flux Hot Channel Factor,FQ 391 "or C... 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor, F 3.2.4 QUADRANT POWER TILT 5.6.3.b The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be: those previously reviewed and approved by the NRC, as described in BAW-lO1 79P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses", or any other new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A. The applicable approved revision number for BAW-10179P-A at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT. The CORE OPERATING LIMITS REPORT shall also list any new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A.

                  *                    [ ~INSERT 5.6.3.d    The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

5.6.3.e The CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC oc e:lControl (A05/ IDesk with copip~d-tohe Regional Administrator a esident Inspector DAVIS-BESSE, UNIT 1 6-14 Amendment No. 1 4 4 54-- 8 -9 , 276 Page 2 of 5 Attachment 1, Volume 16, Rev. 0, Page 118 of 168

Attachment 1, Volume 16, Rev. 0, Page 119 of 168 ITS 5.6 ITS (0 INSERT 1 5.6.3.c As described in reference documents listed in accordance with the instructions given above, when an initial assumed power level of 102% of RATED THERMAL POWER is specified in a previously approved method, an actual value of 100.37% of RATED THERMAL POWER may be used when the input for reactor thermal power measurement of feedwater mass flow and temperature is from the Ultrasonic Flow Meter. The following NRC approved documents are applicable to the use of the Ultrasonic Flow Meter with a 0.37% measurement uncertainty: Caldon Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM*/TM System," Revision 0, dated March, 1997. Caldon Inc. Engineering Report-1 57P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM /TM or LEFM CheckPlus TM System," Revision 5, dated October, 2001. Insert Page 6-14 Page 3 of 5 Attachment 1, Volume 16, Rev. 0, Page 119 of 168

Attachment 1, Volume 16, Rev. 0, Page 120 of 168 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 5.6.1 6.9.1.10 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submittedMbefor, 'May Iff each year. The report by May 15 shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM, and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

      *1                                                 1Add proposed ITS 5.6.1, second paragrp            _   _ ( ý o RADIOACTIVE EFFLUENT RELEASE REPORT 5.6.2 6.9.1.11 The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and the Process Control Program, and (2)in conformance with 10 CFR 50.36a and Section IV.B.I of Appendix I to 10 CFR Part 50.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.g, 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,

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

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

f. Total number and percentage of tubes plugged or repaired 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 and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

Add proposed ITS 5.6.4 L03 4 Add proposed ITS 5.6.5 }

DAVIS-BESSE, UNIT 1 6-15 Amendment No. 86--4.- 276 Page 4 of 5 Attachment 1, Volume 16, Rev. 0, Page 120 of 168

Attachment 1, Volume 16, Rev. 0, Page 121 of 168 ITS 5.6 6.0 ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shal be submitted to the U.S. Nuclear Regulatory oommission in accordance with 10 CFR 5 .4 within the time period specified for each rt. These reports shall be submitted coverin the activities identified below pursuant to th requirements of the applicable reference speci cations:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b. Deleted A07e

c. Deleted

d. Deleted

e. Deleted

f. Deleted

g. Inoperable emote Shutdown System control circuit(s) o transfer switch(es) required for a serious ntrol room or cable spreading room fire, Sp fication 3.3.3.5.2.

6.10 RECORD RETENTIONSeCT 6.0J Records of facility activities shall be retained as described in the USAR Chapter 17 Quality Assurance Program. 6.11 Deleted 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of See ITS] 10 CFR Part 20: 5.7 ], 6.12.1 High radiation areas with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation:

a. Each entry way 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.

DAVIS-BESSE, UNIT 1 6-16 Amendment No.-9;,65;-86r, 94-,--10, 3-5r--70;, 47G,-M-2-7-23i0-2-35;, 276 0 Page 5 of 5 Attachment 1, Volume 16, Rev. 0, Page 121 of 168

Attachment 1, Volume 16, Rev. 0, Page 122 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 6.9.1 requires, in addition to the requirements of 10 CFR, reports be submitted to the Regional Office. ITS 5.6 requires that the reports be submitted in accordance with 10 CFR 50.4. This changes the CTS by removing the explicit requirement to send reports to the Regional Office. 10 CFR 50.4 provides distribution requirements for written communications to the NRC. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.4. This change is designated as administrative because it does not result in technical changes to the CTS. A03 CTS 6.9.1.7 requires, in part, that core operating limits be established and documented in the COLR for the CTS 3/4.1.3.7, "Rod Program," and CTS 3/4.1.3.8, "Xenon Reactivity." ITS 5.6.3.a does not include a reference to these Specifications. This changes the CTS by eliminating the reference to Rod Program and Xenon Reactivity limits being core operating limits that are included in the COLR. The Rod Program Specification is being relocated to the Technical Requirements Manual (TRM) (See CTS 3/4.1.3.7 DOC R01). The Xenon Reactivity program is being removed from the ITS (See CTS 3/4.1.3.8 DOC L01). This information is not included in the ITS, as stated in the various DOCs, and is therefore not included in the list of individual Specifications that address core operating limits in ITS 5.6.3. This change is acceptable because the information contained in the individual Specifications is no longer documented in the ITS. This change is designated as administrative because it does not result in technical changes to the CTS. A04 CTS 6.9.1.7 contains a list of the core operating limits established and documented in the COLR. ITS 5.6.3.a includes additional core operating limits established and documented in the COLR. These are LCO 3.1.1, "SHUTDOWN MARGIN (SDM)"; LCO 3.1.7, "Position Indicator Channels" (SR 3.1.7.1 limits); LCO 3.1.8, "PHYSICS TEST Exceptions - MODE 1"; LCO 3.1.9, "PHYSICS TEST Exceptions - MODE 2"; and LCO 3.9.1, "Boron Concentration." These limits had been previously addressed in the CTS, but are being moved to the COLR in the ITS, and because of this are listed in ITS 5.6.3.a. This changes the CTS by adding core operating limits established and documented in the COLR (and applicable methodology) because they are being moved there as part of changes to other parts of the CTS. Technical aspects of the changes are Davis-Besse Page 1 of 7 Attachment 1, Volume 16, Rev. 0, Page 122 of 168

Attachment 1, Volume 16, Rev. 0, Page 123 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS addressed in the Discussion of Changes for the respective individual ITS Specifications. This change is acceptable because it administratively documents changes made to other parts of the CTS and the COLR. This change is designated as administrative because it does not result in technical changes to the CTS. A05 CTS 6.9.1.7 requires the CORE OPERATING LIMITS REPORT (COLR) to be provided to the NRC document control desk with copies to the Regional Administrator and Resident Inspector. ITS 5.6.3.d requires the COLR to be provided to the NRC. This changes the CTS by removing the specifics regarding distribution of the reports to the NRC. 10 CFR 50.4 provides distribution requirements for written communications to the NRC. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.4. Furthermore ITS 5.6 states that all reports in ITS 5.6 be submitted in accordance with 10 CFR 50.4. This change is designated as administrative because it does not result in technical changes to the CTS. A06 ITS 5.6.5, "Post Accident Monitoring Report," provides the reporting requirements when Condition B of LCO 3.3.17, "Post Accident Monitoring (PAM) Instrumentation," is entered. The CTS does not include this report. This changes the CTS by adding a new PAM Report. This change is acceptable since the PAM Report is being added as a result of changes made to CTS 3.3.3.6. The addition of the allowance to provide a report to the NRC in lieu of a unit shutdown is discussed in ITS 3.3.17, DOC L01. This change is designated as administrative because the addition of the report is a result of a change justified in another Specification of the Davis-Besse ITS submittal. A07 CTS 6.9.2 requires special reports be submitted to the NRC and lists the CTS Specifications that require special reports to be submitted. The ITS does not require these special reports to be prepared and submitted. This changes the CTS by deleting the references to the CTS Specifications requiring special reports. Justification for disposition of each of the special report requirements is addressed by the Discussion of Changes for the respective ITS or CTS Specification. The purpose of CTS 6.9.2 is to identify the Specifications that require special reports to be submitted. This change is acceptable because the special reports are no longer required by the respective Specifications. Justification for disposition of each of the special report requirements is addressed by the Discussion of Changes for the respective ITS or CTS Specification (ITS 3.5.2 DOC L01, ITS 3.5.3 DOC L01, and ITS 3.3.18 DOC R01). This change is designated as administrative because it does not result in technical changes to the CTS. Davis-Besse Page 2 of 7 Attachment 1, Volume 16, Rev. 0, Page 123 of 168

Attachment 1, Volume 16, Rev. 0, Page 124 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS A08 This change to CTS 6.9.1.7 is provided in the Davis-Besse ITS consistent with License Amendment Request No. 05-0007, submitted to the USNRC for approval in FENOC letter Serial Number 3198, from Mark B. Bezilla (FENOC) to USNRC, dated April 12, 2007. As such, this change is administrative. MORE RESTRICTIVE CHANGES M01 The second paragraph of ITS 5.6.1 includes details required to be included in the Annual Radiological Environmental Operating Report. CTS 6.9.1.10 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. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 8 - Deletion of Reporting Requirements) CTS 6.9.1.4 requires the annual reports of CTS 6.9.1.5 to be submitted prior to March 31 of each year. CTS 6.9.1.5.c requires annual reporting of information regarding any instances when the specific activity limit of Specification 3.4.8 is exceeded. ITS 5.6 does not contain any requirements for such a report. This changes the CTS by not including the requirements for the annual reporting of instances when the Technical Specification specific activity limit for the primary coolant is exceeded. The purpose of CTS 6.9.1.4 and CTS 6.9.1.5.c is to specify the requirements for submitting information regarding any instances when the Technical Specification specific activity of Specification 3.4.8 is exceeded in an annual report. This change is acceptable because the regulations provide adequate details of reporting requirements, and the reporting of exceeding the 1-131 specific activity limit does not affect continued plant operation. Operations or conditions prohibited by the plant's Technical Specifications are required to be reported in accordance with 10 CFR 50.73. Subsequent reports would be provided if necessary, without requiring a specific annual report. This change is designated Davis-Besse Page 3 of 7 Attachment 1, Volume 16, Rev. 0, Page 124 of 168

Attachment 1, Volume 16, Rev. 0, Page 125 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS as less restrictive because the reports that would be submitted under the CTS will not be required under the ITS. L02 (Category 1 - Relaxation of LCO Requirements) CTS 6.9.1.10 requires the Annual Radiological Environmental Operating Report to be submitted before May 1 of each year. ITS 5.6.1 requires the Annual Radiological Environmental Operating Report to be submitted by May 15 of each year. This changes the CTS by allowing additional time to submit this report each year. The purpose of the due date for submitting the Annual Radiological Environmental Operating Report is to ensure that the report is provided in a reasonable period of time to the NRC for review. This change is acceptable because the report is still required to be submitted in a reasonable time frame. Given that 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 reports. This change is designated as less restrictive because it allows more time to prepare and submit the report to the NRC. L03 CTS 3/4.4.9.1 provides the requirements for the Reactor Coolant System (RCS) Pressure/Temperature (P/T) Limits. ITS 3.4.3, "RCS Pressure and Temperature (P/T) Limits," Discussion of Change LA02 describes that the specific P/T limits, including the P/T limit curves and the maximum heatup and cooldown rates, are being relocated to the PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR). ITS 5.6.4 provides the requirements for the PTLR. This changes the CTS by adding a PTLR to the Technical Specifications. Creation of a PTLR is consistent with the guidance provided in Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits. GL 96-03 requires that the P/T limits are generated in accordance with the requirements of 10 CFR 50, Appendix G, documented in an NRC-approved topical report incorporated by reference in the Technical Specifications. Accordingly, the Davis-Besse heatup/cooldown curves have been generated using the NRC-approved methods described in BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G," and meet the requirements of 10 CFR 50, Appendix G. Technical Specifications include a Limiting Condition for Operation (LCO) that establishes P/T limits for the RCS. The limits are defined by figures and values that provide an acceptable range of operating temperatures and pressures for heatup, cooldown, criticality, and inservice leak and hydrostatic testing conditions. These parameters are generally valid for a specified number of effective full-power years or for a specified period. License amendments are generally required at the end of the effective period for P/T limit curves or when surveillance specimens are withdrawn and tested. Processing amendment requests for changes to Technical Specification that are developed using an accepted methodology places an unnecessary burden on licensee and NRC resources. An alternative approach for controlling these limits was proposed during the development of the ISTS. This approach, like the one used for the Davis-Besse Page 4 of 7 Attachment 1, Volume 16, Rev. 0, Page 125 of 168

Attachment 1, Volume 16, Rev. 0, Page 126 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS CORE OPERATING LIMITS REPORT, would relocate the P/T curves and maximum heatup and cooldown rates to a PTLR and would reference that document in the affected LCOs and Technical Specification Bases. The guidance contained in GL 96-03 specifically requires licensees wishing to implement this line item Technical Specification improvement to: (1) Reference a methodology for developing the curves and setpoint that has been approved by the NRC; (2) Develop a PTLR or a similar document that contains the figures, values, parameters, and any explanations derived from the methodology; and (3) Make appropriate changes to the applicable sections of the Technical Specifications. The following provides a description of the Davis-Besse compliance with the listed requirements of GL 96-03: (1) The P/T limits currently contained in CTS 3.4.9.1 and to be contained in the PTLR (applicable through 21 EFPY) were generated in accordance with the methods described in BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G," consistent with the requirements of 10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. The NRC has reviewed the methods described in BAW-10046A and approved the topical report by issuance of a Safety Evaluation Report (SER) dated April 30, 1986. (2) Davis-Besse will develop a PTLR consistent with the requirements in ITS 5.6.4 as part of implementing the entire Improved Technical Specifications Amendment. As required by ITS 5.6.4, the initial PTLR will be provided to the NRC upon issuance by Davis-Besse as part of the implementation of the ITS Amendment. (3) Consistent with the guidance in Generic Letter 96-03 and in the format of NUREG-1430, Rev. 3.1, Davis-Besse is providing the proposed Technical Specifications changes associated with the PTLR as part of this ITS Amendment. The ISTS 5.6.4 Reviewer's Note also states that 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.

Davis-Besse Page 5 of 7 Attachment 1, Volume 16, Rev. 0, Page 126 of 168

Attachment 1, Volume 16, Rev. 0, Page 127 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS

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

7. 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 aA) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 2 uA), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology.

The following provides a description of the Davis-Besse compliance with the listed requirements of the Reviewer's Note:

1. The neutron fluence is calculated using the methodology described in BAW-2108, Rev. 1, "Fluence Tracking System," dated May 1992. This methodology was described to the NRC in the Davis-Besse License Amendment request for the current PIT limits curves, dated January 30, 1995 (approved in Amendment 199).

2. The Davis-Besse Reactor Vessel Material Surveillance Program complies with the requirements of Appendix H to 10 CFR 50 and is described in BAW-1543A, "Master Integrated Reactor Vessel Material Surveillance Program." This information was provided to the NRC in the Davis-Besse License Amendment request for the current P/T limits curves, dated January 30, 1995.

3. PORVs are not currently used for LTOP, thus are not required in ITS 3.4.12, "Low Temperature Overpressure Protection (LTOP)."

Therefore, the PTLR will not include any PORV lift setting requirements.

4. Davis-Besse calculates ART in accordance with Regulatory Guide 1.99, Revision 2. This information was also provided to the NRC in the Davis-Besse License Amendment request for the current P/T limits curves, dated January 30, 1995.

Davis-Besse Page 6 of 7 Attachment 1, Volume 16, Rev. 0, Page 127 of 168

Attachment 1, Volume 16, Rev. 0, Page 128 of 168 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS

5. The NRC has previously determined that the current limiting ART calculation complies with NUREG-0800, Standard Review Plan 5.3.2, as documented in the NRC SER approving Amendment 199, dated July 20, 1995.

6. The NRC has previously determined that the Davis-Besse P/T limit curves comply with 10 CFR 50 Appendix G, as documented in the NRC SER approving Amendment 199, dated July 20, 1995.

7. Davis-Besse has not tested any plant-specific capsules since the last time the P/T limit curves were adjusted and approved by the NRC in Amendment 199. Therefore, no changes need to be made at this time.

Based on the above information, the proposed addition of a PTLR to the Davis-Besse Technical Specifications is acceptable. Davis-Besse will continue to meet the requirements of 10 CFR 50, Appendix G and any changes to the Davis-Besse P/T limits will be generated in accordance with the NRC-approved methodology described in BAW-10046A, Rev. 2. Davis-Besse Page 7 of 7 Attachment 1, Volume 16, Rev. 0, Page 128 of 168

Attachment 1, Volume 16, Rev. 0, Page 129 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 129 of 168

Attachment 1, Volume 16, Rev. 0, Page 130 of 168 CTS

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 6.9 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

6.9.1.10 5.6.1 Annual Radiolooical Environmental Operatina Report [A single submittal combine sections ay be made for a multiple unit stationTe submittal should mmon to all units at the station.] 0 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. 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 Oin the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, 0 0 November 19791 In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a r Pt=inr'tiv " supplementary report as soon as possible. 6.9.1.11 5.6.2 Effluent Release Report 0

                                           ----- ------   -    -NOTE-                               --------

[A single submittal m y be made for a multiple unit station. he submittal shall combine sections omnto all units at the station; howev r, for units with separate radwaste ystems, the submittal shall specify th releases of 0 radioactive mater I from each unit.] The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted lprior to May*of each year in accordance with 0 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. BWOG STS 5.6-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 130 of 168

Attachment 1, Volume 16, Rev. 0, Page 131 of 168 CTS Reporting Requirements 5.6 5.6 Reporting Requirements 6.9.1.7 5.6.3 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[The individual speelfications that address core operatgheý limits must be L referenced her94 0

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, peci icayyose Tescribed the in shllowbng documen ed s EmIdentify the Tocal Report(s) by numbeSr and titlear identify the staff Safety Evaluati /n Report for a plant specific meth dology by NRC letter and date. Ther OLR will contain the complete idantification for each of the TS referenced/topical reports used to prepare th.COLR (i.e., report number, title, uevis on, date, and any suppleoren s). e 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 System (ECCS) limits, nuclear limits such as 07 SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met. 0

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

DOC L03 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 as well as heatup ana cooiaown rates snaii be estawuisneu an uoiocumenreu in Une r- I Lr* lor the following:

[The individual sp cations that address RCS pressiand temperature INSERT 4 limits must be r renced here.]

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 eport(s) by number and title or ide tify the NRC Safety Evaluation fc a plant specific methodology by N/C letter and date. The PTLR will cor ain the complete identification for e ch of the TS ISR referenced Topi I Reports used to prepare the PTLA (i.e., report number, title, revision, dýte, and any supplements).] BVWIOG STS 5.6-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 131 of 168

Attachment 1, Volume 16, Rev. 0, Page 132 of 168 5.6 CTS INSERT 1 6.9.1.7 1. SL 2.1.1.1, "Reactor Core Safety Limits";

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

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

4. LCO 3.1.7, "Position Indicator Channels," (SR 3.1.7.1 limits);

5. LCO 3.1.8, "PHYSICS TEST Exceptions - MODE 1";

6. LCO 3.1.9, "PHYSICS TEST Exceptions - MODE 2";

7. LCO 3.2.1, "Regulating Rod Insertion Limits";

8. LCO 3.2.2, "AXIAL POWER SHAPING ROD (APSR) Insertion Limits";

9. LCO 3.2.3, "AXIAL POWER IMBALANCE Operating Limits";

10. LCO 3.2.4, "QUADRANT POWER TILT (QPT)";

11. LCO 3.2.5, "Power Peaking Factors";

12. LCO 3.3.1,"Reactor Protection System (RPS) Instrumentation," Function 8 (Flux - AFlux - Flow) Allowable Value; and

13. LCO 3.9.1, "Boron Concentration."

OINSERT 2 6.9.1.7 as described in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses," or any other new NRC approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A. The applicable approved revision number for BAW-1 01 79P-A at the time of the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT (COLR). The COLR shall also list any new NRC approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A. Insert Page 5.6-2a Attachment 1, Volume 16, Rev. 0, Page 132 of 168

Attachment 1, Volume 16, Rev. 0, Page 133 of 168 5.6 CTS (O INSERT 3 6.9.1.7 c. As described in reference documents listed in accordance with the instructions given above, when an initial assumed power level of 102% of RTP is specified in a previously approved method, an actual value of 100.37% of RTP may be used when the input for reactor thermal power measurement of feedwater mass flow and temperature is from the Ultrasonic Flow Meter. The following NRC approved documents are applicable to the use of the Ultrasonic Flow Meter with a 0.37% measurement uncertainty:

1. Caldon Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMW/TM System," Revision 0, dated March, 1997.

2. Caldon Inc. Engineering Report-157P, "Supplement to Topical Report TM ER-80P: Basis for a Power Uprate with the LEFMq/TM or LEFM CheckPlus System," Revision 5, dated October, 2001.

O INSERT 4 DOC L03 1. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits." O INSERT 5 DOC L03 1. BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G," June 1986. Insert Page 5.6-2b Attachment 1, Volume 16, Rev. 0, Page 133 of 168

Attachment 1, Volume 16, Rev. 0, Page 134 of 168 CTS Reporting Requirements 5.6 5.6 Reporting Requirements DOC L03 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 th calculation of the P-T limits for NRC proval should include the following pr visions:

1. The methodolog shall describe how the neutron fluenc is calculated (reference new egulatory Guide when issued).

2. The Reactor Ve sel Material Surveillance Program shal comply with Appendix H to 1 CFR 50. The reactor vessel material rradiation surveillance sp imen removal schedule shall be provi ed, along with how the specimen e aminations shall be used to update th PTLR curves.

3. Low Temperat re Overpressure Protection (LTOP) Sy tem lift setting limits for the Power perated Relief Valves (PORVs), devel ped using NRC- 0 approved met odologies may be included in the PTLF.

4. The adjusted ference temperature (ART) for each r actor beltline material shall be calcul ted, accounting for radiation embrittle ent, in accordance with Regulato Guide 1.99, Revision 2.

5. The limiting RT shall be incorporated into the calcu ation of the pressure and tempera ure limit curves in accordance with NU EG-080 Standard Review Plan 5.3.2, Pressure-Temperature Limits.

6. The minimu temperature requirements of Append x G to 10 CFR Part 50 shall be inc rporated into the pressure and temper ture limit curves.

7. Licensees ho have removed two or more capsul s should compare for each surve Ilance material the measured increase n reference temperature (RTNDT) to e predicted increase in RTNDT, where he predicted increase in RTNDT is b sed on the mean shift in RTNDT plus th two standard deviation value (2*T specified in Regulatory Guide 1.99, R vision 2. Ifthe measured value exc eds the predicted value (increase in R NOT + 2cG), the licensee should pr vide a supplement to the PTLR to dem nstrate how the results affect the pproved methodology.

BWOG STS 5.6-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 134 of 168

Attachment 1, Volume 16, Rev. 0, Page 135 of 168 CTS Reporting Requirements 5.6 5.6 Reporting Requirements DOC A06 5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.*17 "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the 0 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 Su eillance Report Any abnor I degradation of the containment structure etected during the tests required y the Pre-stressed Concrete Containment T ndon Surveillance Progra shall be reported to the NRC within 30 days The report shall include a 0 descri tion of the tendon condition, the condition of e concrete (especially at tend anchorages), the inspection procedures, th tolerances on cracking, and the orrective action taken. ] 6.9.1.12 5.6 Steam Generator Tube Inspection Report 0 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__ "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG 0

b. Active degradation mechanisms found L_0 0

c. Nondestructive examination techniques utilized for each degradation mechanisrnrJ 0

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications-0y 0

e. Number of tubes plugged Mor repairedMj during the inspection outage for 00 each active degradation mechanisrnm-E 00

f. Total number and percentage of tubes plugged jor repairecg to dateL-0

g. The results of condition monitoring, including the results of tube pulls and in-situ testing 1J 0

0

                        ýh. The effective plugging percentage for all plugging land tube repairsi in each SG~andM~                                                                            0 M*i. Repair method utilized and the number of tubes repaired by each repair              0 method.0 BWOG STS                                           5.6-4                           Rev. 3.1, 12/01/05 Attachment 1, Volume 16, Rev. 0, Page 135 of 168

Attachment 1, Volume 16, Rev. 0, Page 136 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.6, REPORTING REQUIREMENTS

1. The brackets have been removed and the proper plant specific information/value has been provided.

2. ISTS 5.6.3 requires submittal of the Radioactive Effluent Release Report prior to May 1 of each year in accordance with 10 CFR 50.36a. 10 CFR 50.36a states that the report must be submitted within one year of the previous report. The existing Davis-Besse CTS submittal date is also in accordance with 10 CFR 50.36a; a May 1 date is not provided in the CTS. Therefore, the Davis-Besse current licensing basis reporting date has been maintained.

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

4. ISTS 5.6.6 provides requirements for the Tendon Surveillance Report. The Containment design at Davis-Besse does not include pre-stressed concrete tendons.

Therefore, this report is not included in the Davis-Besse ITS, consistent with the current licensing basis. Subsequent Specifications are renumbered as a result of this deletion.

5. These 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.

6. Typographical error corrected.

7. This change to ITS 5.6.3 is provided in the Davis-Besse ITS consistent with License Amendment Request No. 05-0007, submitted to the USNRC for approval in FENOC letter Serial Number 3198, from Mark B. Bezilla (FENOC) to USNRC, dated April 12, 2007. Due to this addition, the remaining paragraphs in ITS 5.6.4 have been renumbered.

Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 136 of 168

Attachment 1, Volume 16, Rev. 0, Page 137 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 137 of 168

Attachment 1, Volume 16, Rev. 0, Page 138 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS 10 CFR 50.92 EVALUATION FOR LESS RESTRICTIVE CHANGE L03 Davis-Besse is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1430, Rev. 3.1, "Standard Technical Specifications Babcock and Wilcox Plants." The proposed change involves making the Current Technical Specifications (CTS) less restrictive. Below is the description of this less restrictive change and the determination of No Significant Hazards Considerations for conversion to NUREG-1430. CTS 3/4.4.9.1 provides the requirements for the Reactor Coolant System (RCS) Pressure/Temperature (P/T) Limits. ITS 3.4.3, "RCS Pressure and Temperature (P/T) Limits," Discussion of Change LA02 describes that the specific P/T limits, including the P/T limit curves and the maximum heatup and cooldown rates, are being relocated to the PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR). ITS 5.6.4 provides the requirements for the PTLR. This changes the CTS by adding a PTLR to the Technical Specifications. Creation of a PTLR is consistent with the guidance provided in Generic Letter (GL) 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits. GL 96-03 requires that the P/T limits are generated in accordance with the requirements of 10 CFR 50, Appendix G, documented in an NRC-approved topical report incorporated by reference in the Technical Specifications. Accordingly, the Davis-Besse heatup/cooldown curves have been generated using the NRC-approved methods described in BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G," and meet the requirements of 10 CFR 50, Appendix G. Technical Specifications include a Limiting Condition for Operation (LCO) that establishes P/T limits for the RCS. The limits are defined by figures and values that provide an acceptable range of operating temperatures and pressures for heatup, cooldown, criticality, and inservice leak and hydrostatic testing conditions. These parameters are generally valid for a specified number of effective full-power years or for a specified period. License amendments are generally required at the end of the effective period for P/T limit curves or when surveillance specimens are withdrawn and tested. Processing amendment requests for changes to Technical Specification that are developed using an accepted methodology places an unnecessary burden on licensee and NRC resources. An alternative approach for controlling these limits was proposed during the development of the ISTS. This approach, like the one used for the CORE OPERATING LIMITS REPORT, would relocate the P/T curves and maximum heatup and cooldown rates to a PTLR and would reference that document in the affected LCOs and Technical Specification Bases. The guidance contained in GL 96-03 specifically requires licensees wishing to implement this line item Technical Specification improvement to: (1) Reference a methodology for developing the curves and setpoint that has been approved by the NRC; (2) Develop a PTLR or a similar document that contains the figures, values, parameters, and any explanations derived from the methodology; and Davis-Besse Page 1 of 4 Attachment 1, Volume 16, Rev. 0, Page 138 of 168

Attachment 1, Volume 16, Rev. 0, Page 139 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS (3) Make appropriate changes to the applicable sections of the Technical Specifications. The following provides a description of the Davis-Besse compliance with the listed requirements of GL 96-03: (1) The P/T limits currently contained in CTS 3.4.9.1 and to be contained in the PTLR (applicable through 21 EFPY) were generated in accordance with the methods described in BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G," consistent with the requirements of 10 CFR 50 Appendix G, and Regulatory Guide 1.99, Revision 2. The NRC has reviewed the methods described in BAW-10046A and approved the topical report by issuance of a Safety Evaluation Report (SER) dated April 30, 1986. (2) Davis-Besse will develop a PTLR consistent with the requirements in ITS 5.6.4 as part of implementing the entire Improved Technical Specifications Amendment. As required by ITS 5.6.4, the initial PTLR will be provided to the NRC upon issuance by Davis-Besse as part of the implementation of the ITS Amendment. (3) Consistent with the guidance in Generic Letter 96-03 and in the format of NUREG-1430, Rev. 3.1, Davis-Besse is providing the proposed Technical Specifications changes associated with the PTLR as part of this ITS Amendment. The ISTS 5.6.4 Reviewer's Note also states that 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 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. The minimum temperature requirements of Appendix G to 10 CFR Part 50 shall be incorporated into the pressure and temperature limit curves.

Davis-Besse Page 2 of 4 Attachment 1, Volume 16, Rev. 0, Page 139 of 168

Attachment 1, Volume 16, Rev. 0, Page 140 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS

7. 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 (2uz4) specified in Regulatory Guide 1.99, Revision 2. If the measured value exceeds the predicted value (increase RTNDT + 20"&), the licensee should provide a supplement to the PTLR to demonstrate how the results affect the approved methodology.

The following provides a description of the Davis-Besse compliance with the listed requirements of the Reviewer's Note:

1. The neutron fluence is calculated using the methodology described in BAW-2108, Rev. 1, "Fluence Tracking System," dated May 1992. This methodology was described to the NRC in the Davis-Besse License Amendment request for the current P/T limits curves, dated January 30, 1995 (approved in Amendment 199).

2. The Davis-Besse Reactor Vessel Material Surveillance Program complies with the requirements of Appendix H to 10 CFR 50 and is described in BAW-1 543A, "Master Integrated Reactor Vessel Material Surveillance Program." This information was provided to the NRC in the Davis-Besse License Amendment request for the current P/T limits curves, dated January 30, 1995.

3. PORVs are not currently used for LTOP; thus are not required in ITS 3.4.12, "Low Temperature Overpressure Protection (LTOP)." Therefore, the PTLR will not include any PORV lift setting requirements.

4. Davis-Besse calculates ART in accordance with Regulatory Guide 1.99, Revision 2. This information was also provided to the NRC in the Davis-Besse License Amendment request for the current P/T limits curves, dated January 30, 1995.

5. The NRC has previously determined that the current limiting ART calculation complies with NUREG-0800, Standard Review Plan 5.3.2, as documented in the NRC SER approving Amendment 199, dated July 20, 1995.

6. The NRC has previously determined that the Davis-Besse P/T limit curves comply with 10 CFR 50 Appendix G, as documented in the NRC SER approving Amendment 199, dated July 20, 1995.

7. Davis-Besse has not tested any plant-specific capsules since the last time the P/T limit curves were adjusted and approved by the NRC in Amendment 199.

Therefore, no changes need to be made at this time. Based on the above information, the proposed addition of a PTLR to the Davis-Besse Technical Specifications is acceptable. Davis-Besse will continue to meet the requirements of 10 CFR 50, Appendix G and any changes to the Davis-Besse P/T limits will be generated in accordance with the NRC-approved methodology described in BAW-10046A, Rev. 2. Davis-Besse Page 3 of 4 Attachment 1, Volume 16, Rev. 0, Page 140 of 168

Attachment 1, Volume 16, Rev. 0, Page 141 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS An evaluation has been performed to determine whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed change is the addition of PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR). The PTLR is created in accordance with the guidance provided by Generic Letter (GL) 96-03 and is consistent with the content of NUREG-1430, Rev. 3.1. The RCS P/T limits will continue to meet the requirements of 10 CFR 50, Appendix G, and will be generated in accordance with the NRC-approved methodology described in BAW-10046A, Rev. 2, "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G." Since the proposed change is essentially administrative in nature and does not involve any change to any values currently required by the Technical Specifications, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. As stated above, the proposed change is essentially administrative in nature. Accident initial conditions and assumptions remain as previously analyzed, and the proposed change does not introduce any new or different accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The margin of safety is not affected by the creation of a PTLR. Operation of the unit in accordance with the limits specified in the PTLR will continue to meet the requirements of 10 CFR 50, Appendix G, and will assure that a margin of safety is not significantly decreased as a result of the change. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, it is concluded that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, there is a finding of "no significant hazards consideration." Davis-Besse Page 4 of 4 Attachment 1, Volume 16, Rev. 0, Page 141 of 168

, Volume 16, Rev. 0, Page 142 of 168 ATTACHMENT 7 ITS 5.7, HIGH RADIATION AREA , Volume 16, Rev. 0, Page 142 of 168
, Volume 16, Rev. 0, Page 143 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 143 of 168

Attachment 1, Volume 16, Rev. 0, Page 144 of 168 ITS 5.7 ITS 6.0 ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission in accordance with 10 CFR 50.4 within the time period specified for each report. These reports shall -be submitted covering the activities identified below pursuant to the requirements of the applicable reference specifications:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

b. Deleted See ITS

c. Deleted

                                                                                                                   -4  5.6   ]

d. Deleted

e. Deleted

f. Deleted

g. Inoperable Remote Shutdown System control circuit(s) or transfer switch(es) required for a serious control room or cable spreading room fire, Specification 3.3.3.5.2.

6.11 Deleted 5.7 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601 (c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20: 5.7.1 6.12.1 High radiation areas with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation: 5.7.1 .a a. Each entry way 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. DAVIS-BESSE, UNIT I 6-16 Amendment No. % 6,49-3--9-f6, 11174-l-?21-,23-;23-,276 Page 1 of 4 Attachment 1, Volume 16, Rev. 0, Page 144 of 168

Attachment 1, Volume 16, Rev. 0, Page 145 of 168 ITS G IITS 5.7 6.0 ADMINISTRATIVE CONTROLS 6.12.1 (Continued) 5.7.1 .b b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures. 5.7.1 .c c. Individuals qualified in radiation protection procedures (e.g., health physics personnel) and personnel continuously escorted by such individuals may be exempted from the requirement for a RWP or equivalent while performing their assigned duties provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas. 5.7.1 .d d. Each individual (whether alone or in a group) entering such an area shall possess: 5.7.1 .d.1 1) A radiation monitoring device that continuously displays radiation dose rates in the area; or 5.7.1 .d.2 2) A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 5.7.1 .d.3 3) A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or Add proposed Specification L01

5. . -5.7.1 .d.4(i)

4) A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and be under the surveillance, asfcloed specified in cr

                                                                    ......          the RWP" ort equivalent, l      while in the area, by "and     with the means communicate    with to" 5.7.1.d.4.(ii)     -         means of closed circuit television, by personnel qualified in radiation protection                            individuals inthe area procedures responsible for controlling personnel radiation exposure in the area.;--                             who are covered by such surveillance.

5.7.1.e e. Except for individuals qualified in radiation protection procedurestentry into such areas shall bermade only after dose rates in the area have been determined and entry personnel M', are knowledgeable of them. \ Sorpersonnel " lcontinuously es corted/ 5.7.2 6.12.2 _ok_ e igh radiation areas with dose rates greater than 1.0 rem/hour at 30 centimeters by such individuals, from the radiation source or from any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation: 5.7.2.a a. Each entryway to such an area shall be conspicuously posted as a high radiation area and L0 shall be.provided with a locked ]door, gate, 9,r'6ther bari that prevents unauthorized entry, and, inaddition: L_ or continuousl guardeddoor or 0t (L003 SThese continuously escorted personnel will receive a pre-job briefing prior to entry into

                  --     uch areas. This dosequatedeoterminnation, nok*wledgieiaandpre-job briefing does not                                          @

DAVIS-BESSE, UNIT 1 6-17 Amendment No."25t, 276 Page 2 of 4 Attachment 1, Volume 16, Rev. 0, Page 145 of 168

Attachment 1, Volume 16, Rev. 0, Page 146 of 168 ITS 5.7 ITS 6.0 ADMINISTRATIVE CONTROLS 6.12.2.a (Continued) 5.7.2.a.1 1) All keys to such doors, gates, or other barriers shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee. 5.7.2.a.2 2) Doors, gates, or other barriers shall remain locked except during periods of personnel or equipment entry or exit. 5.7.2. b b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures. 5.7.2.c c.. Individuals qualified. in radiation protection procedures may be exempted from the requirement for a RWP or equivalent while performing radiation surveys in such areas provided that they are following plant radiation protection procedures for entry to, exit from, and work in such areas. 5.7.2.d d. Each individual (whether alone or in a group) entering such an area shall possess: 5.7.2.d.1 I) A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or 5.7.2.d.2 2) A radiation monitoring device that continuously transmits dose rate and cumulative dose infornation to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or 5.7.2.d.3 3) A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, 5.7.2.d.3.(i) Wi Be under the surveillance, as specified in the RWP or equivalent, while in the area, by an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or 5.7.2.6.3.(i) (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, by 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, or DAVIS-BESSE, UNIT 1 6-1 8 Amendment No. .23t, 276 Page 3 of 4 Attachment 1, Volume 16, Rev. 0, Page 146 of 168

Attachment 1, Volume 16, Rev. 0, Page 147 of 168 ITS 5.7 ITS 0 6.0 ADMINISTRATIVE CONTROLS 6.12.2.d (Continued) 5.7.2.d.4 4) In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation r s . monitoring device that continuously displays radiation dose rates in the area. continuously escorted

                                                                                                                       ý 7              by sch individuals, 5.7.2.e        e. Except for an individual qualified in radiation protection procedures,entry into such areas shall be made only after dose rates in the area have been determined and entry personne are knowledgeable of them." q*      These  continuously to entry            escorted into such areas. This dose ratewill personnel        receive a pre-job determination,       briefing prior knowledge, and                      L02 S   pre-job briefing does not require documentation prior to initial entry.

i*. , 5.7.2.f f. Such individual areas that are wiin a arger area that is controlled as a high radiation 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, but shall be barricaded and conspicuous, and a clearly visible flashing light shall be activated at the area as a warning device. 6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" See CTS 6.0 J (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License NPF-3 dated October 24, 1980. 6.13.2 By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified. 6.14 Deleted DAVIS-BESSE, UNIT 1 6-19 Order dated 10/24/80 Amendment No. 86,-l-7Or,23-,-2-3-,

                                                                                                               ,          276 0

Page 4 of 4 Attachment 1, Volume 16, Rev. 0, Page 147 of 168

Attachment 1, Volume 16, Rev. 0, Page 148 of 168 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox 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 6.12.1.d.4 provides one of the options for an individual entering a high radiation area, and requires the individual to be under the surveillance, by means of closed circuit television, by personnel qualified in radiation protection procedures. ITS 5.7.1 .d.4.(ii) includes a similar option; however it includes an additional requirement that the person have a means of communicating with the individuals in the high radiation area who are covered by such surveillance. This changes the CTS by requiring means to communicate with the associated individuals in the high radiation area. The purpose of CTS 6.12.1 .d.4 is to ensure personnel in high radiation areas are properly monitored. This change is acceptable because it provides additional guidance to ensure the personnel in the high radiation areas can be contacted if the need arises. This change is also consistent with a similar option provided for personnel entering a very high radiation area. This change is designated as more restrictive because additional requirements are added for personnel entering high radiation areas. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 6.12.1.d.4 states that each individual that enters a high radiation area with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation shall possess a self-reading dosimeter and be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, by personnel qualified in radiation protection procedures responsible for controlling personnel radiation exposure in the area. Davis-Besse Page 1 of 3 Attachment 1, Volume 16, Rev. 0, Page 148 of 168

Attachment 1, Volume 16, Rev. 0, Page 149 of 168 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA ITS 5.7.1.d.4.(ii) retains this same requirement. However, ITS 5.7.1 .d.4.(i) provides an additional option in lieu of that required by CTS 6.12.1.d.4. ITS 5.7.1.d.4.(i) states that each individual that enters a high radiation area with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation shall possess a self-reading dosimeter and 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. This changes the CTS by allowing an individual to be monitored directly instead of indirectly (i.e., by closed circuit television) when entering a high radiation area with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation. The purpose of CTS 6.12.1 .d.4 is to provide the method for monitoring the exposure of individuals in high radiation areas. This change is acceptable because it provides adequate means of monitoring the personnel in the high radiation areas, yet provides added flexibility for how to do it. Furthermore, this proposed option is currently allowed by CTS 6.12.2.d.3.(i) when entering high radiation areas with dose rates greater than 1.0 rem/hour at 30 centimeters. This change is designated as less restrictive because additional methods for monitoring personnel in high radiation areas have been provided. L02 (Category I - Relaxation of LCO Requirements) CTS 6.12.1.e states that except for individuals qualified in radiation protection procedures, entry into such areas (a high radiation area with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation) shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. CTS 6.12.2.e states that except for individuals qualified in radiation protection procedures, entry into such areas (a locked high radiation area with dose rates exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation) shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. ITS 5.7.1.e states that except for individuals qualified in radiation protection procedures, "or personnel continuously escorted by such individuals," entry into such areas (a high radiation area with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation) shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. Furthermore, ITS 5.7.1 .e requires that these continuously escorted personnel will receive a pre-job briefing prior to entry into such areas, and that this dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. ITS 5.7.2.e provides identical requirements for entry into a high radiation area 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 500rads/hours at 1 Meter from the radiation source or from any surface penetrated by the radiation. This changes the CTS by allowing additional personnel to enter high radiation areas prior to determining the current dose rates. Davis-Besse Page 2 of 3 Attachment 1, Volume 16, Rev. 0, Page 149 of 168

Attachment 1, Volume 16, Rev. 0, Page 150 of 168 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA The purpose of CTS 6.12.1 .e and CTS 6.12.2.e is to provide the entry requirements for high radiation areas. This will allow personnel to perform the task that requires entry into the high radiation area simultaneously with the radiation protection technicians performing the radiation survey. This allowance could reduce the total dose accumulated by site personnel for a given task. This change is acceptable because it provides adequate means to monitor the personnel and the personnel will receive a pre-job brief to ensure the dose is maintained as low as reasonably achievable. This change is designated as less restrictive because additional methods for monitoring personnel before dose rates in the area have been determined have been provided. L03 (Categoty 1 - Relaxation of LCO Requirements) CTS 6.12.2 requires 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/hours at 1 meter from the radiation source or from any surface penetrated by the radiation shall be locked and CTS 6.12.2.a states that the areas will be locked by a door, gate, or other barrier. ITS 5.7.2.a allows the areas to either be locked or continuously guarded by a door, gate, or other barrier. This changes the CTS by allowing the doors, gates, or other barriers to be guarded instead of being locked. The purpose of CTS 6.12.2.a is to prevent unauthorized access to a high radiation area 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/hours at 1 meter from the radiation source or from any surface penetrated by the radiation. This change is acceptable because adequate controls are maintained to prevent an unauthorized access, while allowing the reasonable flexibility in determining the proper methods to ensure unauthorized access. A guarded door, gate, or barrier provides a similar control over the area; an individual is providing the controls in lieu of a mechanical device (a lock). This change is designated as less restrictive because an additional method to prevent unauthorized access to a high radiation area is allowed in the ITS than is allowed in the CTS. Davis-Besse Page 3 of 3 Attachment 1, Volume 16, Rev. 0, Page 150 of 168

Attachment 1, Volume 16, Rev. 0, Page 151 of 168 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 0, Page 151 of 168

Attachment 1, Volume 16, Rev. 0, Page 152 of 168 CTS gHigh Radiation AreaQ 5.7 5.0 ADMINISTRATIVE CONTROLS 6.12 *5.7 High Radiation Area 0 As~provided in paragraph 20.1.601(c) of 10 CFR Part 20, the following controls shall .be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20: 6.12.1 5.7.1 High Radiation Areas with Dose Rates Not Exceedinq 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation 6.12.1.a a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment. 6.12.1.b b. Access to, and activities in. each such area shall be controlled by means of Radiation Work Permit (RVVP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures. 6.12.1.c c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. whether alone or in a group) 6.12.1.d d. Each individual org foupentering such an area shall posses 6.12.1.d.1) 1. A radiation monitoring device that continuously displays radiation dose rates in the area; 0g 6.12.1.d.2) 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoin tl M. 6.12.1 .d.3) 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area or 0 6.12.1 .d.4) 4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, DOC L01 (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, ffJgn 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 BVVOG STS 5.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 152 of 168

Attachment 1, Volume 16, Rev. 0, Page 153 of 168 CTS 1High Radiation .AreaO 5.7 0 15.7 High Radiation Area 0 0 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated 'by the Radiation (continued) 6.12.1.d.4) (ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance. 6.12.1.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. 6.12.2 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 remlhour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation 6.12.2.a a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded d gate at prevents unauthorized entry, and, in addition:oQ 6.12.2.a.1) 1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee 6.12.2.a.2) 2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit. 6.12.2.b b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures. 6.12.2.c c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas. BWOG STS 5.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 153 of 168

Attachment 1, Volume 16, Rev. 0, Page 154 of 168 CTS [High Radiation Areal 5.7 0 15.7 High Radiation Area 0Q 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 remn/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.12.2.d d. Each individual lor (_w*hether alone or in a grou_) ofuplentering such an area shall possess Pone of the Q followingg 6.12.2.d.1) d 1 A radiation monitoring device that continuously integrates the radiation ates in the area and alarms when the device's dose alarm)

                                                                                                                                '0 setpoint is reached, with an appropriate alarm setpoin t_.

6.12.2.d.2) 2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the areaw or y_ 6.12.2.d.3) 3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and I--] 6.12.2.d.3)(i) (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area._ -CE 6.12.2.d.3)(ii) (ii) Be under the surveillance, as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area. [Specifications 5.7.2.d.2 and 5.7.2.d.3 6.12.2.d.4) 4. In those cases where Foptions and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area. 6.12.2.e e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry. BVVOG STS 5.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 154 of 168

Attachment 1, Volume 16, Rev. 0, Page 155 of 168 CTS Diigh Radiation Areal 5.7 0

        !5.7 High Radiation Area V                                                                         0 5.7.2    High Radiation Areas with Dose Rates Greater than 1.0 rerm/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued) 6.12.2.f               f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

BWOG STS 5.7-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 16, Rev. 0, Page 155 of 168

Attachment 1, Volume 16, Rev. 0, Page 156 of 168 JUSTIFICATION FOR DEVIATIONS ITS 5.7, HIGH RADIATION AREA

1. The brackets have been removed and the proper plant specific information/value has been provided.

2. ISTS 5.7.1.d and ISTS 5.7.2.d allows only one member of a group of individuals.

entering a high radiation area to meet the associated monitoring requirements. ITS 5.7.1.d and ITS 5.7.2.d requires every individual, even if in a group, who enters a high radiation area to meet the associated monitoring requirements. This is consistent with the Davis-Besse current licensing basis.

3. Change made to be consistent with another similar Specification (i.e., ITS 5.7.2.d).

4. These 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.

5. Typographical/grammatical error corrected.

6. Change made to be consistent with the Davis-Besse current licensing basis. Not all entryways are doors or gates.

7. The proper Specification numbers have been provided.

Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 156 of 168

Attachment 1, Volume 16, Rev. 0, Page 157 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 157 of 168

Attachment 1, Volume 16, Rev. 0, Page 158 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.7, HIGH RADIATION AREA There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 158 of 168

Attachment 1, Volume 16, Rev. 0, Page 159 of 168 ATTACHMENT 8 DELETED CURRENT TECHNICAL SPECIFICATIONS Attachment 1, Volume 16, Rev. 0, Page 159 of 168

, Volume 16, Rev. 0, Page 160 of 168 CTS 6.0, ADMINISTRATIVE CONTROLS , Volume 16, Rev. 0, Page 160 of 168
, Volume 16, Rev. 0, Page 161 of 168 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 0, Page 161 of 168

Attachment 1, Volume 16, Rev. 0, Page 162 of 168 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.5.3 TECHNICAL REVIOW AND CONTROL ACTIVITIES 6.5.3.1 Activities which a t nuclear safety shall be conducted as follow

a. Plant procedures r uired by Section 6.8.1 and changes thereto sh I be prepared, reviewed and appro ed. Each such procedure or procedure chan shall be reviewed by an individual/grour. other than the individual/group which prepar the procedure or procedure change, ut who may be from the same organization the individual/group which prepared th procedure or procedure change. Plant proced res, (including plant administrative pro ures), Physical Security Plan Implementini Procedures and Davis-Besse Emergency Ian Implementing Procedures will be approv by procedurally authorized indivi s.

b. Temporary appro al of changes to plant procedures cited in S 'on 6.8.1 which clearly do not change th intent of the approved procedures, can be ma e by two members of the plant managem i t staff, at least one of whom holds a Senior R ctor Operator's License.

For changes to p ant procedures, which may involve a change i intent of the approved procedures, the on authorized in Section 6.5.3.1 a to appro e the procedure shall approve the ch ge LAO1

c. Proposed chan s or modifications to plant structures, system and components shall be reviewed as de ignated by procedurally authorized individual . Each such modification shall be reviewvd by an individual/group other than the indi i ual/group which designed the modificati , but who may be from the same organizatio as the individual/group which design the modifications. Implementation of modi cations to plant structures, systems and mponents shall be approved by proecdurally thorized individuals.

d. Proposed tes and experiments which affect plant nuclear s fety and are not addressed in the Safety An lysis individual/griup Report which shall be reviewed by an individu 1/group other than the prepared the proposed test or exp inent and shall be approved by procedura ly authorized individuals.

e. Individuals r ponsible for reviews performed in accordan e with Section 6.5.3.1 a, b, c and d above hall meet or exceed the appropriate qualifica on requirements of Section 4.2, .3.1, 4.4 or 4.6 of ANSI 18.1, 1971, and be p eviously designated by procedurall authorized individuals. Each such review sh I include a determination of whether an dditional, cross disciplinary, review is necess . If deemed necessary, such review shal be performed by the review personnel of the ppropriate discipline.

f. Each revie will include a determination of whether prio NRC approval is required pursuant t 10 CFR 50.59.

DAVIS-BESSE, UNIT 1 6-4 Amendment No. -109,-39r248,22* 276 Page 1 of 3 Attachment 1, Volume 16, Rev. 0, Page 162 of 168

Attachment 1, Volume 16, Rev. 0, Page 163 of 168 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission in accordance with 10 CFR 50.4 within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specifications:

a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.

See ITS 1 5.6

b. Deleted

c. Deleted

d. Deleted

e. Deleted

f. Deleted

g. Inoperable Remote Shutdown System control circuit(s) or transfer switch(es) required for a serious control room or cable spreading room fire, Specification 3.3.3.5.2.

6.10 RECORD RETENION LA02 Records of facility a vities shall be retained as described in the USS Chapter 17 Quality Assurance Pro 6.11 Deleted 6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of See ITS] 5.7] 10 CFR Part 20: 6.12.1 High radiation areas with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation:

a. Each entry way 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.

DAVIS-BESSE, UNIT 1 6-16 Amendment No. 9-65-865-, 7-9*-;94,-I-6,

                                                             -"";"",-14,--8,2-,-23-,23,         276 Page 2 of 3 Attachment 1, Volume 16, Rev. 0, Page 163 of 168

Attachment 1, Volume 16, Rev. 0, Page 164 of 168 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.12.2.d (Continued)

4) In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

e. Except for an individual qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel See ITS]

5.7 are knowledgeable of them.

f. Such individual areas that are within a larger area that is controlled as a high radiation 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, but shall be barricaded and conspicuous, and a clearly visible flashing light shall be activated at the area as a warning device.

6.13 ENVIRONMENTA QUALIFICATION 6.13.1 By no later than ahn J eq in tin the facility shall 30,1982 all safety-relatedelectrical equipm be qualified in accordance n ith the provisions of Division of Operating e n ea eactors "Guidelines for Evaluating Environmenn ne Qualification of Class IE Electrical Equipm in inn Operating Reactors" (DOR Guidelines); or, ra G-0588 "Interim Staff Position on qE n ur Enviro eental Qualification of Safety-Related Electrinrd F Equipment", qQ December 1979. Copies o0)f th vtra ees documents are attached to Order for Modl!ificati 5 at. ce f th. n of License NPF-3 dated October 24, 19,,80. 1980 6.13.2 By no later tht0 December uuf 1, 1980, complete and auditibleu rds must be available and maintained at a cen location 0 which describe the environmental 'al q lification 0 method used for all safety-rclated el tinical j equipment in sufficient detail to docurn t s0gp the degree tc of compliance with the DOR Gui lines j or NUREG-0588. Thereafter, such reco ds shouldC.be updated and maintained current as cmt equipment is replaced, further tested, or o sse RW further qualified. 6.14 Deleted DAVIS-BESSE, UNIT I 6-19 Order dated 10/24/80 Amendment No. 86,-- 272. 276 Page 3 of 3 Attachment 1, Volume 16, Rev. 0, Page 164 of 168

Attachment 1, Volume 16, Rev. 0, Page 165 of 168 DISCUSSION OF CHANGES CTS 6.0, ADMINISTRATIVE CONTROLS ADMINISTRATIVE CHANGES A01 CTS 6.13 requires that by June 30, 1982, all safety-related electrical equipment be environmentally qualified in accordance with the Division of Operating Reactors (DOR) Guidelines or NUREG-0588. It further requires that complete and auditable environmental qualification records be available and maintained at a central location by December 1, 1980. ITS Chapter 5.0 does not retain these requirements. This changes the CTS by deleting the requirement related to complying with the 10 CFR 50.69 requirements. These requirements have already been satisfactorily met by Davis-Besse, therefore this historical requirement is not needed to be maintained in the ITS. Environmental qualification requirements are adequately addressed in the Davis-Besse procedures implementing the requirements of 10 CFR 50.49, and need not be repeated in the ITS. This change is designated as an administrative change and is acceptable since the requirements that have been fulfilled and it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, QAPM, IST Program,or liP) CTS 6.5.3, Technical Review and Control, explains how activities that affect nuclear safety (i.e., changes to procedures, changes or modifications to plant structures, systems, and components, and proposed tests and experiments) shall be conducted. ITS Chapter 5.0 does not retain these requirements. This changes the CTS by moving the Technical Review and Control requirements to the Quality Assurance Program Manual (QAPM). The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is acceptable because these types of procedural details will be adequately controlled in the QAPM. Any changes to the QAPM 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 requirements are being removed from the Technical Specifications. Davis-Besse Page 1 of 2 Attachment 1, Volume 16, Rev. 0, Page 165 of 168

Attachment 1, Volume 16, Rev. 0, Page 166 of 168 DISCUSSION OF CHANGES CTS 6.0, ADMINISTRATIVE CONTROLS LA02 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, UFSAR, ODCM, QAPM, IST Program,or liP) CTS 6.10, Record Retention, requires that records of activities be retained as described in the USAR Chapter 17 Quality Assurance Program. ITS Chapter 5.0 does not retain this requirement. This changes the CTS by moving the record retention requirements to Quality Assurance Program Manual (QAPM). The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is acceptable be these types of procedural details will be adequately controlled in the QAPM. Any changes to the QAPM 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 requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 16, Rev. 0, Page 166 of 168

Attachment 1, Volume 16, Rev. 0, Page 167 of 168 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 0, Page 167 of 168

Attachment 1, Volume 16, Rev. 0, Page 168 of 168 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 6.0, ADMINISTRATIVE CONTROLS There are no specific NSHC discussions for this Specification. Davis-Besse Page 1 of 1 Attachment 1, Volume 16, Rev. 0, Page 168 of 168}}

ML072200485

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