ML20197J251

Application for Amend to License NPF-3.Amend Changes TS Definitions 1.12,TS 3/4.9.5 & New TS 3.0.6

ML20197J251
Person / Time
Site:	Davis Besse
Issue date:	12/23/1997
From:	Jeffery Wood CENTERIOR ENERGY
To:
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ML20197J241	List: ML20197J236 ML20197J251 ML20197J262
References
NUDOCS 9801020121
Download: ML20197J251 (14)

Text

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Dock 0t Number-50-346-

-Lic0n00 Number NPF-3

=Scrici Number 3462 Enclosure

~Page 1 >

APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 ,

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station,

. Unit Number 1 Facility Operating License Number NPF-3. Also included is the

-Safety Assessment and Significant Hazards Consideration.

The proposed changes (submitted under cover letter Serial Number 2462) concern:

Appendix Ai Technical Specifications: -

Pages VIII and XII Index 1.12 Definitions - Core Alteration

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J.0.6 Applicability - Limiting Condition for Operation 3/4.9.5 Refueling Operations - Communications Appendix A, Technical Specification Bases:

3.0.6 Applicability 3/4.9.5 Refueling Operations - Communications By: i ,

J.fWo[d,VicePresident-Nuclear Sworn to and subscribed before me this 23rd day-of December, 1997 nuJ %f>&

Nodary Public,-State of Ohio LAURA A.JEP#450N i Notary PuWic, State ed Olwo My Commish Empires t@2001 9901020121 971223 PDR ADOCK 05000346

- P POR

I

= Dock 0t Number 50-346 Lic0n03 Number NPF-3 S0rici Number 2462 Enclosure Page 2 The following information is provided to support issuance of the requested-changes to the Davis-Besse Nuclear Power Station (DBNPS), Unit Number 1, Facility Operating License Number NPF-3, Appendix A, Technical Specifications.

The changes involve Technical Specification-(TS; Definition 1.12 - Core Alteration, TS 3.0.6 - Applicability - Limiting condition for Operation and-associated Basea, and TS 3/4.9.5 - Refueling Operations - Communications and associated Bases.

A. Time Required to Implement: These' changes will be implemented within 90 days after the NRC issuance of the License Amendment.

B. Reason for Change (License Amendment Request 97-0002):

This application proposes to revise the TS 1.12 Core Alteration definition to clarify what constitutes a core alteration.

The proposed changes will also. add new TS 3.0.6 and Bases, which addresses the return of equipment to service to perform testing required to demonstrate operability. Having the means to return equipment to operable status from its tripped condition can help prevent an unnecessary plant transient or shutdown.

1 In addition, this application proposes to relocate TS 3/4.9.5 - Refueling Operations - Communications and Bases from the Technical Specificationt to the Updated Safety Analysis Report Technical Requirements Manual. Future changes to the relocated specification will be evaluated under Section 50.59 of Title 10 of the Code of Federal Regulations.

These changes are in accordance with NUREG-1430, " Standard Technical Specifications, Babcock and Wilcox Plants," Revision 1, dated April 1995, as modified by a pending NUREG-1430 change approved by the NRC, Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler Number 165.

C. Safety Assessment and Significant Hazards Consideration: See Attachment.

-Dock;t Number 50-346 '

Lic;n;0_ Number NPF-3  ;

Serial' Nwrbr 2462 Attachment- ,

h SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR a LICENSE AMENDMENT. REQUEST NO. 97-0002 (45 pages follow)

.s b

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i LAR 97-0002 Page 1 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDEPJsTION FOR LICENSE AMENDMENT REQUEST NUMBER 97-0002 TITLE:

Proposed Modification to the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A, Technical Specifications to Revise the Technical Specification (TS) Definition of Core Alteration; Relocate TS 3/4.9.5 -

Refueling Operations - Communications and its Bases to the Technical Requirements Manual; and Add TS 3.0.6 and Associated Bases Regarding the Return to Ser/ ice of Inoperable TS Equipment.

DESCRIPTION:

The purpose of this License Amendment Request is to modify the Davis-Besse Nuclear Power Station (DBNPS) Operating License NPP-3, Appendix A, Technical Specifications (TS) by revising TS 1.12 - Core Alteration, to adopt the language of NUREG-1430, Revision 1, "Standacd Technical Specifications, Babcock and Wilcox Plants," dated April 1995, which more. clearly states what constitutes a core alteration. In addition, this request proposco the relocation of TS _,

3/4,9.5 - Refueling Operations - Communicar. ions and its Bases to the DBNPS Updated Safety Analysis Report (USAR) Technical Requirements Manual (Tmi), and the addition of TS 3.0.6 - Applicability - Limiting condition for Operation and associated Bases, addressing the return to service of inoperable equipment.

Proposed TS 3.0.6 and Bases adopt the text of NUREG-1430 TS 3.0.5 and Bases, as modified by a pending NUREG-1430 change approved by the NRC, Technical Spacification Task Force (TSTF) Standard Technical Specification Change Traveler Number 165. Each of these changes is discussed in more detail as follows:

TS 1.12 - CORE ALTERATION Revision Technical Specification 1.12 pres;ntly states:

CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

The proposed revised definition would read as follows:

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel wit'n the vessel

-head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The main purpose of the proposed change is to clarify the definition by more specifically describing the type of components of interest to include " fuel, sources, or reactivity control components." In addition, as noted above, the proposed revised definition is in verbatim agreement with NUREG-1430.

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i LAR 97-0002 Page 2 Felocation of TS 3/4.9.5 - Pefueling Ormra ti ons - Communicationn TS 3/4.9.5 - Refueling Operations - Communications and associated Bases 3/4.9.5 - Communications, are being proposed for relocation from the TS to the USAR TRM with the same content as they possessed as part of the Operating License. This proposed change is unrelated to the above-mentioned proposed changes. This relocation is in accordance with Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR) and the Nuclear g Regulhtory Commission's " Final Policy Statement on TS Improvements for

' Nuclear Power Reactors," dated July 22, 1993, which provides NRC policy on what is required to be included in the Technical Specifications and gives guidance on relocation of TS requirements to other station documents. The NRC policy estabiiches that any TS requirements which do not moet any of the four delineated criteria may be proposed for relocation to licensee controlled documents such as the USAR. The requirement for communications during refueling does not meet any of these criteria, and is, therefore, proposed for relocation. This revision will allow the DBNPS to evaluate future proposed changes to the communications requirements under Section 50.59, " Changes, Tests, and Experiments," of Title 10 of the Code of Federal Regulations (10 CFR), and implement those changes which are not an unroviewed safety question or do not involve a change to the TS. This relocation is also consistent with NUREG-1430. Associated changes to the TS Index to note the deletien of TS 3/4.9.5 and associated Bases 3/4.9.5 are also proposed.

IS 3.0,6 and Banen Additi2D The proposed TS 3.0.6 would allow " Equipment removed from service or declared inoperable to comply with ACTIONS . . . [to} be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITT of other equipment. " This proposed change is unrelated to the above-mentioned proposed changes. It has been the NRC's long-standing position that this practice is acceptable, as reaf firmed in the NRC's letter to Niagara Mohawk Power Corporation, dated November 21, 1996, regarding the Nine Mile Point Nuclear Station, Unit 2.

This change is in accordance with NUREG-1430. Having~ the means to return equipment to operable status f rom its tripped condition can help prevent an unnecessary plant transient or shutdown.

The new TS 3.L.6 would state:

Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. Tais is an excepcion to Specification 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY, y Associated with this change, a new Bases 3.0.6 would be added. This change is in accordance with NUREG-1430, as modified by a pending NUREG-1430 change

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LAR.97-0002 Pago 3 approved.by the NRC, Technical Specification Task Force (TSTF) Standard

, Technical Specification Change Traveler Humber 165. The new Bases would read as.follows:

Specification 3.0.6 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to on cification 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of required testing to demonstrates

a. The OPERABILITY of the equipment being returned to service; or

b. The OPERABILITY of other equipment.

The administrative controls cnsure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY. This Specification does not provide time to perform any other preventive or corrective maintenance.

An example of demonstrating the OPERAB7LITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions, and must be reopened to perform the required testing.

An example of demonstrating the OPERABILITY of other equipment being returned to service is taking an inoperable channel or trip system out of the tripped condition tr prevent the trip function from occurring during the performance of .vguired testir.g on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of

-the tripped condition to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system.

Each of the proposed revisions are shown on the attached marked-up Operating License pages.

SYSTEMS. Q,QMPONENTS , AND ACTIVITIES AFFECTED:

The proposed changes clarify the core alteration definition, thereby clarifying the need to apply various Technical. Specification requirements during

< performance of refueling outage activities. No systems or components are affected by this change.

The proposed changes also relocate the Technical Specification and Bases for 3/4.9.5 - Refueling Operations - Communications to the USAR TRM. No systems, components, or activities are'affected by this change.

(, . -

LAR 97-0002 Page 4 l

The proposed changes to add new TS 3.0.G and associated Bases to address the

)

return to service of inoperable equipment is an administrative change and does J not affect any systems, components, or activities. The activities described by the proposed new TS 3.0.6 and-associated Bases reflect current practice at the DBNPS.

FUNCTIONS OF THE AFFECTED SYSTFMS. COMPO[LE_NTS . AND ACTIVITI,SE1 The core alteration da'inition identifies activities in the reactor pressure vessel that could pocentially cause an unwanted reactivity change. Other Technical Specifications establish requirements which must be met when conduct-ing these activities, such an establishing containment closure (or the ability to isolate, if necessary), and reactivity contro) requirements. These actions ensure personnel safety and minimize radiation exposure risk to the public, The communications requirements during core alte2ations ensure that refueling station personnel can be promptly informed of significant changes in the station status.or core reactivity condition.

The addition of new TS 3.0.6 and associated Bases addresses the return to service of inoperable equ, , ment.

EEEECTS ON SAFETY:

TS 1.12 - CORE ALTERATION Revision The Tochnical Specifications require that certain conditions be met during core alterations to ensure adequate reactivity monitoring, reactivity control, and the ability to mitigate the effects of a fuel handling accident. The proposed revised core alteration definition will continue to ensure that appropriate Technical Specification requirements are applied 'or activities that have the potential to affect reactivity.

In order to br1ng the revised definition into agreement with NUREG-1430, it is also proposed to' delete the word " conservative" from the TS statement "Suspen-sion of CORE ALTERATIONS shall not preclude completion of movement of a compo-nent to a safe conservative position. " The meaning of the word " conservative" is similar to che word " safe" in this statement. This change is administrative.

Action statements for TS 3/4.1.2.1 - Boration Systems - Flow Paths - Shutdown, TS 3/4.1.2.5 - Decay Heat Removal Pump - Shutdown, TS 3/4.1.2.6 - Boric Acid Pump - Shutdown, TS 3/4.1.2.8 - Borated Water Sources - Shutdown, and'TS 3/4.9.1

- Refueling Operations - Boron Concentretion require the suspension of core alterations or positive reactivity changes. These requirements ensure that activities are suspended when sufficient reactivity controls may not exist to keep the reactor core shutdown by the required amount in the event of a fuel assembly or control component being mishandled. Ths ravised definition will continue to ensure that these requirements are met "hea

• he potential for a loss of required shutdown mhrgin exists because activities which could significantly alter core reactivity will continue to be identified as core alterations.

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~LAR-97-0002 Page 5 TS 3/4.9.2 - Refueling Operations - Instrumentation specifies requirements for reactor core neutron flux monitoring during refueling. These requirements ensure that sufficient reactivity monitoring is available to indicate changes in the core reactivity condition. The Action statement requires that core l

alterations or positive reactivity changes be suspended when sufficient monitoring instrumentation is not available. This requirement prevents the occurrence or an undetected, inadvertent change in reactivity condition. In addition, the surveillance requir?ments of this TS require performance of a Channel Functional Test within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initial start of core alterations, a Channel Punctional Test at least once per 7 days during core alterations, and a Channel Cb9ck at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during core alterations. These requirements ensure detector operability during core alterations or positive reactivity changes, which provides increased protection against an undetected increase in core reactivity. The revised core alteration definition will continue to identify activities which could result in signifi-

-cant reactivity changes. Therefore, neutron flux monitors will continue to be required during. activities in which significant reactivity changes could occur.

TS 3/4.8.1.2 - Electrical Power Systems - A.C. Sources - Shutdown ensures that electrical power sources are available to ensure that the station can be maintained in the shutdown or refueling condition for extended time periods and that sufficient instrumentation and control capability is available for monitoring and maintaining the station status. The Action statement requires the suspension of core alterations or positive reactivity changes with less than the minimum required A. C. electrical power sources operable. The revised core alteration definition will continue to identify the movement of reactivity related items as core alterations.

The requiremen'r. of TS 3/4.9.4 - Refueling Operations - Containment Penetrations ensure that the .elease of fissic. product radioactivity to the environment due to a fuel element rupture during shutdown is mi;.imized. The TS requires that paths from containment to the outside atmosphere be closed or capable of being closed during core alterations or movement of irradiated fuel within containment. The revised core alteration definition will continue to identify movement of fuel assemblies within the reactor vessel as core alterations. In addition, the DBNPS has established administrative controls to ensure the safe and reliable handling of heavy loads. As documented in the October 29, 1984 NRC Safety Evaluation Report (SER), " Control of Heavy Loads - Phase I," the DBNPS conforms to the general provisions for load handling specified in Section 5.1.1 of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." As summarized in the SER, these general provisions for load handling are met via development and implementation, through procedures and operator training, of safe load travel paths such that, to the maximum extent practical, heavy loads are not

• carried over or near irradiated fuel or safe shutdown equipment. As further stated in the SER, these gedbral provisions for load handling a a also met by providing. sufficient operator training, handling system design, load handling instructions, and equipment inspection to ensure reliable operation of the handling system.

The Action statement for TS 3/4.9.5 - Refueling Operations - Communications requires the suspension of core alterations when direct communications between the. control room and personnel at the refueling station are not availah'.e. The communicatione requirements during core alterations ensure that refueling

LAR 97-0002

.Page 6 station personnel can be promptly informed of significant changes in the station

-status or core reactivity condition. The revised core alteration definition will continue to identify activities which could result in significant reactivity changes. Therefore, after relocation to the USAR TRM, the communications capability will continue to be required during activities in which significant reactivity changes could occur. As described previously, unrelated to the core alteration definition change, TS 3/4.9.5 and associated Bases are proposed to be relocated to the USAR TRM.

-Based on the above, the proposed change to TS 1.12 will have no adverse effect on plant safety.

Relocation of TS 3/4.9.5 - Refuelino Doera,tions - Communications 10 CFR Section 50.36 establishes the regulatory requirements for licensees to include TS as part of applications for operating licenses. In addition, the Nuclear Regulatory Commission's " Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," dated July 22, 1993, provides guidance regarding the required content of TS. The fundamental purpose of the TS, as described in the NRC's Final Policy Statement, is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. This is accomplished by identifying those features that are of controlling importance to nuclear safety and establishing on them certain TS conditions of operation which cannot be changed without prior NRC approval.

The NRC's Final Policy Statement recognized, as had previous statements related to the NRC Staff's TS Improvement Program, that implementation of the policy would result in the relocation of some existing TS requirements to licensee-controlled documents such as the USAR. Those items relocated to the USAR would, in turn, be controlled in accordance with the requirements of 10 CFR Section 50.59 10 CFR Section 50.59 provides criteria to determine when facility or operating changes planned by a licensee require prior NRC approval in the form of a license amendment in order to address any unreviewed safety questions or TS changes.

The four criteria of 10 CFR 50.36 that call for retaining a particular item in the TS ares (1.) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation.of t.he reactor coolant pressure boundary.

As described in the-Federal Register notice (

Reference:

58 FR 39132) of the NRC's Final Policy Statement, the purpose of thir criterion is to ensure that TS control those instruments spe'ifically installed to detect excessive Reactor Coolant System leakage. The Federal Register states that this criterion should not be interpreted to include

. instrumentation to detect precursors to reactor coolant pressure boundary leakage oz instrumentation to identify the actual leakage.

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LAR 97-0002

-page 7 (2.) A process variable,- design feature, or operating restriction that is

.an initial condition of a Design Basis Accident or Transient analysis that either assumes the failur,e cf or presents a_ challenge to the integrity of a fission product barrier.

As described in the Final policy Statement's Federal Register notice, '

the purpose of this criterion is to capture thoae process variables that havs initial values assumed in the Design Basis Accident and transient analyses, and which are monitored and controlled during .

power operation. As long as these variables are maintained within the '

established values, risk to the public cafety is presumed to be acceptably low. This criterion also includes active design features (e.g., high pressure / low pressure system valves and interlocks) and operating restrictions (e.g. , pressure / temperature limits) needed to preclude unanalyzed accidents and transients.

(3.)- A structure, system, or component that is part of the primary success

• path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. '

As described in the Final policy Statement's Federal Register notice, the purpose of this criterion is to capture only those structures, systems, and components that are part of the primary success path of a safety sequence analysis. Aleo captured by this criterion, are those support and netuation systems that are necessary for items in the primary success path to successfully function. The primary success path for a particular mode of operation does not include backup and  ;

diverse equipment (e.g., safety valves which are a backup to low temperature overpressure relief valves during cold shutdown).

(4.) A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As described in the Final policy Statement's Federal Register notice, the purpose of this criterion is to retain in TS those requirements that the probabilistic safety assessment or cperating experience shows

• as significant to public health and safety.

Technical Specification 3/4.9.5 does not meet these criteria for inclusion in tha Technical Spe:ifications, as discussed in further detail below:

criterion 1- The communications system is not installed instrumentation that is used to detect degradation of the reactor coolant pressure boundary. This communications system is used to promptly notify station personnel of significant changes in the facility status or core reactivity conditions during core alterations.

Criterion 2 Communications capability is not a process variable, design

. feature, or operating restriction that is an initial condition of a Design Basis Accidtat or Transient analysis that either assumes the

{

. LAR 97-0002 Page 8 failure of or presents a challenge to the integrity of a fission product barrier. Communications capability is not a process variable, design feature, or operating restriction that was an initial condition assumed in the analysis of any accident described in the DBNPS USAR.

Criterien 3 The communications system does not provide a function or actuate in order to mitigate the consequences of a Design Basis Accident or Transient. Communications between the refueling station personnel and the control room is not a mitigating factor in any accident described in the DBNPS USAR.

Criterion 4 Communications capability has not been shown to be significant to public health and safety by either operational experience or Probabilistic Risk Assessment. Communications were not identified in the DBNPS Probabilistic Risk Assessment (i.e., the Individual Plant Examination (IPE)) as being significant to public health and ,

safety.

In summary, the proposed relocation of TS 3/4.9.5 to the USAR TPM will not impact the requirements regarding communications capability. Further, t;he proposed change will not affect testing requirements for the communications system. Any future changes to the relocated specification will be evaluated as required by 10 CFR 50.59. It is concluded that there is no adverse effect en plant safety as a result of this relocation.

TS 3.0.6 and Bases Addition Several TS Action statements require placing equipment that has been declared inoperable in a " tripped" condition. For example, with a containment isolation valve inoperable due to a slow stroke timo, under Limiting condition for Operation (LCO) 3.6.3.1 Action b, the isolation valve may be required to be deactivated in the closed position. In order to return the valve to operable status following maintenance, it would be necessary to stroke the valve. As ,

another example, with an instrumentation channel bistable found to be out of i adjustment during a monthly surveillance test, the functional unit would be declared inoperable anc placed in the tripped condition in accordance with the applicable TS Action statement. In order to return the functional unit to operable statun, the bistable would need to be untripped to determine and, potentially, adjust its precise setting and ensure its proper operation. The addition of Specification 3.0.6 would address both of these concerns by allowing

" Equipment removed from service or declared inoperable to conply with ACTIONS

. . . (to) be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment."

By returnina a repaired instrumentation channel to its normal operating condition, the likelihood of a spurious reactor trip or inadvertent actuation of safety relate? equipment is greatly reduced. For example. Action 10 of TS Table 3.3-1, Reactor Protection System (RPS) Instrumentation, requires that if two of four Reactor Coolant High Temperature channels are inoperable, one be placed in a tripped condition and one be placed in bypass. In this condition if one of l

LAR 97-0002-Page 9 th' two remaining operable RPS channels failed, RPS would _ trip the reactor.

With '.he _ return of the repaired, tripped channel to service under the proposed TS 3.0.6 in-order to test it, two of the three inservice channels would have to fail in order to cause an-inadvertent reactor trip.

Returning equipment to its untripped condition to verify operability may temporarily place the plant in a condition whereby, in the unlikely event of an accident occurring while the ecmponent is in an untripped condition, it could potentially fail to perform its safety related function, requiring both of the remaining operable channels, in the above example, to function properly to generate an RPS trip. However, the likelihood is small of an accident occurring during the time the channel is removed from the tripped condition since administrative controls ensure that the time the channel is returned to service is limited to only the time needed to perform the testing. Furthermore, the channel would be returned to service for testing only if it was believed the repaired channel was now operable. As stated in the proposed Bases 3.0.6, no other preventive or corrective maintenance-will be performed on the equipment which has been returned to service under these circumstances.

As previously referenced, the proposed change is in accordance with recent NRC guidance and is in agreement w!ch NUREG 1430, as modified by a pending NUREG-1430 change approved by the NRC, TSTF Standard Technical Specificat ion Change Traveler Number 165. The dministrative controls on time allowed to test the operability of equipment returned to service will limit the time that the plant is in this condition. The incro ssed safety risk involved during operability testing is more than offset by the decreased risk of an undesirable transient or shutdown. Therefore, the addition of LCO 3.0.6 and its Bases will not involve an adverse effect on plant safety.

SIGNIFICANT HAZARDS CONSIDERATION:

The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92 (c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changas wouldt (1) Not involve a significant increase in the probability or consequences of an accident previously evaluatedi (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the proposed changes and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station,. Unit Number 1, in accordance with these changes would la. Not involve a significant increase in the probability of an accident previously evaluated because the probability of previously analyzed accidents is not affected by the criteria in the core alteration definition (Technical' Specification (TS).1.12). Nor do these changes, the proposed relocation of the refueling communications TS 3/4.9.5 and Bases to the DBNPS Updated Safety Analysis Report-(USAR) Technical Requirements Manual (TRM),

or the proposed addition of new TS 3.0.6 and Bases regarding return to service of inoperable equipment, affect _any accident initiator, or

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LAR 97-0002 Page 10 assumption made in any safety analysis. The proposed changes are administrative inLnature and are consistent with NUPP" 1430, Revision 1,

" Standard Technical Specifications, Babcock and Wilcox Plants," dated April, 1995, as modified by a pending NUREG 1430 change approved by the NRC, Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler Number 165.

Ib. Not involve a significant increase in the consequences of an accidan' previously evaluated because the proposed changes do not affect c ..t conditions or assumptions used in evaluating the radiological cont . .aences of an accident. The proposed changes do not significantly -iter the source term, containment isolation, or allowable radiological re eases.

2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed changes do not change the way the plant is operated. No new or different types of failures or accident initiators are introduced by the proposed changes.

3. Not involve a significant reduction in a margin of safety because no inputs into the calculation of any Technical Specification Safety Limit, Limiting Safety System Settings, Technical Specification Limiting Condition for Operation, or other previously defined margins for any structure, system, or component important to safety are being affected by the proposed changes.

CONCLUSIONt On the basis of the above, the Da'ris-Besse Nuclear Power Station has determined that the License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Speczfications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.

ATTACHMElfr:

Attached are the proposed marked-up changes for the operating License.

REFERENCES:

1. Davis-Besse Nuclear Power Station (DBNPS) Unit No. 1, Operating License NPF-3, Appendix A. Technical Specifications, through Amendment 215.

2. " Standard Technical Specifications, Babcock and Wilcox Plants," NUREG-1430, Revision 1, dated April 1995.

3. Technical Specification Task Force (TSTF) Standard Technical Specification

' Change Traveler Number 165.

4. USAR Section 9.5.2, " Communications Systems," through Revision 20.

-5. USAR Section 15.4.7, "Puel-Handling Accident," through Revision 20.

'LAR 97-0002 Page 11 6.: 10 CPR 50.36, " Technical Specifications.*

7. 30 CFR 50.59,." Changes, Tests, and Experiments."

8. The NRC " Final Policy Statement on Technical specifications Improvements for Nuclear Power Reactors," (58 FR 39132, dated July 22, 1993).

9. Letter from S. Singh Bajwa, NRC, to Ralph Sylvia, Niagara Mohawk Power Corporation, regarding Technical Specification 3.0.2, Nine Mile Point Nuclear Station, dated November 21, 1996.

• 10 NUREG-0612,- " Control of Heavy Loads at Nuclear Power Plants, Resolution of Generic Technical Activity A-36," published July 1980.

11. NRC letter dated October 29, 1984 (Toledo Edison Log Number 1634) .

..w 4