ML17263A641

Application for Amend to License DPR-18,revising TS Administrative Controls Section 6.0 Consistent W/Improved TS

ML17263A641
Person / Time
Site:	Ginna
Issue date:	05/13/1994
From:	MECREDY R C ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A642	List: ML17263A641 ML17263A643 ML17309A543
References
NUDOCS 9405170066
Download: ML17263A641 (36)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of))Rochester Gas and Electric Corporation

)(R.E.Ginna Nuclear Power Plant))Docket No.50-244 APPLICATION FOR AMENDMENT TO OPERATING LICENSE AND UALITY ASSURANCE PROGRAM Pursuant to Section 50.90 of the regulations of the U.S.Nuclear Regulatory Commission (NRC), Rochester Gas and Electric Corporation (RGEE), holder of Facility Operating License No.DPR-18, hereby requests that the Technical Specifications set forth in Appendix A to that license be amended.This request for change in Technical Specifications is to revise the Administrative Controls e section consistent with Improved Technical Specifications.

Several requirements are also relocated to other programs and documents in accordance with the criteria contained in the NRC Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors.A description of the amendment request, necessary background information, justification of the requested change, safety evaluation and no significant hazards and environmental considerations are provided in Attachment A.This evaluation demonstrates that the proposed changes do not involve a significant change in the types or a significant increase in the amounts of effluents or any change in the authorized power level of the facility.The proposed changes also do not involve a significant hazards consideration.

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A marked up copy of the current Ginna Station Technical Specifications which show the requested changes is set forth in Attachment B.The proposed revised Technical Specifications are provided in Attachment C.A marked up copy of Standard Technical Specifications is provided in Attachment D which shows all applicable changes.WHEREFORE, Applicant respectfully requests that Appendix A to Facility Operating License No.DPR-18 be amended in the form attached hereto as Attachment C.Rochester Gas and Electric Corporation By R bert C.Hecredy Vice President Nuclear Production Subscribed and sworn to before me on this 13th day of Hay 1994.Notary Public

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Attachment A R.E.Ginna Nuclear Power Plant License Amendment Request Upgrade of Administrative Controls Technical Specification 6.0 This attachment provides a description of the amendment request and necessary justification for the proposed changes.The attachment is divided into seven sections as follows.Section A identifies all changes to the current Ginna Station Technical Specifications while Section B provides the background and history associated with the changes being requested.

Section C provides detailed justification for the proposed changes including a comparison to Improved Technical Specifications as applicable.

A safety evaluation, significant hazards consideration evaluation, and environmental consideration of the requested changes are provided in Sections D, E, and F, respectively.

Section G lists all references used in this attachment.'.

DESCRIPTION OF AMENDMENT REQUEST This License Amendment Request (LAR)proposes to revise Ginna Station Technical Specification (TS)6.0 as summarized below.Attachment B contains a marked up copy of the current Ginna Station Technical Specifications showing the requested changes.The Technical Specifications following the proposed changes is provided in Attachment C.Note that format issues and minor wording changes are excluded from the discussion below: 2.3.5.Technical Specification Table of Contents i.Revised to support proposed changes.Technical Specification 3.1.4.3.a i.Deleted special report related to exceeding I-131 equivalent activity limits.Technical Specification 3.5.5.3 Deleted"Semiannual" from the name of the Radioactive Effluent Release Report.Technical Specification 3.6 Revised bases to replace reference to TS 6.8 with the Quality Assurance Program.Technical Specification 3.15.1.3 i.Revised reference to special report technical specification section.

Technical Specification 3.16.1 TS 3.16.1.3 is revised to replace references to Table 6.9.2 with the Offsite Dose Calculation Hanual (ODCH).TS 3.16.1.4 is revised to delete"Semiannual" and add"Release" to the name of the Radioactive Effluent Release Report.Technical Specification 6.1 TS 6.1.1 is revised to include a statement that the Plant Hanager shall approve each proposed test, experiment or modification to structures, systems or components that affect nuclear safety.Added a new requirement (TS 6.1.2)which outlines the Shift Supervisor's (SS)responsibilities and who can be designated to assume the control room command function when the SS is absent.Technical Specification 6.2 TS 6.2.l.a is revised to include a requirement for functional descriptions

'f departmental responsibilities and relationships and job descriptions of key personnel positions.

TS 6.2.1.b, 6.2.1.c, 6.2.2.a, 6.2.2.b, 6.2.2.c, 6.2.2.d, 6.2.2.f and Table 6.2-1 are deleted.V.Vl.TS 6.2.2.a is replaced with a statement defining when non-licensed operators are required to be assigned to the shift crew.This requirement currently exists in Table 6.2-1.TS 6.2.2.b is replaced with a statement defining the conditions and durations that operating shift crew compositions may be less than that specified in 10 CFR 50.54(k), (l), and (m), and in the Technical Specifications.

This statement currently exists in Table 6.2-1.TS 6.2.2.e is renumbered and revised to allow the individual qualified in radiation protection procedures to be absent for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.TS 6.2.2.g is renumbered and revised to reference a NRC approved program for controlling overtime.viii.Added a new requirement (new TS 6.2.2.e)which defines the responsibilities and qualifications of the'Shift Technical Advisor (STA).

l,y 9.Technical Specification 6.3 i.TS 6.3.1 is revised to delete reference to the STA function and standardize the reference to Regulatory Guide 1.8.10.Technical Specification 6.4 i.Deleted entire section.ll.Technical Specification 6.5 i.Deleted entire section.'2.

Technical Specification 6.6 i.Deleted entire section.13.Technical Specification 6.7 i.TS Section is now 6.8.ii.TS 6.7.a is revised to update reference to 10 CFR 50 36(c)(1)(i)(A) iii.TS 6.7.b and 6.7.d are revised to replace the reference to.NSARB with"offsite review function" and Senior Vice President, Production and Engineering, with"corporate executive responsible for overall plant nuclear safety." iv.TS 6.7.c is revised to replace reference to PORC with"onsite review function." 14.Technical Specification 6.8 i.TS Section is now 6.4.ii.TS 6.8.l.a is revised to reference Revision 2 of Regulatory Guide 1.33, Appendix A, instead of Revision 0.iii.TS 6.8.l.b, 6.8.l.c, 6.8.1.d, 6.8.l.e, 6.8.l.f, 6.8.1.g, 6.8.l.h, 6.8.l.i, 6.8.2, and 6.8.3 are deleted.iv.TS 6.4.l.b is added which requires written emergency operating procedures which implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1.v.TS 6.4.l.c is added which requires written procedures for all programs specified in new TS 6.5.15.Technical Specification 6.9 i.TS Section is now 6.6.

ii.TS 6.9.1.1, 6.9.1.5, 6.9.2.1, 6.9.2.3, 6.9.2.4, 6.9.2.5, and Table 6.9-2 are deleted.iii.TS 6.9.1.2 is revised to require documentation of all challenges to the pressurizer power operated relief valves or the pressurizer safety valves.iv.TS 6.9.1.3 is revised to delete all discussion related to format and content of the Annual Radiological Environmental Report including deletion of Table 6.9-1.References to 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C are added and the submittal date is changed from Hay 1st of each year to Nay 15th.V.TS 6.9.1.4 is revised to delete all discussion related to format and content of the Radiological Effluent Release Report.References to 10 CFR 50.36a and 10 CFR 50, Appendix I, Sections IV.B.1 are added and the semiannual submittal requirement is changed to an annual report.vi.TS 6.9.2.2 is revised to update the references to 10 CFR 20 and specify a required submittal date.16.Technical Specification 6.10 i.Deleted entire section.17.Technical Specification 6.11 i.Deleted entire section.18.Technical Specification 6.12 i.Removed previously deleted section.19.Technical Specification 6.13 i.TS Section is now 6.7.ii.Revised entire section consistent with new 10 CFR 20.20.Technical Specification 6.14 i.Removed previously deleted section.21.Technical Specification 6.15 Significantly revised entire section and moved the Offsite Dose Calculation Hanual to new TS 6.5.1.22.Technical Specification 6.16 i.Deleted entire section. 23.Technical Specification 6.17 i.Deleted entire section.24.New Technical Specification Section 6.5 i.Added new TS 6.5.1 which defines the Offsite Dose Calculation Manual.ii.Added new TS 6.5.2 which defines the Post Accident Sampling Program.BACKGROUND 1.History On February 6, 1987, the NRC issued an interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors which proposed criteria for defining the scope of what should be contained within technical specifications.

These criteria were developed as a result of efforts within the NRC and nuclear industry to relocate many existing requirements to other more appropriate programs and documents in order to streamline the technical specifications.

The relocated requirements would then be controlled by other existing regulations (e.g., 10 CFR 50.59).This relocation would result in more concise technical specifications which focused on the most important requirements without any reduction in safety.Following issuance of the interim Policy Statement, the NRC and nuclear industry began development of new standard (or improved)technical specifications which implemented the proposed criteria, provided greater emphasis on human factors principles, and contained detailed bases for each retained specification.

The end product of this effort was the release of five NUREGs in September 1992 which contained the Improved Technical Specifications (ITS)for each of the four NSSS vendors (two NUREGs were issued for BWRs).These NUREGs contained all of the improvements discussed above and also so-called"line-item" improvements which are generic refinements to the technical specifications that were identified subsequent to the interim Policy Statement.

The NUREGs were then followed by a final Policy Statement which became effective on July 22, 1993, and encouraged licensees to implement the ITS.In response to the release of the ITS and the continued publication of new"line-item" improvements, RGEE began consideration of implementing NUREG-1431 (Reference 1)which contained the new technical specifications for Westinghouse plants.Following discussions with the NRC staff, RGLE decided to initiate a Technical Specification Improvement Program (TSIP)and implement the ITS for Ginna Station (Reference 2).This program will consist of relocating current requirements in accordance with the final Policy Statement on Technical Specifications and reformatting the remaining requirements consistent with NUREG-1431.

In general, new requirements contained in NUREG-1431 that are not in the current Ginna Station Technical Specifications will not be added.

I Isa Based on discussions with the NRC staff, the first step in the TSIP is conversion of the Administrative Controls section of technical specifications.

This section was selected since RG&E previously submitted a LAR related to the Plant Operations Review Committee (PORC)membership requirements which are contained in the Administrative Controls section (Reference 3).That LAR has not been approved by the NRC and is superseded by this submittal.

The remaining sections of technical specifications will be converted at a later date.Subsequent to the issuance of the NUREGs, the NRC proposed that the content of the ITS Administrative Controls section be significantly reduced to eliminate duplication with other regulatory requirements (Reference 4).The nuclear industry evaluated the proposed changes and formally submitted revisions to the ITS Administrative Controls section based on the information provided by the NRC and additional research and study (Reference 5).RG&E has evaluated the nuclear industry proposed changes and decided to submit a LAR which implements the ITS format for the Administrative Controls section (Ginna Station Technical Specification Section 6)and relocates requirements consistent with Reference 5.Administrative controls that are contained in NUREG-1431 which are related to requirements that will be addressed during the conversion of the remaining Ginna Station Technical Specifications are not included with this LAR.Examples of requirements which are not added at this time include the Safety Function Determination, Explosive Gas and Storage Tank Radioactivity Monitoring, and the Bases Control Programs.In addition, a current requirement for addressing Safety Limit Violations is retained within the Administrative Controls section in order to prevent the need to revise additional Technical Specification sections.The Safety Limit Violation requirement will be relocated to Ginna Station Technical Specification Section 2 during the conversion of the remaining sections.2.Hardware Modifications There are no plant modifications which are required to implement the changes requested in this LAR.Section 6 of the Technical Specifications addresses the administrative controls which are in place to assure that Ginna Station is operated in a safe manner.These controls typically consist of programs and periodic reporting requirements.

Revising the current administrative controls, removing current controls which duplicate requirements contained within the Code of Federal Regulations, or relocating controls to other programs does not require any plant modifications.

Any changes to plant procedures, the Updated Final Safety Analysis Report (UFSAR), or the gA Program which are necessary to support the changes requested by this LAR will also not result in any plant modifications. C.JUSTIFICATION Reformatting the current Administrative Controls section will provide a significant human factor s improvement by locating similar requirements

,-,within the same section and provide a standard structure.

There are several types of changes that are being requested by this LAR in order to perform the reformatting.

The technical and significant administrative changes, and the basis for these changes, are summarized below.Since the upgrade of the Administrative Controls section is being performed consistent with the ITS, a marked up copy of Section 5.0 of NUREG-1431 is provided in Attachment D.This attachment shows all proposed changes with respect to ITS for Westinghouse plants.A cross reference is provided in the left margin of each ITS specification that has been changed by use of a circle containing section numbers from below.Requirements contained within the current Ginna Station Technical Specifications are identified as"TS";within the proposed final Ginna Station Technical Specifications (i.e., Attachment C)as"new TS";and within NUREG-1431 as"ITS." Relocation of Requirements Within Technical Specification Section 6 Many current requirements are moved to support consolidation of similar administrative controls within the same section.Since the requirements are only being relocated within the technical specifications, there is no reduction in safety.The following Ginna Station requirements were relocated within the Administrative Controls Section 6: TS 6.2.l.c-the responsibilities of the Plant Manager are deleted since this is addressed by new TS 6.1.1.TS 6.3.1-the current requirements for Shift Technical Advisors (STAs)are moved to new TS 6.2.2.e.The reference to the RGKE letter dated December 30, 1980, was replaced with a statement requiring the STAs to"meet the qualifications specified within a NRC approved STA training program." This change is different than the wording contained in NUREG-1431 which requires the STA to meet, the qualifications specified in the Commission Policy Statement on Engineering Expertise on Shift (i.e., Generic Letter 86-04).The current STA program at Ginna Station is discussed in References 6 and 7 and has been reviewed and approved by the NRC.Since the STA program does not meet all the requirements contained within GL 86-04, the proposed wording is considered more appropriate and consistent with the current technical specifications.

The deletion of the specific referenceto an RG&E letter eliminates the need to revise technical specifications if the STA program is later revised.

J f,ik t iii.TS 6.4-the requirements for a Training Program are deleted since TS 6.3 specifies that all members of the Ginna Station staff must meet or exceed the minimum requirements of ANSI Standard N18.1-1971.This ANSI Standard requires a retraining and replacement training program which ensures"safe and efficient operation of the facility." Requirements for licensed operator training are addressed by 10 CFR 55.TS 6.8.l.b, 6.8.l.c, and 6.8.l.f-the requirements for written procedures which address refueling operations, surveillance and test activities of safety related equipment, and Fire Protection program implementation are deleted since these are specified in Regulatory Guide 1.33, Appendix A, Revision 2 (new TS 6.4.l.a).V.TS 6.8.l.g-the requirement for written procedures for the radiological environmental monitoring program is deleted since this is addressed in the specification for the Offsite Dose Calculation Hanual (new TS 6.5.1).The Offsite Dose Calculation Hanual is required to contain"the methodology and parameters used in...the conduct of the radiological environmental monitoring program." vi.TS 6.8.l.h-the requirement for written procedures for the Offsite Dose Calculation Hanual is moved to new TS 6.5.1.TS 6.9.1.5 and 6.9.2.4-the requirement for reporting pressurizer power operated relief valve and pressurizer safety valve challenges is moved to the Honthly Operating Report Specification (new TS 6.6.4).viii.TS 6.15-the Offsite Dose Calculation Hanual is revised consistent with NUREG-1431 and ITS Traveller BWOG-09 and moved to new TS 6.5.1.Elimination of Duplicated Regulatory Requirements Several administrative controls currently duplicate other regulatory requirements.

The removal of these type of administrative controls eliminates the need to change technical specifications when there are rule changes.As an example, the recent revisions to 10 CFR 20 will require RG3E to revise the technical specifications.

Since all licensees must meet the applicable requirements contained in the Code of Federal Regulations, or have NRC approved exemptions, there are sufficient regulatory controls in place to allow elimination of duplicated requirements from technical specifications.

The implementation of these requirements are contained in procedures and other licensee controlled documents.

The following requirements were removed from the Administrative Controls Section 6 (or revised as noted)since they duplicate existing regulatory requirements:

TS 3.1.4.3.a and 6.9.2.3-reporting requirements for exceeding the primary coolant specific activity limits listed in TS 3.1.4.l.a and 3.1.4.l.b are specified in 10 CFR 50.73(2)(i)(A).

TS 6.2.2.a, 6.2.2.b, 6.2.2.c, 6.2.2.d, and Table 6.2-1 required operating crew compositions are specified in 10 CFR 50.54(k), (l), and (m)and new TS 6.2.2.a, 6.2.2.b, and 6.2.2.e.TS 6.6-requirements for reportable events are specified in 10 CFR 50.72, 10 CFR 50.73, and other regulations.

TS 6.8.l.d-the requirement for written procedures which implement the Security Plan are specified in 10 CFR 50.54(p)(l) and 10 CFR 73.55(b)(3).

Elimination of this requirement is consistent with Generic Letter 93-07 (Reference 8).TS 6.8.l.e-the requirement for written procedures which implement the Emergency Plan are specified in 10 CFR 50.54(q)and 50.54(t), and 10 CFR 50, Appendix E, Section V.Elimination of this requirement is consistent with Generic Letter 93-07 (Reference 8).TS 6.9.1.3 and Table 6.9-1-the format and content of the Annual Radiological Environmental Report is contained in the Offsite Dose Calculation Hanual specification (new TS 6.5.1)and the referenced Radiological Assessment Branch Technical Position, Revision 1, November 1979.References to 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C were added to support deletion of the report format and content and the submittal date was changed from Hay 1st of each year to Hay 15th consistent with NUREG-1431.

TS 6.9.1.4-the format and content of the Radiological

'.Effluent Release Repor t is contained in Offsite Dose Calculation Hanual specification (new TS 6.5.1).References to 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1 were added consistent with NUREG-1431.

The semiannual submittal requirement is also changed to an annual report in accordance with 10 CFR 50.36a(a)(2).

TS 6.9.2.1-reporting requirements related to sealed sources are specified in 10 CFR 30.50.TS 6.9.2.5-licensee submittals are addressed by 10 CFR 50.4.TS 6.11-the requirement for a Radiation Protection Program is specified in 10 CFR 20.1101.

E y~I'\ xi.TS 6.17-changes to the Radioactive Waste Treatment systems are controlled by 10 CFR 50.59.NRC notification of significant changes to these systems is addressed by 10 CFR 50.59(b)(2).

3.Relocation of Current Administrative Control Requirements To Other Controlled Documents The relocation of certain requirements contained in the Administrative Controls section to other licensee controlled documents (i.e., gA Program and plant procedures), does not eliminate the program.Instead, the requirements are relocated to other more appropriate documents and programs which have sufficient controls in place to manage implementation and future changes (e.g., 10 CFR 50.54(a)(3) and 10 CFR 50.59).The relocation of these items will enable RGKE to more efficiently maintain the requirements under existing regulations and reduce the need to request technical specification changes for issues which do not affect public safety.The following sections were relocated from the Administrative Controls Section 6 to the gA Program or plant procedures:

TS 3.16.1.3 and Table 6.9-2-Reporting levels for radioactivity concentrations in environmental samples will be addressed in and implemented by procedure RD-15.0, Offsite Dose Calculation Hanual (Reference 9).Relocating the reporting levels to the Offsite Dose Calculation Manual is consistent with the required content of this manual as specified in new TS 6.5.1.Changes to the Offsite Dose Calculation Hanual is also addressed by new TS 6.5.1 which provides sufficient control with respect to maintaining appropriate reporting levels for radioactivity concentrations.

TS 6.2.1.b, 6.7.b and 6.7.d-the responsibilities of the Senior Vice President, Production and Engineering.

New TS 6.2.I.a requires that"lines of authority, responsibility and communication shall be established and defined throughout the highest management levels." Specifying management positions by title within the technical specifications creates a licensee and NRC burden when titles are changed or responsibilities are shifted.The responsibilities of key organization positions are specified in the Ginna Station gA Program.Since changes to the gA Program are controlled by 10 CFR 50.54(a)(3), equivalent control is provided.TS 6.2.1;d-the capability of training, health physics and quality assurance to have direct access to responsible corporate management to support mitigation of their concerns.New TS 6.2.l.a requires that"lines of authority, responsibility and communication shall be established and defined throughout the highest management levels." The organizational structure is specified in the Ginna Station gA Program.Since changes to the gA Program are controlled by 10 CFR 50.54(a)(3), equivalent control is provided. 1V~TS 6.5, 6.6.l.b, 6.7.b, 6.7.c and 6.7.d-the Review and Audit functions, including the compositions, meeting frequencies, and responsibilities of the PORC, NSARB and gA.The necessary changes to the gA Program to support relocation are being requested separately.

Changes to the gA Program are controlled by 10 CFR 50.54(a)(3) which provides equivalent control.v.V1~TS 6.8.l.i and 6.16-The Process Control Program is addressed in and implemented by procedure RD-16.0, Process Control Program (Reference 10).This program ensures compliance with 10 CFR 20, 10 CFR 61, and 10 CFR 71 which regulate the processing and packaging of solid radioactive wastes.The failure to meet these regulations is sufficiently addressed within the Code of Federal Regulations.

Since the Process Control Program only ensures compliance with existing regulations, it is acceptable to relocate this program to plant procedures which are under the control of 10 CFR 50.59.TS 6.8.2 and 6.8.3 and Bases for TS 3.6-control of changes to written procedures and the use and approval of temporary procedure changes.The necessary changes to the gA Program to support relocation are being requested separately.

Changes to the gA Program are controlled by 10 CFR 50.54(a)(3) which provides equivalent control.TS 6.10-Record Retention requirements.

These are currently specified in the gA Program.Changes to the gA Program are controlled by 10 CFR 50.54(a)(3) which provides equivalent control.Addition of New ITS Administrative Control Requirements There are several requirements contained in NUREG-1431 which are not currently in the Ginna Station Technical Specifications.

The following NUREG-1431 (ITS)requirements were added to the current Ginna Station Administrative Controls section in order to provide a more complete specification:

ITS 5.1.1 (TS 6.1.1)-Plant Manager responsibilities related to approving each proposed test, experiment or modifications to systems, structures or components that affect nuclear safety.ITS 5.1.2 (TS 6.1.2)-Shift Supervisor responsibilities and allowed absences from the control room.1V.ITS 5.2.l.a (TS 6.2.l.a)-a requirement for ,functional descriptions of departmental responsibilities and relationships and job descriptions of key personnel positions.

ITS 5.2.l.c (TS 6.2.2.e)-allows the individual qualified in radiation protection procedures to be absent for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

'I II v.ITS 5.7.l.l.a (TS 6.4.l.a)-requires written procedures in accordance with Revision 2 of Regulatory Guide 1.33, Appendix A, instead of the current reference to Revision 0.vi.ITS 5.7.l.l.b (TS 6.4.l.b)-requires written emergency operating procedures which implement the requirements of NUREG-0737 and NUREG-0737, Supplement l.vii.ITS 5.7.l.l.g (TS 6.4.l.c)-requires written procedures for all programs specified in TS 6.5.r viii.ITS 5.7.2.6 (TS 6.5.2)-requires a Post Accident Sampling Pro'gram.ITS Administrative Controls Deferred Until Conversion of Remaining TS Sections Several ITS Administrative Control requirements were not added since they will be addressed during the conversion of.the remaining technical specifications sections.Their addition is delayed since they are currently addressed elsewhere within the Ginna Station Technical Specifications or are directly related to the conversion of the remaining sections.The following NUREG-1431 administrative control requirements (or issues)will be evaluated during the conversion of the remaining chapters: The use of Modes 1 through 6.This was replaced with the current nomenclature of Refueling,'old Shutdown, Hot Shutdown, and Operating (see TS 1.2).The numbering of the Administrative Controls section remains 6.0 (versus 5.0)until the consolidation of Technical Specifications Sections 3.0 and 4.0.ill~1V.ITS 5.6, Technical Specifications Bases Control-this program will be implemented following conversion of the remaining chapters.ITS 5.7.1.1.e, gual i ty Assurance for Effluent and Environmental Monitoring

-the requirements for effluent and environmental monitoring are addressed by TS 3.16, Radiological Environmental Monitoring, and TS 4.10, Radiological Environmental Monitoring Surveillances.

The Ginna Station Technical Specifications currently require (and the gA Program will require)independent audits at least once per year to ensure"the conformance of facility operation to all provisions contained within Technical Specifications." Therefore, sufficient quality assurance of these activities exists.v~" ITS 5.7.2.4, Primary Coolant Sources Outside Containment

-the control of leakage from systems that could contain highly radioactive fluids and which exit containment is addressed by TS 4.4.3, Recirculation Heat Removal Systems. V1.V11.ITS 5.7.2.5, In Plant Radiation Monitoring Program-the capability to determine airborne iodine concentrations in vital areas under accident conditions is addressed by TS 3.5.4 and Table 4.1-5.ITS 5.7.2.7, Radioactive Effluent Controls Program-the requirement to provide limitations on liquid and gaseous releases is addressed by TS 3.9, Plant Effluents, and TS 4.12, Effluent Surveillance.

V111~1X.ITS 5.7.2.8, Radiological Environmental Monitoring Program-the requirement to monitor, sample, and report radiation and radionuclides in the environment is addressed by TS 3.16, Radiological Environmental Monitoring, and TS 4.10, Radiological Environmental Monitoring Surveillances.

ITS 5.7.2.10, Pre-Stressed Concrete Containment Tendon Surveillance Program and ITS 5.9.2.d, Special Report related to abnormal degradation observed by the Pre-Stressed Containment Tendon Surveillance Program-requirements related to containment tendon degradation are addressed by TS 4.4.4, Tendon Stress Surveillance.

X.X1~X11.X111.X1 V.ITS 5.7.2.11, Inservice Inspection Program-requirements for an inservice inspection program are addressed by TS 4.2, Inservice Inspection.

ITS 5.7.2.12, Inservice Testing Program-requirements for an inservice testing program are addressed by TS 4.2.1.6.ITS 5.7.2.13, Steam Generator Tube Surveillance Program and ITS 5.9.2.e, Special Report for Steam Generator Tube Surveillance Program-requirements for a steam generator tube surveillance program are addressed by TS 4.2.1 and 4.2.1.4.ITS 5.7.2.15, Ventilation Filter Testing Program-testing requirements for Engineered Safety Feature filter ventilation systems is addressed by TS 4.5.2.3, Air Filtration System.ITS 5.7.2.16, Explosive Gas and Storage Tank Radioactivity Monitoring Program-the control and monitoring of potentially explosive gas mixtures in radioactive systems is addressed by TS 3.5.5 and 3.9.2.5.XV.XV1.ITS 5.7.2.17, Diesel Fuel Oil Testing Program-the control, monitoring, and sampling requirements of diesel fuel oil is addressed by TS 4.6.I.d.ITS 5.8.1, Safety Function Determination Program-this program will be implemented following conversion of the remaining chapters.

ITS 5.9.1.6, Core Operating Limits Report-this report is only required following removal of existing technical specification core operating limits and will be utilized following conversion of the remaining sections.xviii.ITS 5.9.1.7, Reactor Coolant System Pressure and Temperature Limits Report-this report is only required following removal'of existing technical specification RCS pressure and temperature limits and will be utilized following conversion of the remaining sections.ITS 5.9.2.b, Special Report following four or more valid failures of an individual emergency diesel generators in the last 25 demands-This requirement is currently not specified in the electric power sections of Technical Specifications and will be addressed during their conversion.

XX.ITS 5.9.2.c, Special Report related to Post Accident Monitoring (PAM)Instrumentation

-the requirement for a special report is currently specified in TS Table 3.5-6.ITS Administrative Controls Not Added The following NUREG-1431 administrative control requirements were not added to the technical specifications based on ITS Traveller BWOG-09 or issues specific to Ginna Station: ITS 5.1.1 (TS 6.1.1)-the requirement for Plant Manager approval of proposed tests, experiments or modifications to be contained in approved administrative procedures was not added.This would only duplicate a technical specification requirement within procedures and does not provide any benefit to safety.Also, Plant Manager approval of"proposed changes" was deleted since his approval is already required"prior to implementation" which provides sufficient control of unit changes.ITS 5.1.2 (TS 6.1.2)-the requirement for an annual management directive to all station personnel related to the shift supervisor control room command function was not added.A management directive is not necessary since the authority and responsibility of each unit organization is specified in the gA Program and addressed by TS 6.2.l.a.ITS 5.2.2.f-the requirement for the Operations Manager to hold an SRO license was not added since the qualifications of this position are addressed in ANSI Standard N18.1-1971 (see TS 6.3). v.V1.ITS 5.3.1 (TS 6.3)-a requirement for qualifications of staff not covered by Regulatory Guide 1.8 was not added.This requirement was not considered necessary since all activities which affect nuclear safety are controlled by other technical specification requirements, existing regulations, and the gA Program.Also, Revision 1 of Regulatory Guide 1.8 was not revised to Revision 2 in order to maintain consistency with the current gA Program and existing procedures.

ITS 5.7.2.9-the requirement for a Component Cyclic or Transient Limit Program was not added.The tracking of cyclic and transient occurrences within a specified program to ensure that components are maintained within the design limits is not considered necessary due to existing programs such as surveillance testing and the ISI program.Maintaining records of transient and operational cycles is currently required by TS 6.10.g and the gA Program.ITS 5.7.2.14-the requirement for controlling and monitoring secondary water chemistry was not added since this requirement is specified as a condition to the Ginna Station Operating License dated December 10, 1984 (see Reference 12).ITS 5.7.2.18-the requirement for a Fire Protection Program were not added consistent with BWOG-09.The administrative control for this program was originally added to ensure that the unit was capable of providing alternate/dedicated safe shutdown in accordance with 10 CFR 50, Appendix R following a fire.Changes to the Fire Protection Program may be made without prior NRC approval only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.This requirement is specified as a condition to the Ginna Station Operating License dated December 10, 1984 (see Reference 13).Changes to plant procedures that implement the Fire Protection Program are addressed under 10 CFR 50.59.ITS 5.9.2.a-the requirement for a special report following the actuation and injection of an Emergency Core Cooling System (ECCS)above Node 4 was not added consistent with BWOG-09.Reporting requirements related to ECCS actuations are addressed by 10 CFR 50.73(a)(2)(iv) which specifies a Licensee Event Report within 30 days.Other Changes to Technical Specification Section 6 The following revisions to the Administrative Controls Section 6.0 are minor changes only and do not involve any technical issues: Reformat of the entire Section 6.0 consistent with the structure of NUREG-1431.

r Renumbering of remaining Section 6.0 requirements consistent with the format of NUREG-1431 and ITS Traveller BWOG-09.1V~v.V1.Update of references to sections of the Code of Federal Regulations.

Standardization of report names within Technical Specifications.

Removal of previously deleted sections within Technical Specifications.

ITS 5.l.1 (new TS 6.1.1)-the wording was revised to require the Plant Hanager to approve changes to"structures, systems or components" that affect nuclear safety consistent with Ginna Station terminology.

vii.ITS 5.1.2 (new TS 6.1.2)-replaced"valid" license with"active" license consistent with ITS Traveller BWOG-09 and standard nomenclature at Ginna Station.1X.X.ITS 5.2.2.e-revised the overtime requirement for unit staff who perform safety related functions to require control in accordance with an NRC approved program.RG&E currently utilizes a staff working hour control program which slightly differs from the NRC Policy Statement on Working Hours (Generic Letter 88-12).This program was previously reviewed and approved by the NRC (Reference ll).The proposed wording is considered more appropriate and consistent with the current technical specification.

TS 6.9.1.1 (ITS 5.9.1.1)-deleted the requirement since a Startup Report is more appropriately addressed in the NRC Safety Evaluation Report authorizing an Operating License, increased licensed power level, installation of a new nuclear fuel design or manufacturer, or modifications which significantly alter the nuclear, thermal, or hydraulic performance of the plant.The'Startup Report is required to be submitted within 90 days following completion of the above activities and does not require NRC approval.Therefore, inclusion of the requirement for this report in Technical Specifications is not necessary to assure safe plant operation.

ITS 5.9.1.2.a (new TS 6.6.1)-A submittal date of"on or before April 30 of each year" was added to the Occupational Radiation Exposure Report consistent with 10 CFR 20.2206(7)(c). xi.ITS 5.11 (new TS 6.7)-replaced references to"health physics" with"radiation protection" throughout section.Also replaced the">1000 mrem/hr" with"Z 1000 mrem/hr" in ITS 5.11.3 to make it consistent with ITS 5.11.2.In addition, deleted the requirement for the Radiation Protection Manager to sign Radiation Work Permits (RWPs)as specified in ITS 5.11.I.a.The issuance of RWPs is addressed in procedures under the control of 10 CFR 50.59.Specifying management positions by title within the technical specifications creates a licensee and NRC burden when titles are changed or responsibilities are shifted.8.Retained Administrative Control Requirements Not In ITS Ginna Station Technical Specification 6.7 lists the required actions in the event that a Safety Limit is violated.This requirement will be relocated to Technical Specification 2.0 during the conversion of the remaining sections to eliminate the need to revise additional technical specifications if the requirement were to be relocated now.SAFETY EVALUATION The changes requested in this LAR involve either 1)the relocation of existing requirements within the technical specifications or other RGSE controlled documents, 2)elimination of specifications duplicated by regulatory requirements, 3)addition of new controls in accordance with NUREG-1431, or 4)administrative changes only.These changes are consistent with the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors and NUREG-1431.

Therefore, the safety impact of making these changes has been previously considered and found acceptable by the NRC.The Administrative Controls section was also converted to the ITS format;however, not all NUREG-1431 requirements were implemented.

These additional requirements were not added since they are currently addressed elsewhere within the Ginna Station Technical Specifications, are directly related to the conversion of the remaining sections, or are not considered necessary to provide a complete specification.

The impact of not implementing all of the NUREG-1431 requirements was evaluated as follows: There is no increase in the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR related to not implementing all of the NUREG-1431 administrative controls.The requirements which were not added as identified in section C.5 and C.6 above are currently addressed by existing technical specifications and regulations, the Ginna Station license, plant procedures, or the gA Program.Therefore, there is no increase in the probability or consequences of an accident or malfunction of equipment by not adding these requirements to the Technical Specifications. The possibility of a new or different kind of accident or a malfunction of a different type than previously evaluated in the UFSAR is not created by not implementing all of the NUREG-1431 administrative controls.The Administrative Controls section contains those requirements that are not covered by other technical specifications which are considered necessary to assure safe operation of the facility.These controls typically supplement existing regulations and do not affect the accident analyses contained in the UFSAR.Since the affected requirements identified in section C.5 and C.6 above are currently addressed by existing technical specifications and regulations, the Ginna Station license, plant procedures, or the gA Program, a new or different kind of accident or malfunction is not created by not adding them to the Technical Specifications.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification related to not implementing all of the NUREG-1431 administrative controls since these requirements are currently not contained in the Ginna Station Technical Specifications.

Based on the above, the proposed amendment does not involve an unreviewed safety question and will not adversely affect or endanger the health and safety of the general public.SIGNIFICANT HAZARDS CONSIDERATION EVALUATION The proposed changes to the Ginna Station Technical Specifications do not involve a significant hazards consideration as discussed below: Operation of Ginna Station in accordance with the proposed changes does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The changes are consistent ,with the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors and NUREG-1431 and have therefore, been previously evaluated by the NRC.Implementation of these changes is expected to result in a significant human factors improvement and enable RGEE and the NRC to focus on the most important requirements without any reduction in safety.The changes which do not duplicate NRC guidance in NUREG-1431 are currently addressed by existing technical specifications and regulations, the Ginna Station license, plant procedures, or the gA Program.

II ,g 4 C e 044 2.3.Operation of Ginna Station in accordance with the proposed changes does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The Administrative Controls section contains those requirements that are not covered by other technic'al specifications which are considered necessary to assure safe operation of the facility.The majority of changes are consistent with the Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors and NUREG-1431 and have therefore, been previously evaluated by the NRC.The requirements which were not added only supplement existing regulations and do not affect the accident analyses contained in the UFSAR.Operation of Ginna Station in accordance with the proposed changes does not involve a significant reduction in, a margin of safety.All requirements removed from technical specifications are relocated to other programs and documents.

These alternative programs and documents are controlled by existing'regulations which provide a more appropriate vehicle for addressing changes and compliance.

There were no administrative control requirements which were removed from technical specifications and not addressed by other regulations.

Therefore, there is no significant reduction in a margin of safety.Based upon the above information, it has been determined that the proposed changes to the Ginna Station Technical Specifications do not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident previously evaluated, and does not involve a significant reduction in a margin of safety.Therefore, it is concluded that the proposed changes meet the requirements of 10 CFR 50.92(c)and do not involve a significant hazards consideration.

ENVIRONMENTAL CONSIDERATION RG8E has evaluated the proposed changes and determined that: The changes do not involve a significant hazards consideration as documented in Section E above;2.The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite since all administrative controls related to offsite releases are retained or addressed by existing regulations; and 3.The changes do not involve a significant increase in individual or cumulative occupational radiation exposure since all administrative controls related to maintaining'adiation exposure as low as reasonably achievable are retained or addressed by existing regulations.

~4 I Accordingly, the proposed changes meet the el igibi 1 i ty cri teri a for categorical exclusion set forth in 10 CFR 51.22(c)(9).Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed changes is not required.REFERENCES 2.NUREG-1431, Standard Technical Specifications, kfestinghouse Plants, September 1993.Letter from R.C.Hecredy, RG&E, to A.R.Johnson, NRC,

Subject:

Conversion to Improved Technical Specifications, dated February 28, 1994.3.4, 5.6.7.8.9 10.12.13.Letter from R.C.Mecredy, RG&E, to A.R.Johnson, NRC,

Subject:

Application for Amendment to Technical Specification 6.5.1,"Plant Operations Review Committee (PORC)Function," dated March 23, 1993.Letter from W.T.Russell, NRC, to Improved Technical Specification Owners Group Chairpersons,

Subject:

Content of Standard Technical Specifications, Section 5.0, Administrative Controls, dated October 25, 1993.Improved Technical Specification Traveller BWOG-09.Letter from R.W.Kober, RG&E, to H.Fairtile, NRC,

Subject:

Policy Statement on Engineering Expertise on Shift (Generic Letter 86-04), dated Hay 14, 1986.Letter from R.C Hecredy, RG&E, to A.R.Johnson, NRC,

Subject:

NUREG-0737, Item 1.A.1.1, Shift Technical Advisor, dated October 12, 1989.Generic Letter 93-07,

Subject:

Hodification of the Technical Specification Administrative Control Requirements for Emergency Security Plans.Ginna Stat'ion Procedure RD-15.0, Offsite Dose Calculation Hanual.Ginna Station Procedure RD-16.0, Process Control Program.Letter from W.A.Paulson, NRC, to R.W.Kober, RG&E,

Subject:

Plant Staff irking Hours and Reporting Requirements for Safety Valve and Relief Valve Failures and Challenges, dated January 31, 1984.Letter from D.H.Crutchfield, NRC, to L.D.White, RG&E,

Subject:

Issuance of Amendment No.33 to Provisional Operating License No.DPR-18, dated June 13, 1980.Letter from A.R.Johnson, NRC, to R.C.Mecredy, RG&E,

Subject:

Issuance of Amendment No.40 to Facility Operating License No.DPR-18, R.E.Ginna Nuclear Power Plant (TAC No.N83568), dated September 21, 1992.