ML20210T216

Draft SE Supporting Proposed Conversion of Current TS to ITS for Plant

ML20210T216
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Site:	Farley
Issue date:	08/06/1999
From:	NRC (Affiliation Not Assigned)
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ML20210T210	List: ML20210T202 ML20210T216
References
NUDOCS 9908190035
Download: ML20210T216 (325)

Text

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[". +i UNITED STATES o,s di j e

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION e 1 RELATED TO AMENDMENT NOS. AND TO FACillTY p 3 OPERATING LICENSES NPF-2 AND NPF-8 i

% l JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 AND UNIT 2 p y SOUTHERN NUCLEAR OPERATING COMPANY p z DOCKET NUMBERS! 50-348150-364

l. INTRODUCTION _

,l l-Joseph M. Farley Nuclear Plant Unit 1 and Unit 2 (FNP) has been operating with Technical Specifications (TS) issued with original operating licenses on June 25,1997, foi Unit 1 and March 31,1981, for Unit 2, as amended from tim 3 to timef By letter dated March 12,1998,as supplemented by letters dated April 24,1998, August 20,1998, November 20,1998, February 20,1999, April 30,1999, May 28,1999, June 30,1999, and July 27,1999, Southern Nuclear Operating Company (SNC or the licensee) proposed to amend Appendix A of Operating Licenses No. NPF-2 and NPF- 8 to completely revise the FNP TS. The proposed amendment was based upon NUREG-1431, " Standard Technical Specifications - Westinghouse Plants,"

Revision 1, dated April 1995, and upon guidance in the "NRC Final Policy Statement on Technical Specification improvements for Nuclear Power Reactors" (Final Policy Statement),

published on July 22,1993 (58 FR 39132). The overall objective of the conversion, consistent with the Final Policy Statement, is to rewrite, reformat, and streamline the TS for FNP to be in accordance with 10 CFR 50.36.

Hereinafter, the proposed TS are referred to as the improved TS (ITS), the existing FNP TS are referred to as the current TS (CTS), and the TS in NUREG-1431 are referred to as the standard TS (STS). The corresponding TS Bases are ITS Bases, CTS Bases, and STS Bases, respectively.

In addition to basing its ITS on STS and the Final Policy Statement, SNC retained portions of the CTS as a basis for the ITS. Plant-specific issues, including design features, requirements, and operating practices, were discussed with the SNC during a series of conference calls and meetings that concluded on April 20,1999. In addition, SNC proposed matters of a generic nature that were not in STS. The NRC staff requested that SNC submit such generic issues as a proposed change to STS through the Nuclear Energy institute's Technical Specifications Task Force (TSTF). These generic issues were considered for specific applications in the FNP ITS.

Consistent with the Final Policy Statement, SNC proposed transferring some CTS requirements to SNC-controlled documents. In addition, human factors principles were emphasized to add clarity to the CTS requirements being retained in the ITS and to define more clearly the Enclosure 1 9908190035 990806 PDR ADOCK 05000348 p PDR

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'v appropriate scope of the ITS Further, significant changes were proposed to the CTS Bases to make each ITS requirement clearer and easier to understand. f8. ,

y The Commission's proposed action on the FNP application fian amen'dment deled March 12,1998, was published in the FEDERAL REGISTER on Mar 25,1999 (64 FR 28218).

The Staff's evaluation of the application and supplementaldiformation thit res

.thisrequests Safety Evaluation for information (SE). These and plant-specific changes discussions epys ticiarify the ITS with respect with isiuseenhedin to SN the guidance in the Final Policy Statement and STS. . Therdue,f9te changes are within the scope of the action described in the initial and supplemental FEDERAL REGISTER notices.

y 9,a During its review, the NRC staff relied on the Final Poli y^Statemenih,b STS as guidance for acceptance of CTS changes. This SE provides a basis hir[Ste'NRC staff conclusion that FNP can develop ITS based on STS,lais uteSed by pladt specific changes, and that the use of the ITS is acceptable for continued operallon.EThe NRC staff also acknowledges that, as indicated in the Final Policy Statement, the conversion to STS is a voluntary process.

Therefore, it is acceptable that the .lTS differs from STS, reflecNng the current licensing basis for FNP. The NRC staff approves SNC's slidnges to the CTS with modifications documented in the revised submittals. [#

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p8 in , A; For the reasor,s stated infra in this SE, the NRC ataff finds that the TS issued with this license amendment comply with Sedian %2a of the Atomic Energy Act,10 CFR 50.36, and the guidance in Die Final Policy Statement, and that they are in accord with the common defense and security snd provide adeqdate protection of the health and safety of the public.

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11. BACKGROUND !  %

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Section 182a of the AtorrNe Energy Act requires that applicants for nuclear power plant operating licenses will state: h,'

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.(S)uch technical spealllcations, including information of the amount, kind, and source of 2 apecial nuclear material required, the place of the use, the specific characteristics of the

[ facility, and such other information as the Commission may, oy rule or regulation, deem U nocessary in order to enable it to find that the utilization . . . of special nuclear material will The in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. in dok g so, the Commission placed emphasis on those matters related to the prevention

- of accidents and the mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS "those items that ar6 directly related to maintaining the integrity of the physical barriers designed to contain radioactivity." Statement of Consideration, " Technical Specifications for Facility Licenses; Safety Analysis Reports,"

33 FR 18610 (December 17,1968). Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories: (1) cafety limits, limiting safety system settings and

. limiting control settings; (2) limiting conditions for cperation (LCOs); (3) surveillance

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requirements (SRs); (4) design features; and (5) administrative controls; However, the rule

. does not specify the particular requirements to be included in a plant's TS.i 4 '

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For several years, NRC and industry representatives have sothhi to desalop guidelines for e improving the content and quality of nuclear power plant T On February 6 Commission issued an interim policy statement on TS impp(svemen NAny_ Statement on Technical Specification improvements for Nuclear Power Reactors" (52 FR 3788).' During the period from 1989 to 1992, the utility Owners Groups aisil tis DStC staff developed improved

' standard technical specifications that would establish modils of ths' Commission's policy for each primary reactor type. In addition, the NRC staff, licensees land Owners Groups developed l

generic administrative and editorial guidelines in the form 5f a "W4ers Ginde" for preparing TS, L which gives greater consideration to human factors principles and Esas'uesd throughout the l development of licensee-specific ITS. gQg W7 ftwQ gnyx &

In September 1992, the Commission issued NUREG-1431lwhich_was developed using the guidance and criteria contained in the Comadssion's inierinipogey" statement. STS were established as a model for developing ITS for Westirighouse M in general. STS reflect the l results of a detailed review of the appliostion of the interim policy statement criteria to generic l system functions, which were published in a " spa Report" isisued to the Nuclear Steam System l Supplier (NSSS) Owners Groups in May 1988ASTS also reflect the results of extensive i discussions concerning various~ drafts of STSiso;that#ie application of the TS criteria and the

j. Writer's Guide would consisten8y reRect detailed system configurations and operating characteristics for au NSSS W.j As such, the generic Bases presented in NUREG-1431 provide an stiundance of inforrhation regarding the extent to which the STS present l

requirements that are necessary to protect public health and safety.

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!> On July 22,1993, the Commiselon issued its Final Policy Statement, expressing the view that j satisfying the" guidance in the policy statement also satisfies Section 182a of the Act and 1 10 CFR 50.36 (58 FR 39132)Alho Final Policy Statement described the safety benefits of the I

. improved STS, and encouraged licensees to use the improved STS as the basis for plant-

, specific TS amendments, and for complete conversions to improved STS. Further, the Final L {

Policy Statement gave guidance for evaluating the required scope of the TS and defined the i guidance criteria to be used in determining which of the LCOs and associated surveillances l

should remem in the TS. The Commission noted that, in allowing certain items to be relocated to licermeo controlled documents while requiring that other items be retained in the TS, it was l adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263,273 (1979).

t There, the Appeal Board observed:

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[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is i legally binding upon the licensee unless and until changed with specific Commission l- approval. Rather, as best we can discem it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed 1

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I necessary to obviate the possibility of an abnormal situation or event giving rise to an

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immediate threat to the public health and safety. -

By this approach, existing LCO requirements that fall within or satisfy any of the enteria in the l Final Policy Statement st.ould be retained in the TS; those LCO requirerpents that do not fall within or satisfy these criteria may be relocated to licenseehntrolled dodume'nts?Tho' Commission codified the four criteria in 10 CFR 50.36 (60 FR 36593,' July 19,1995). The Final Policy Statement criteria are as follows: k,%  %

Criterion 1 [y

,p y Installed instrumentation that is used to detect, and indicate ih the control room, a

, significant abnormal degradation of the reactor coolant pressure boundary.

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! Criterion 2 I.

f: % j l 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 failure of or presents a challenge to the integrity of a fission product barrier, r s l '. Y'

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l Criterion 3 '

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A structure,' system, or component that is'part of the primary success path and which functions or actuates to mitgate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

l Criterion 4 l

A structure, system, or component which operating experience or probabilistic safety l assessment has shown to be significant to public health and safety. l Part til of this SE explains the NRC staff conclusion that the conversion of the FNP CTS to those l based on STS; as modsfied by plant-specific changes, is consistent with the FNP current

! licerming basis and the requirements and guidance of the Final Policy Statement and l 10 CFR 50.36.

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l Ill. EVALUATION l

The NRC staff's ITS review evaluates changes to CTS that fall into five categories defined by SNC and includes an evaluation of whether existing regulatory requirements are adequate for controlling future changes to requirements removed from the CTS and placed in SNC-controlled documents.

In addition to the initial submittal of March 12,1998, as supplemented, the NRC staff review identified the need for clarifications and additions to the submittal in order to establish an appropriate regulatory basis for translation of CTS requirements into ITS. Each change proposed in the amendment request is identified as either a discussion of change (DOC) to CTS l

a or a justification for deviation from STS. The NRC staff comments were documented as j requests for additional information (RAls) and forwarded to SNC. SNC provided written i

responses to the NRC staff requests in supplementalletters indicated aboves The docketed l letters clarified and revised SNC's basis for translating CTS requirements into ITS. ,The NRC l staff finds that SNC's submittels provide sufficient detail to abow the staff to reach a conclusion I regarding the adequacy of SNC's proposed changes. 2 [ NQ f; , #

! Q f C l ~ The license amendment application was organized such that changes were included in each of l the following CTS change categories, as appropriate; administridive changes, technical changes - less restrictive (specific), technical changes - less reelrkslue (generic), technical changes - more restrictive, and relocated specifications: (f "" p N a l'ek.

(1) Administrative Changes, (A), i.e., non-technical changes in t guesentation of existing l

l requirements.

n fh {M' f.i l (2) ' Technical Changes - More Restrictive, (M), i.e$%giv &ineweraddhional CTS require

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-(3) Technical Changes - Less Restrictive (specific), (L), le., changes, deletions and relaxations of existing TS requirementsii f f; $ l (4) Technical Changes - Lsss Restrictive','(LA), i.ei.; deletion of existing TS requirements by movement of information and requirements from existing specifications (that are otherwise being retained) to SNC-controlled documents, including TS Bases.

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(5) Technic'al Changes - Less Restrichve, (LB), i.e., relaxation of existing TS requirements by providing an allowance to use' a simulated or actual signal to verify the automatic actuation of specific components in the surveillance test requirements of the TS.

(6) ~ Relocated Specifications, (R), i.e., relaxations in which whole specifications (the LCO i Jand associated action and SR) are removed from the existing TS (an NRC-controlled

- document) and placed in SNC-controlled documents.

A These general categories of changes to SNC's CTS requirements and STS differences may be

better understood as follows: 4 yx  ; l A. Administrative Changes Administrative (non-technical) changes are intended to incorporate human factors principles into the form and structure of the ITS so that plant operations personnel can use them more easily.

These changes are editorial in nature or involve the reorganization or reformatting of CTS requirements without affecting technical content or operational restrictions. Every section of the ITS reflects this type of change in order to ensure consistency, the NRC staff and SNC have used STS as guidance to reformat and make other administrative changes. Among the changes proposed by SNC and found acceptable by the NRC staff are:

(1) Providing the appropriate numbers, etc., for STS bracketed information (information that i must be supplied on a plant-specific basis and that may change from plant to plant). l 4

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e A (2) Identifying plant-specific wording for system names, etc. 54 y y,

(3) ~~ Changing the wording of specification titles in STS to' conform t{Ieeletindplant practices. g Q 2 S [%gSN J, l- (4) Splitting up requirements currently grouped unde s' single current specificadon to more appropriate locations in two or more specificationdef_ITSP  %&

'Qp;i (5) Combining related requirements currently'presentedirisoperate specifications of the CTS into a single specification of ITS. f g

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n %i&S,W Table A lists the administrative changes proposed inIT8CTable A is osgenised by the corresponding ITS section DOC, and provides a,sunumery description of the administrative changa that was made, and CTS and ITS LCo references.tThe NRC staff reviewed all of the

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administrative and editorial changes proposed by SNC and Ands Wiern acceptable, because they are compatible with the Writers Guide and STS, do not result in any' substantive change in l . operating requirements and are consistint with tha' Commission's regulations.

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~ 8. Technical Changes -yp More Restrictive p- A ' w/

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SNC, in electing to implement the specifications of STS proposed a number of requirements more restridive than those in tio CTSn ITS re$irements in this category include requirements that are either now, more conservative than corresponding requirements in the CTS, or that have additional restrictions that are not in the CTS but are in STS. Examples of more restrictive requirements are placing en LCO on plant equipment which is not required by the CTS to be operable, more restrictive requusments to restore inoperable equipment, and more restrictive SRs.- Table M lists all the' more restrictive changes proposed in ITS. Table M is organized by the corresponding ITS section DOC and provides a summary description of the more restrictive change that was adopted, and CTS and ITS LCO references. These changes are additional restrictions on plant operation that enhance safety and are acceptable.

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s C. Technical Changes L Less Restrictive (Specific)

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Less feetricWve requirements include changes, deletions and relaxations to portions of CTS requirements that are not being retained in ITS or relocated to a SNC-controlled document.  !

When requirements have been shown to give little or no safety benefit, their change or removal from the TS may be appropriate. In most cases, relaxations previously granted to individual j plants on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3)

. resolution of the Owners Groups comments on STS. The NRC staff reviewed generic relaxations contained in the STS and found them acceptable because they are consistent with current licensing practices and the Commission's regulations. The FNP design was also reviewed to determine if the specific design basis and licensing basis are consistent with the technical basis for the model requirements in STS, and thus provide a basis for ITS . '

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1 A significant number of changes to the CTS involved changes, deletions and relaxations to portions of CTS requirements evaluated as Categories I through Vil that follow Category 1 - Ralaxation of Modes of Applicability /? Njg fx , _

Category II - Relaxation of Surveillance Frequency [

p[ \pw Category 111 - Relaxation of Completion Time 4 -

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% g-Category IV - Relaxation of Required Actions gM ;,

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Category V - Relaxation of Surveillance Requirement Acceptance Criteria Category VI - Relaxation of LCO f.l  %'

l#~"&',2 . l Category Vil - Deletion of SR  ?  :

l Category Vlli - Deletion of Requirembnt for 30 day Special Report to NRC l: , j j?

The following discussions address why various'TS within each of the eight categories of  ;

l information or specific requiremients are not required to be included in ITS .

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Relaxation of Modes of A_ -(Category l}

Reactor operating conditions are used in CTS to define when the LCO features are required i to be operable. CTS apphcabilities can be specific defined terms of reactor conditions: hot shutdown, cold shutdown, reactor critical or power operating condition. Applicabilities can also be more general. Depend ng on the circumstances, CTS may require that the LCO be maintained within limits in "all modes" or "any operating mode." Generalized applicability i

conditions are not contained in STS, therefore ITS eliminate CTS requirements such as "all modes" or "any operating mode," replacing them with ITS defined modes or applicable conditions that are consistent with the application of the plant safety analysis assumptions for operability of the rpquired features.

l$ another appl tion of this type of change, CTS requirements may be eliminated during condihons for which the safety function of the specified safety system is met because the feature is performing its intended safety function. Deleting applicability requirements that are indeterminate or which are inconsistent with application of accident analyses assumptions is acceptable because when LCOs cannot be met, the TS are satisfied by exiting the applicability thus taking the plant wt of the conditions that require the safety system to be operable. These changes are consistent wih STS and changes specified as Category I are acceptable.

Relaxation of Surveillance Freauency (Category ll)

CTS and ITS surveillance frequencies specify time interval requirements for performing surveillance requirement testing, increasing the time interval between surveillance tests in

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the ITS results in decreased equipment unavailability due to test which also increases equipment availability. In general, the STS contain test frequencies that are consistent with industry practice or industry standards for achieving acceptable levels of equipment reliability.

Adopting testing practices specified in the STS is acceptable based on similar design, like-component testing for the system application and the availability of otherTS requirements

, which vrovide regular checks to ensure limits are met. (

Reduced testing can result in a safety enhancement thh unavailability due to test is i

reduced; in turn, reliability of the affected structure, system or component should remain constant. Reduced testing is acceptable where operating experience, industry practice or the industry standards such as manufacturers' recommendstions hous shown that these

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components usually pass the Surveillance when performed at the spoollied interval. Thus, the frequency is acceptable from a reliability standpoint.; Surveillanos frequency changes to incorporate alternate train testing have been shown to be acceptable ~where other qualitative or quantitative test requirements are required which are antaMahed predictors of system performance, e.g., a 31 day air flow test is an indicator that posibve pressure in a controlled space will be maintained because this test would use the s'ame fans as the less frequent ITS 36 month pressurization test and industry experience shows that components usually pass j the pressurization test. Additionally, surveillance frequency extension can be based on staff- i approved topical reports. The NRC staff has accepted topical report changes where topical l report analyses bound the plant-specific design and component reliability assumptions.

i These changes are consistent with STS and changes specified as Category 11 are acceptable.

1 Relaxation of Comoletion Time (Category lli)

Upon discovery of a failure to meet an LCO, STS specify times for completing required actions of the associated TS conditions. Required actions of the associated conditions are l used to establish remedial measures that must be taken within specified completion times j l (allowed outage times). These times define limits during which operation in a degraded condition is permitted.

l Adopting completion times from the STS is acceptable because completion times take into l account the operability status of the redundant systems of TS required features, the capacity l and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a design basis accident (DBA) occurring during the repair period. These changes are consistent with STS and allowed outage time extensions specified as Category lli are acceptable.

Relaxatic of Reauired Actions (Category IV)

CTS require that in the event specified LCOs are not met, penalty factors to reactor operation, such as resetting setpoints, and power reductions shall be initiated as the method to reestablish the appropriate limits. The iTS are constructed to specify actions for conditions of required features made inoperable. Adopting ITS action requirements for exiting LCO applicabilities is acceptable because the plant remains within analyzed parameters by performance of required actions, or the actions are constructed to minimize risks associated

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l with continued operation while providing time to repair inoperable features. Such actions add margin to safety or verify equipment status such as interlock status for the mode of operation, thereby providing assurance that the plant is configured appropriately or operations that could l

result in a challenge to safety systems are exited in a time period that is commensurate with the safety importance of the system. Additionally, other changes to TS actions include placing the reactor in a Mode where the specification no'ionger appii5s, usdaEy resutting in an extension to the time period for taking the plant into shutdown conditions. These actions

are commensurate with industry standards for reductions in thermal power in an ' orderly fashion without compromising safe operation of the plantc These changes are consistent with STS and changes specified as Category IV are acceptable.g . ,,

4 4' Relaxation of Surveillance Reauirement Acceptance Criteria (Category V)

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CTS require safety systems to be tested and vertfied operable prior to" entering applicable conditions. ITS provide the additional requaement to verify operabildy by actual or test conditions. Adopting the STS allowance for " actual" coriditions is ' acceptable because TS required features cannot distinguish between an "octual" signal or a " test" signal. Category V also includes changes to CTS requirements that are replaced in the ITS with separate and distinct testing requirements which when combined include operability verification of all TS required components for the features specified in the CTS. Adopting this format preference in the STS is acceptable because TS SRs that remain include testing of all previous features required to be verified operable. CTS provide an a!!owance to bypass an inoperable channel for surveillance testing of other channels. ITS provide the allowance to bypass the inoperable channel when making required setpoint adjustments on the other channels as well as performing surveillance tests on other channels. CTS test extensions allow inoperable channels to be bypassed for surveillance testing when sufficient equipment is required to be operable by TS requirements to provide an acceptable level of safety system protection. SR relaxations include the recognition in ITS that administrative controls exist which provide assurance that any changes to component status, such as valve position, are recorded and tracked. Thus, ITS extend the option to verify penetration integrity by administrative control to valves outside containment, whereas CTS permits this option only for valves inside containment. These changes are consistent with STS and changes specified as Category V  !

are acceptable.

L- . . . 3 Relaxation of LCO (Category VI) l CTS provides lists of acceptable devices that may be used to satisfy LCO requirements. The l ITS reflect the STS approach to provide LCO requirements that specify the protective limit l inat is required to meet safety analysis assumptions for required features. The protective limits replace the lists of specific devices previously found to be acceptable to the NRC staff for meeting the LCO. The ITS changes provide the same degree of protection required by the safety analysis and provide flexibility for meeting limits without adversely affecting operations rance equivalent features are required to be operable. These changes are consistent with STS and changes specified as Category VI are acceptable.

q Deletion of SR (Category Vil)

Both CTS and ITS include applicability requirements which specify that failure to meet an SR or failure to perform an SR within the specified time interval constitutes a failure to meet I operability requirements for an LCO. As an adjunct to the TS conversion process, CTS SRs are reviewed to establish an appropriate level of testing for LCO regulremehts retained in the ITS. One outcome of this review is a determination that it is appropriate to make changes to CTS SRs. CTS SRs can be deleted as a result of adopting ITS format; such as eliminating i detector testing for components that are not susceptible to drift and then simplifying surveillance test by revised testing criteria. CTS SRs can be replaced with a like-test that

. verifies operability of components but at a less frequent' test interval because the conditions required for testing make it safe to reduce testing since other information is available to ensure components are operable. CTS may also contain specific requirements to perform testing which verifies a criterion of a component design.: Explicit com'ponent verification is subject to TS requirements to establish component operability, thus TS testing is simplified in the ITS by eliminating such narrowly focused test criteriaQSurvedlance frequency requirements for components may be, revised to correspond to industry standards resulting in SR interval extensions. Relaxations to SR can be made by deleting the requirement to perform a SR test for a class of componentscFor components whose status can be adequately be controlled by STS administrative means, options add flexibility to testing where determinations to include components in a test scheme can be evaluated based on the status of the component during specified plant conditions. Some CTS surveillances are changed to a' low testing or repairs on redundant components which as a result temporarily eliminates protection afforded by the redundant component. Category Vll changes include deletion or modification of CTS surveillance testing requirements not needed to establish equipment operability. 1 Deletion of Reauirement for 30-day Soecial Report to NRC (Category Vill)

CTS include requirements to submit Special Reports when specified limits are not met.

Typically, the time period for the report to be issued is within 30 days. However, the STS eliminates the TS administrative control requirements for Special Reports and instead relies on the reporting requirements of 10 CFR 50.73. ITS changes to reporting requirements are acceptable because 10 CFR 50.73 provides adequate reporting requirements, and the special reports do not affect continued plant operation, Therefore, this change has no impact on the safe operation of the plant. Additionally, deletion of TS reporting requirements reduces the administrative burden on the plant and allows efforts to be concentrated on restoring TS required limits. These changes are consistent with STS and changes specified as Category Vill are acceptable.

Table L lists all CTS requirements that have been deleted and which pertain to Category I through IX and to the specific listing of changes discussed above. Table L includes all LB l changes and is organized by ITS section which specifies: the section designation followed by the DOC identifier, e.g.,1.1 L.1 (ITS Section 1.1, DOC L.1); a summary description of the change; CTS and ITS LCO references; a reference to the specific change category as discussed above (if applicable); and a characterization of the DOC.

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For the reasons presented above, these less restrictive requirements are acceptable because they will not affect the safe operation of the pisnt. The TS requirements that remain are consistent ~with current licensing practices, operating experience, plant accident and transient analyses, and provide reasonable assurance that public health and safety will be protected.

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D. Relocated Less Restrictive Requirements y g ( -, .

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When requirements have been shown to give little or no salsty benefit, their removal from the TS may be appropriate. In most cases, relaxations previously granted to individual plants on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the Owners Groups comments on STS. The NRC staff reviewed gene'ricJelaxations contained in STS and found them acceptable because they are, consistent with current licensing practices and the Commission's regulations. The FNP design was also reviewed,to determine if the specific design tesis and licensing basis are consistent with the technical basis for the model requirements in S TS, and thus provide a basis for ITS/ A significant number of changes to the CTS involved *.nts removal of specific requirements and detailed information from individual specifications evaluated to be Types 1 through 4 that follow: E

{ 4 Type 1 Details of system desigriand system description including design limits i

Type 2 Descriptions of systems operation Type 3 Procedural details for requirements and related reporting problems  !

l Type 4 Administrative requirements redundant to regulations l The following discursions address why each of the four types of information or specific requirements are not required to be included in ITS .

Details of System Jesian and System Description includina Desian Limits (Type 1) l The design of th s facility is required to be described in the UFSAR by 10 CFR 50.34. In addition, the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50 require that plant desig a be documented in controlled procedures and drawings, and maintained in accordance wita an NRC-approved QA plan (UFSAR Chapter 17). In 10 CFR 50.59 controls are specified for changing the facility as described in the UFSAR, and in 10 CFR 50.54(a) criteria are specitad for changing the QA plan. In ITS, the Bases also contain descriptions of system design. IT3 5.5.14 specifies controls for changing the Bases. Removing details of system design from the CTS is acceptable because this information will be adequately controlled in the UF! AR, controlled design documents and drawings, or the TS Bases, as appropriate. Cycle apecific design limits are moved from the CTS to the Core Operating Limits Report 'COLR)in accordance with Generic Letter (GL) 88-16. ITS Administrative Controls are revised to include the programmatic requirements for the COLR.

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Descriotions of Systems Operation (Type 2)

The plans for the normal and emergency operation of the facility are r$ quired to be described l in the UFSAR by 10 CFR 50.34. ITS 5.4.1.a requires written procedures to be established, l implemented, and maintained for plant operating proced,uires including procedures recommended in Regulatory Guide (RG) 1.33, Revisiorf2,' Appendix A', Fetiruary 1978.

Controls specified in 10 CFR 50.59 apply to changes in procedures as described in the UFSAR. In ITS, the Bases also contain descriptions of' system operation. It is acceptable to remove details of system operation from the TS becauss this type of information will be adequately controlled in the UFSAR, plant operating procedures, and the TS Bases, as appropriate. 6 %g e, w . ..

Procedural Details for Meetino TS Reauirements & Related Reportina Problems (Type 3)

a Details for performing action and SRs are more appropnately'specified in the plant i procedures required by ITS 5.4.1, the UFSAR, and ITS Bases. For example, control of the plant conditions appropriate to perform a surveillance test is an issue for procedures and scheduling and has previously been determined to be unnecessary as a TS restriction. As indicated in G

. 91-04, allowing this procedural control is consistent with the vast majority of l other SRs that do not dictate plant conditions for surveillances. Prescriptive procedural  !

information in an action requirement is unlikely to contain all procedural considerations necessary for the plant operators to complete the actions required, and referral to plant procedures is therefore required in any event. Other changes to procedural details include those associated with limits retained in the ITS. For example, the ITS requirement may refer to programmatic requirements such as COLR, included in ITS Section 5.5, which specifies the scope of the limits contained in the COLR and mandates NRC approval of the analytical methodology.

The removal of these kinds of procedural details from the CTS is acceptable because they will be adequately controlled in the UFSAR, plant procedures, Bases and COLR, as appropriate. This approach provides an effective level of regulatory control and provides for i a,more appropriate change control process. Similarly, removal of reporting requirements I from LCOs is appropriate because ITS 5.6,10 CFR 50.36 and 10 CFR 50.73 adequately l cover the reports deemed to be necessary.

y Administrative Reauirements Redundant to Reaulations (Type 4)

Certain CTS administrative requirements are redundant to regulations and thus are relocated to the USAR or other appropriate SNC-controlled documents. The Final Policy Statement allows licensees to relocate to licensee-controlled documents CTS requirements that do not l

meet any of the criteria for mandatory inclusion in the TS. Changes to the facility or to procedures as described in the USAR are made in accordance with 10 CFR 50.59.

Changes made in accordance with the provisions of other licensee-controlled documents are subject to the specific requirements of those documents. For example,10 CFR 50.54(a) govems changes to the QA plan, and ITS 5.5.1 governs changes to the Offsite Dose Calculetion Manual (ODCM). Therefore, relocation of the administrative details identified above, is acceptable.

1 I

i

Table LA lists CTS specifications and detailed information removed from individual specifications that are relocated to SNC-controlled documents in ITS. Table LA is organized by ITS section and includes the following:

, f

• DOC identifiers, e.g., Section 2.0,11-LA (ITS Section 2.0, DOC 11, LA *No Significant Hazards Evaluation")

+ CTS reference

['

' '4 y

+ the name of the document that retains the CTS requirements the summary description of the change t

+ the method for controlling future changes to relocated requirements a reference to the specific change type, as discussed above ,for not including the information or specific requirements in ITS , ,

s

,,5 x The NRC staff has concluded that these types of detailed information arid specific requirements are not necessary to ensure the effectiveness of ITS to#:-Fy protect the health and safety of the public. Accordingly, these requirements may be rnoved to one of the following SNC-controlled documents for which changes are adequately govemed by a regulatory or TS requirement: .

'N '

?

p  ; p i .

(1)TS Bases controlled by ITS 5.5.14 " Technical Specifications Bases Control Program."

(2)UFSAR (includes the Technical Requireinents Manual (TRM) by reference) controlled by 10 CFR 50.59. D (3)The ODCM controlled by 10 CFR 50.59.

(4)The QA plans as approve'd by the NRC and contained in UFSAR Chapter 17 and controlled by 10 CFR Part 30, Appendix B.

For each of these changes, Table LA also lists SNC-controlled documents and the TS or regulatory requirements goveming changes to those documents.

To the extent that requirements and information have been relocated to SNC-controlled documents, such information and requirements are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

Further, where such information and requirements are contained in LCOs and associated requirements in the CTS '.a NRC staff has concluded that they do not fall within any of the four crite'ria in the Final Polic Statement (discussed in Part 11 of this SE). Accordingly, existing detailed information ano specific requirements, such as generally described above, may be deleted from the CTS.

E. Relocated Specifications The Final Policy Statement states that LCOs and associated requirements that do not satisfy or fall within any of the four specified criteria may be relocated from existing TS (an NRC-controlled document) to appropriate licensee-controlled documents. These requirements include the LCOs, Action Statements (ACTIONS), and associated SRs. In its application, SNC proposed relocating such specifications to the UFSAR (includes the TRM by reference). The staff has reviewed SNC's submittals, and finds that relocation of these requirements to the FSAR (and TRM) is acceptable, in that changes to these documents will be adequately controlled by l

l l

p 10 CFR 50.59 These provisions will continue to be implemented by appropriate plant procedures (i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and work control procedures). A c

& Y~'_

SNC, in electing to implement the specifications of STS, al proposed,jn'acqordance with the criteria in the Final Policy Statement, to entirely remove in TS from the CISimid place them in SNC-controlled documents noted in Table R. Table R 1 all speelfications and specific CTS details that are relocated, based on the Final Policy State te 8NC-controlled documents in ITS. Table R provides: a DOC identification number refe toCTS; a CTS reference; a summary description of the requirement; the name of the M O st retains the CTS requirements; and the method for controlling future changs to M seguirements. The NRC staff evaluation of each relocated specification any specific C15desB presented in Table R is provided below. g$u EW Boration CTS jy kSQygg%gy ,

g vpgW l Requirements to maintain a source of boraled wate one or niese flow paths to inject this borated water into the reactor coolant,aystem (RC8), and appiropriate charging pumps to provide the necessary charging head to overcones reactor psessure for boron injection are relocated to the TRM. The relocation of the fotowing six CTS specifications addressing the boration subsystem of the chemical and volume controfeystem (CVCS) are addressed as a group because each represents an element of the boration subsystem and as such there are common functional requirements as, well similaTrelationships to DBAs:

g -

CTS 3/4.1.2.'1 Flow Path $huloown CTS 3/4.1.2.2 Flow Paths - Operating CTS 3/4,1.2.3 Charging Pump - Shutdown CTS 3/4.1.2.4 Charging Pumps - Operating

' CTS 3/4.1.2.5 Borsted Water Sources - Shutdown CTS 3/4.1.2.6 Borated Water Sources - Operating The operability of those boration subsystems or components required to mitigate a DBA or transient are retained,in Chapter 3.5 TS for emergency core cooling systems (ECCS) since they meet Creerion 3 of the NRC policy statement.

i em V

l The boration subsystem of the CVCS provides the CVCS capability to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin

- (SDM). The operation limits retained in SDM and rod insertion TS provide adequate assurance that the required parameters (SDM and rod position) are maintained within the design and analyses lines. The boration subsystem is not specifically assumed to be operable or credited in the applicuble safety analysis to mitigate the consequences of a DBA or transient, including

. the limiting case of a boron dilution event. In the case of a malfunction of the CVCS causing a boron dilution event, operator response is to close the appropriate valves in the reactor water makeup system. The calculations supporting the analyses of the boron dilution event show i

adequate time exists for operator action to mitigate the event before SDM is lost, criticality is reached or for some form of boration to be initiated to restore the SDM (FSAR 15.2.4).

l l

.15--

CTS 3/4.3 3.7 - Hiah Enerov Line Break isolation Sensors g

Requirements for instrumentation used to either detect and mitigate the discharge of steam or

. water into plant areas or to provide control room operators witt alarms to alert them of a line break event are relocated to the TRM The instrumentatior sonsists of pressure switches which -

monitor the air pressure in piping penetration and equipmen)t' rooms containingMen

- outside of containment. In addition, CTS 3/4.3.3.7 contains requirements for levet sullches'used to detect flooding in the main steam (MS) valve room. Ths instrumentation addressied tpy CTS 3/4.3.3.7 functions to actuate the following isolation valves ~on'a hlgh room air pressure signal:

Instrument air supply isolation valves, Nitrogen supply isolationyelves, steam generator (SG)

. blowdown sample isolation valves, Pressurizer steam space sampleisolation valves, Pressurizer liquid sample isolation valves, RCS loops 2 and 3 sam ~ple isolellon valves, CVCS letdown isolation valves, and SG blowdown isolation salues. .The level aseches are used to detect flooding in the MS valve room and function'1Diftp $te mein feedw'alBr pumps.

[  %%O The instrumentation addressed by CTS 3/4.8.3.7 initiates the;actuallon'of equipment required to prevent damage to the surrounding safety 4 elated systems and structures outside containment.

The functions performed by the instrumentation in CTS 3/4.3.3.7 are not functions that are required by a safety analysis to mitigets the consequences of a design basis (line break) accident described in the FSAR. Valve isolation actuations required to mitigate the consequences of design basis jilpe' rupture adoidents described in the FSAR are performed by ESFAS signals auch as Phase A and B containment isolation, steam line isolation, and main feedwater isolation. ;These ESFAS isolation fuhdions continue to be required operable by the ITS ESFAS LCO." in addition, the'feedwater pump trip actuated by the flood detection instrumentation in CTS 3/4.3.3.7 is unrelated to meeting the acceptance criteria for feedwater line breaks described in FSAR 15.4.2.2.4 (limiting the primary and secondary pressures and ensuring the reactor core remains adequately covered). The reactor trip signals and safety injection actuation signals which are described in the FSAR (15.4.2.2.1) as providing protection against a main feedwater pipe rupture continue to be required operable by the ITS ESFAS and RTS LCOs.

CTS 3/4.3.3 2 - Movable incore Detectors Requirements of the lhlovable incore Detector Specification, CTS 3/4.3.3.2, used to ensure operability'of the instrumentation for monitoring flux distribution within the core are relocated to the TRM. ~ The movable incore detector instrumentation provides information used for periodic surveillances of the reactor power distribution, and for calibration of the excore detectors. The l power distribution surveillances verify that peaking factors are within the assumptions of the design analysis, These TS more directly address the required reactor power distribution limits and peaking factors in ITS section 3.2, Power Distribution Limits. In addition, the Reactor Trip System (RTS), LCO 3.3.1 contains SRs that require calibration of the excore detectors. As such, the RTS specification requires the movable incore detectors to support test verification that the excore detectors are properly calibrated by the information provided by the incore detectors. However, the movable incore detectors the do not directly address the required parameters or the equipment calibrated.

l

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j CTS 3/4.3 4 - Turbine Overspeed Protection Requirements for existing TS 3/4.3.4 conditions, ras, and S_Rs for the turbine overspeed protection system instrumentation are relocated to the TRMT This specification requires the turbine overspeed protection instrumentation and turbine shed control valves'to be operable to protect the turbine from excessive speed which prevents the' generation of potentially damaging i missiles. Although the design basis accidents and transients include a variety of system failures I and conditions which might result from turbine missiles striking various plant systems and equipment, the system failures and plant conditions could be caused by other events as well as turbine failures. 4

-n N.. 1 The operation of the turbine overspeed protection system is not assurned in or credited by any l design basis accident analysis. A turbine missile probabihty analysis wa's performed to determine the potential of turbine missiles to be generated, strike, and cause the failure of safety related equipment. The probability analysis was base,d on the surveillance frequencies of the turbine overspeed protection testing program described in FSAR section [10.2.2 ?) because CTS 3/4.3.4 does not contain SRs for the turbine o'v erspeed protection system. The analysis j showed a low likelihood of turbine missiles being generated from turbine overspeed. The j relocation of this TS does not impact the test program described in FSAR section [10.2.2 7] and I as such, the relocation of CTS 3/4.3.4 does notimpact the analysis of the probability of turbine missiles to be generated, stnke, and cause the failure of safety related equipment, in view of the low likelihood of turbine missiles, the turbine overspeed scenario does not constitute a part of ,

the primary success path to prevent or mitigate such design basis accidents and transients.

CTS 3/4 4.2 - Safety Valves - Shutdown (Modes 4 and 5)

J Requirements for Safety Valves - Shutdown (Modes 4 and 5) are relocated to the TRM. The pressurizer safety valves protect the RCS from being pressurized above the RCS pressure l Safety Limit. In ITS, the pressurizer safety valves are required operable to provide overpressure j protection from operating conditions (Modes 1-3) down to the RCS temperature at which the low temperature overpressure protection system (RHR relief valves) are required operable (Mode 4  ;

s 325'F). Therefore, ITS LCO 3.4.10, Pressurizer Safety Valves, and ITS LCO 3.4.12, Low Teniperature Overpressure Protection System, requirements provide continuous RCS overpressure protection. As such, the CTS 3/4.4.2, Safety Valve - Shutdown, requirement for a single pressurizer safety valve to be operable during all of Modes 4 and 5 is not required for RCS overpressure protection. In addition, the operability of a single safety valve in Modes 4 and 5 is not an assumption of any safety analysis for the mitigation of a design basis accident or transient in Modes 4 and 5.

CTS 3/4 4.8 - RCS Chemistry The RCS chemistry limits of existing TS 3/4.4.8 are relocated to the TRM. The reactor coolant chemistry program provides limits on particular chemical properties of the primary coolant, and surveillance practices to monitor those properties, to ensure that degradation of the reactor coolant pressure boundary is not exacerbated by poor chemistry conditions. However, degradation of the reactor coolant pressure boundary is a long-term process, and there are

g weg-m

y l

I I

i 1

. other, direct means to monitor and correct the degradation of the reactor coolant pressure

. boundary which are controlled by regulations and TS; for example, in-sendoe inspection and

_ primary coolant leakage limits are provided to prevent long term degradallon'of.the reactor 1 2 coolant pressure boundary materials, and provide long term Jacintenancaief acceptable i structural conditions of the system. These limitations are not of immediate' to the operator, and are not required to ensure operability of the IES presspboun , l[ j

$3 p &

CTS 3/4 4.10.2 - Pressurizer k['Oy p%.

DWh Pressurizer temperature limits are relocated to the TRM. Lunile;are pieced on pressurizer operation to prevent non-ductile failure of piping. These lenits areW with the accepted

( structural analysis. Since the pressurizer normally operates in tempt ranges above those for which there is a reason for concem of nonductile eBiseitemperat@0HD8 sits are pressurizer to assure compatibility of operation uSi Wie,@ analysis' performed in accordance with the ASME Code requiremente. While,$tese Enghs represent operating

' restrictions and Cnterion 2 does not apply since it incinefes opesetg sestrictions, Criterion 2 applies only to those operating restrictions sequired 46 preclude'unenalyzed accidents and j transients be included in TS. Howeveria failure of pressurizar integrity would result in an '

l analyzed event (loss of coolant accident) for whah numerous systems and components are required and retained in the TS. Therefore, the pressuriaer temperature limits are not relied on to prevent or mitigate a DBA or'Irenelent, norpre these limits an operating restriction that is j required to produde an unanalysed occident or transient. .

I y ^ '%

I CTS 3/4 412 --- Ramedar Vesset "--S.tras s s Requirements for.RCS Vents are reloosted to the TRM. The RCS Vents exhaust non-

! condensable gases andfor steem from the RCS which may inhibit natural circulation core

! cooling following any event :r;;14 a loss of offsite power and for which long term cooling, such as a loss-of-coolant accident (LOCA) is required. The functional capabilities, and testing requirements for reactor vessel head vents are consistent with the requirements of item II.B.I of l NUREG-0737, " Clarification of TMl Action Plan Requirements," however, the operation of RCS l Vents is not assumed in any safety analysis since operation of the vents is not part of the

!- primary success path for any design basis event. The operation of these vents is an operator action eller the event has occurred, and is only required when there is indication that natural circUlsilon'is not occurring.

w; l CTS 3/4.7.2 - Steam Generator Pressure /Temoerature Limitation Requirements for the steam generator pressure / temperature limits in CTS 3/4.7.2 are relocated  :

to the TRM. These pressure and temperature limits ensure that the pressure induced stresses l are within the maximum allowable fracture toughness stress limits. The values of the pressure and temperature limits are based on maintaining steam generator RTNDT st a level sufficient i

. enough to prevent brittle fracture. However, if failure of steam generator integrity occurs, the plant condition that results is bounded by the analysis of a steam generator tube rupture or other

' loss of coolant accident events for which adequate mitigation systems and components are provided. The systems and components provided to mitigate analyzed events resulting from a failure of steam generator integrity are retained in the ITS. The Steam Generator l i

I

~- . _%

p i

Pressure / Temperature Limitation is not an initial condition of an DBA or transient, noris this limitation an operating restriction that is required to preclude an unanalyzed accident or L transient. ~

i

' CTS 3/4.7.9 - Snubbers r kr &$

AN%?? ~f.

e Q ,V A y Vw Existing TS 3/4.7.9, " Snubbers," states that all snubbers shot be operable. Snubbers are" passive devices that are designed to prevent unrestrained pW motion under dynamlolloods and allow normal thermal expansion of piping and nozzles to eliminete' excessive thermal stresses during heetup or cooldown. The TS action statement for snubbers only requires that an inoperable snubber be replaced or repaired. The SRs forenubbeqis tiet they be periodically examined under the inservice inspection program. The,requirementsgemishng TS 3.7.4 that all j snubbers be operable are requirements that do notimpact teactor opeenm,:do not identify a i parameter that is an initial condition assumption feir~a (Skortransient, 'do not identify a

]

significant abnormal degradation of the reactor coolantpressure boundary, and do not form part l of the primary success path which functions or actuates to mWgste a' design basis accident or i transient. Requirements for snubber operability are relocated to the TRM.

p. p p CTS 3/4 7.10 - Sealed Source Cont $mination[

A

[ I 14  ?

Existing TS 3/4.7.10 " Sealed source Contamination,"ioquires that sealed sources containing radioactive matenal shall be free of a specifie'dremovable contaminatio.1; thereby ensuring that leakage from byproduct, soume and special nuclear material sources will not' exceed allowable values specified in 10 CFR Part 20.! The associated action statement requires that if the removable contamination exceeds limitabons, the sealed source shall be either decontaminated or disposed of. The hmitations expressed in this TS do not impact leactor operation or the safety of reactor opershons.1 Requirements specified in the existing TS have been relocated to the TRM. -

? ,

CTS 3/4.7.13 - Area Temperature Monitorina (Unit 2 Oniv)  ;

r Requirements in CTS 3/4.7.13 for area temperature monitoring are relocated to the TRM.

These limits ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures .

coul'dl degrade equipment over the long-term and may impact equipment operability. However, i equipment ambient temperatures do not give a direct status of the operability of specified equipment,' rather it is only one of many factors used in the evaluation of operability of safety related equipment. Ultimately the operability of safety-related equipment is determined and controlled within the TS by the definition of operability and the individual TS which require the safety-related equipment operable. Area temperatures will continue to be monitored and evaluated for their affect on equipment operability, in accordance with the requirements retained in the TS for the affected equipment. Therefore, the existing safety-related equipment TS and the TS definition of operability provide adequate assurance that safety-related equipment is operable.

i

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[ CTS' 3/4 8.31 - Containment Penetration Conductor Overcurrent Protective Devices

};y y

' Requirements for CTS 3/4.8.3.1, Containment Penetration Conductor Overcupent Protective

' Devices are' relocated to the TRM. This LCO contains requisements fofinadeBed overcurrent .

devices and breaker position or fuse status to minimize the ential for arialesgricalfault in a component inside containment orarewithin n of the cliitainttient. Qt g i Containment electrical penetration conductors protectethe cabling penet ,d by de-energizing breaker trip or removed fuse for circuits not required durinissectof bperation or by inadeBed overcurrent circuit breakers which are periodically surveilledtoensure operability. De-

' energizing an AC circuit minimizes the potential for an electdosi faut in a component inside containment from propagating to equipment outside contain~meritied p ' etentially damaging the penetration? However,10 CFR Part 50 Appendix J leakrate testing required monitoring of all containment penetrations for degradMon'These dev { provide protection for y the circuit conductors against damage or failure dietis'ouemunent heating effects, but are not '

relied upon in any design basis accident or tranIslent. J jf? /yg . WSp i;%g CTS 3/4.8.3.2 - Motor-Ooerated Valves Thermal Overload Paesetion Devices  ;

JY', f jf l Requirements of CTS 3/4.8.3.2, MotorOperatedValve Thermal Overload Protection Devices are relocated to the TRM. This LCO'contains tiiquirements that ensure the thermal overload protection'will not prevent a saisty-islated maior opereled valve from performing its intended safety function.1These devices protect the equipment from potential damage to maintain the i capability of the equipment, but are not relied upon in the primary success path to mitigate a

' design basis accident'or transient. The volves protected by thermal overload devices are required to be operable to support the operability of their associated system. Thus, the operability of the thermal overload protection devices is sufficiently controlled by the LCOs for those systems containing the volves designed with such devices. Moving these requirements outside TS will not, by itself, reduce the existing operability requirements for safety-related motor-operated valves or relex the associated operational restrictions imposed by the applicable system LCOs.

• CTS 3/4.9.3 - Decav Time i n 17 TheyTS for Oecay Time is relocated to the TRM. The Decay Time TS requirement places a  ;

time Smit on reactor subcriticality prior to the movement of irradiated fuel assemblies out of the l reactor vessel i This ensures that sufficient time has elapsed for the radioactive decay of l short-lived fission products and is consistent with the assumptions used in the fuel t andling accident safety analysis. l However, the schedule restraints associated with a normal refueling shutdown always ensures the movement of irradiated fuel does not occur prior to the CTS Decay Time TS limit of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. Refueling outage schedule restraints include RCS cooldown, depressurization, boration,  !

removal of the reactor vessel head and upper intamals, flooding the reactor cavity to the required level, as well as various required testing and maintenance activities. The activities and requirements associated with a normal refueling shutdown are inherent in the design and operating restrictions (i.e., cocidown and depressurization TS requirements, water level TS

,. requirements, boron concentration TS requirements, and TS requirements to maintain and test j ' equipment to ensure operability) associated with pressurized water reactors;

+  ;-

' Although Criterion 2 of the Final Policy Statement would requiis existin[T8 34.9.3 " Decay-

j. Time" to be retained in the ITS, the requirement for a 100 ' decay tirpsifoggeing subcriticality..

before commencing movement of irradiated fuel in the rea r vessel will'alwayspeinst for a.

refueling outage. The operations requircd prior to moving isted fuelin the reactorvessel require in excess of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to complete before irradiatig$selcon be moved. Th'erefore, the requirement is unnecessary and hn been relocated from ticispholAcations to the TRM.

n

(

CTS 3/4 9.5 - Communications JhY&gh 4r

- M&c Requirements contained in CTS 3/4.9.5, "CommunM,are propoeq$5S relocated to the TRM. This specification establishes requirements'te ntSibeln communibation between the control room and the refueling station during refueling operations to eneure that refueling ,

personnel can be promptly aformed of sig "aant chan0es hitSteplant status or core reactivity conditions observed on the controt room instrumentdion. Thinequirement represents good operational practice and is designed t,oerisure sais" refueling 6perations; however the refueling system design accident or transient leaponse does not take' credit for communications.

Therefore, the requirements have been relocated.

m j4

/y Q CTS 3/4.9.6 -ManioulatoLfdEDS h, t, M~ ,

Requirements of CTS 3/4.9.6, "Menipulator Crane" are relocated to the TRM. The requirements )

'of this specification ensure that the manipulator crane and auxiliary hoist will have sufficient load handling capacity, an appropnate overload cut off limit for the manipulator crane, and load indicators for the' auxiliary hoist during refueling operations. Additionally, this specification ensures that the core intemals and reactor vessel are protected from excessive lifting force during refueling operations in the event they are inadvertently engaged during lifting operations.

Although this specification contains requirements designed to prevent damage to fuel assemblies, core intemals, and reactor vessel, these requirements are not relied upon to prevent or mitigate the consequences of a DBA. The limitations of this specification only apply to design requirements / Design control requirements are adequately governed by regulation and.seguired QA plan; y ,

CTS 3M.9.7.1 - Crane Travel - Soent Fuel Storaae Pool Buildina Bridae Crane The crane travel limits specified in existing TS 3/4.9.7.1 support the safety analysis assumption that loads in excess of 3000 pounds will not be moved over fuel assemblies stored in the spent fuel racks. Ve action statements of are provided by physical design and administrative controls. The applicable safety analysis assumes a 3,000 pound load dropped at a height of 42 inches. The results of tMs Analysis show that the accident would not result in damage to the fuel assemblies or an unsafe geometric spacing of the fuel assemblies. This analysis conservatively bounds the drop of a standard fuel assembly with a control rod and handling fixture at the g-.%a.m-u

i maximum lift height of 39.5 inches. The only other heavy load handled by the bridge crane is the spent fuel pool transfer slot gate. The transfer slot gate is moved from its normal position directly to its stored position without moving over stored fuel. However, the storage racks are designed to withstand a transfer slot gate drop. There are no other heavy loads that are handled by this crane. gj g yp -

V; p  % ..

The spent fuel storage pool building bridge crane travel limits are physical limits venfied in' accordance with the requirements provided in the applicable plant procedure and not process variables which are monitored and controlled by the operat'or. Therefore, the requirements have been relocated to the TRM. y4 y g, CTS 3/4.9 7.2 - Soent Fuel Cask Crane ,, . '(

j7 , 'q.

Requirements for CTS 3/4.9.7.2, " Spent Fuel Cask Crane" are relocated to the TRM. The spent fuel cask crane is used for handling shipping casks containing spent fuel assemblies. The crane transfers the shipping cask between the offsite transportation and the cask wash and storage areas. Fuel assemblies are placed in th.e cask by the bridge crane ' The cask crane is prevented from moving above or into the vicinity of the spent fuel pool by rail stops and mechanical bumpers which are permanently attached. The cask crane does not move loads over the spent fuel pool. ,

q ,~,

Although this specification contains requirements designed to ensure correct operation and

maintenance of the spent fuel cask crane, these requirements are not relied upon to prevent or mitigate the consequences of a' DBA. The limitations of this specification only apply to design requirements. Design control requirements are adequately governed by regulation and required QA plan.  %

CTS 3/4 9.12 - Storace Pool Ventilation (Fuel Storaae)

Requirements of CTS 3/4.9.12, " Storage Pool Ventilation (fuel storage)" are relocated to the TRM. This specification contains requirements for operability of the penetration room filtration system whenever irradiated fuel is stored in the spent fuel pool.

The, penetration room filtration system limits releases of radioisotopes to the environment which may leak from the ECCS pump rooms and penetration rooms during the recirculation phase of a design basis LOCA accident. The system also processes the air from the fuel handling area following a fuel handling accident. The system is manually aligned by operators to operate in s one of two modes (fuel handling or LOCA). Normally the system is aligned to automatically process the exhaust air from the spent fuel pool area upon receipt of an actuation signal initiated by either high radiation or low flow in the spent fuel pool exhaust system. In the event of a LOCA, the system is manually realigned to operate in the LOCA mode prior to the end of the injection phase of a LOCA.

i Operation of the penetration room filtration system is assumed in the safety analysis for a fuel handling accident inside the fuel handling building. Therefore, this system lmust be operable and aligned for this purpose when the potential for a fuel handl.ing accident exists in the refueling section of the CTS, this system is addressed in two esperate LCOs,34.9.12 for fuel storage and 3/4.9.13 for fuel movement and crane operatiorQ ^

Since the, potential for a fuel handling accident exists during irradiated fuel movement ttie' CTS 3/4,9.13 (fuld Mt) is -

retained consistent with the STS in the plant systems chapitsr of the TS. Howev5r,'theLCTS .

3/4.9.12 (fuel storage) requires the penetration room filtration system operable and aligned to the fuel handling building whenever irradiated fuel is storedin the spent fuel pit. As such, the potential for a design basis fuel handling accident in the fuel butdin0.cen only exist during movement of irradiated fuel and this is adequately addressed by'Ste soleined CTS 3/4.9.13.

Therefore, the CTS 3/4.9.12 (fuel storage), applicable when there iiiino fuel movement, is not relied on to prevent or mitigate the consequences ofisselpt basis s'ccident and is proposed for relocation.

$g% pef 7 CTS 3/4.10.5 - Position Indication System ASht4 dawn Test Ehn

}if T. .j n n &

The test eheeption for Position Indication bystemshhutdowb allows the CTS 3/4.1.3.3

- requirement that a single digital rod poeltion indicotor be operable for each rod not fully inserted to be suspended in Modes 3,4, and 5 for the purpose of rod drop time measurements.

,, p tr pn vn The control rod position indicatin0 system provides iridication of rod position to the operator.

This indication is used by the operator to verify'that the rods are correctly positioned. In operating Modes 1 and 2, this indication is used during reactor startup and operation to monitor rod position to verify insertion and agg nment limits are met and to verify that the rods are fully inserted into the core immediately following a reactor trip. Rod position indication requirements during startup and operation are addressed in the ITS by the LCO, " Rod Position Indication" which satisfes Criterion 2 of the NRC Policy Statement (verification of initial conditions of a DBA). -

pg v ~

The CTS 3/4.1.3.3 requirement for rod position indication during shutdown, Modes 3,4, and 5, with the reactor trip brealcers closed, specifies that a single digital rod position indicator is required to be operable for each rod not fully inserted. The associated CTS test exception (3/4.10.5) allows this' requirement to be suspended in Modes 3,4, and 5 for the purpose of rod drop'Emo measurements. In the shutdown Modes, the position indicating system provides

- infornistion only and is not relied on by the operators to verify insertion or alignment limits (which are only required in Modes 1 and 2). Therefore, in the shutdown Modes the rod position indication system is not used to verify the initial conditions of a DBA. Additionally, during shutdown Modes, the rod position indication system is not used to verify a reactor trip, or assist in the mitigation of any other DBA or transient. )

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Conclusion The relocated CTS discussed above are not required to be in the TS under 10 CFR 50.36 and do not meet any of the four criteria in the Final Policy Statement.?They are not needed to obviate the possibility that an abnormal situation or event wd give rise to an immediate threat to the public health and safety. In addition, the NRC staff finds that sufficidnt reg'ulatory controls exist under the regulations cited above to maintain the effect of the. provisions in these; specifications. The NRC staff has concluded that approprble controls have been established for all of the current specifications, information, and require'monts that are being moved to SNC-controlled documents. This is the subject of a license condition estabkshed herewith. Until incorporated in the UFSAR and procedures, changes to these s s, info'rmation, and requirements will be controlled in accordance with the applicable cungmt procedures that control these documents. Following implementation, the NRC win audit the removed provisions to ensure that an appropriate level of control has been achieved. tThe NRC staff has concluded that, in accordance with the Final Policy Statement, sufficient regulatory' controls exist under the regulations, particularly 10 CFR 50.59. Accordingly, these specificebons, information, and requirements, as described in detail in this SE, may be relocated fro ~m CTS and placed in the UFSAR or other SNC-controlled docume'nts as specified in SNC's letters dated (SNC to supply dates). # '

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F. Control of Specifications, Refuirement , and Information Removed from the CTS j y The facility and procedures described in the UFSAR and TRM, incorporated into the UFSAR by reference, can only be revised in accordance with the provisions of 10 CFR 50.59, which ensures records are maintained and establishes appropriate control over requirements removed from CTS and over future changes to the requirements. Other licensee-controlled documents contain provisions for making changes consistent with other applicable regulatory requirements:

for example, the ODCM can be changed in accordance with ITS 5.5.1, the emergency plan implementing procedures (EPIPs) can be changed in accordance with 10 CFR 50.54(q); and the administrative instructions that implement the QA Plan can be changed in accordance with 10 CFR 50.54(a) and 10 CFR Part 50, Appendix B. Temporary procedure changes are also controlled by 10 CFR 50.54(a). The documentation of these changes will be maintained by SNC in accordance with the record retention requirements specified in SNC's OA plan for FNP and such applicable regulations as 10 CFR 50.59.

The license condition for the relocation of requirements from the CTS will address the implementation of the ITS conversion, and when the relocation of the CTS requirements into licensee-controlled documents will be completed. The submittal of the updated licensee-controlled documents (e.g., FSAR) to the Commission will be as required by, and in accordance with, the regulations (e.g.,10 CFR 50.71(e) for the updated FSAR), and not be as part of the implementation of the ITS.

G. Evaluation of Other TS Changes included in the Application for Conversion to improved Technical Specifications - f fs p ff, This section addresses the beyond-scope issues in which FNP proposed dienges to both the CTS and STS. The NRC published a notice of consideratiorifor these tqSecope issues in the FEDERAL REGISTER on [date] (64 FR xxxxx). ' The ps discussed tielow an tisted in the order of the applicable ITS specification or section, as opriate..

&$ y%sd ITS 3.2.1 ~- Deletina FOC(Z) and FQW(Z) +& 4

. pg,Oh SNC deleted STS terms FQC(Z) and FQW(Z) and used the terms @,'fsteady state" limit, and " transient" limit. Since the term FQW(Z) does not define the transienttait, it may be deleted. Furthermore, if SNC deletes the term FQM(2), there is no niind to specify a term FQC(Z). These limits may then be referred to as shoody stetsiand transient limits instead of FQC(Z) and FQW(Z). Since these changes do'not alterthe technmalintent of the CTS requirements, the proposed change is acceptable.

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ff ITS 3.2.4 - Limit Thermal Power Rathe),than Reduce Thermist Power b' fa /s

' When_the quadrant power tilt ratio (QPTR) exceeds 1.02, the plant enters CTS action a.2.b which requires operators to redGee thermal powerlevel.tThe STS also requires reducing thermal power under these conditens in the'flS, the phrase " reduce thermal power" is replaced with" limit thermal power.") During startup, QPTR may exceed 1.02 because of transient core conditions due to kenon. These conditions are usually self-correcting as a result of power ascension and xenon bum up. Changing the language to " limit thermal power" rather than

• reduce thermal _ power" would allow incnpases in power within the mode of applicability (when QPTR exceeds 1.02), but would still continue to limit power below the appropriate RTP level.

Therefore, this change is'%.

y .

ITS 3.2.4 - AFD and QPTR Aeolicability SNC revised CTS 3/4.2.41 be consistent with the Mode of Applicability in the AFD TS, and revised the QPTR Applicability from "above 50% RTP" to "250% RTP." Although this change introduces a slightly more conservative Applicability for the QPTR LCO, it maintains the CTS consistency between the AFD and QPTR Applicabilities. This change has no practicalimpact on plant operation and no impact upon safety, yet it eliminates subtle differences between the LCOs. This change is acceptable.

ITS 3.3.2 - Main Steam Isolation Instrumentation SNC in their ITS conversion are not adopting STS footnote (i) and are replacing it with ESFAS Table 3.3.2-1 footnote (d) to take credit for the FNP MS system design. FNP's MS isolation

~ instrumentation is required to be operable during Modes 1 and 2. STS footnote (i) provides an exception to the Mode applicability for the MS line isolation function when all the MS line l

J h c,- * ,_ . . .

• Ta-"- - -~^^-+ ' " * * " - - '

isolation valves are closed.' Once the MS isolation valves are closed, the safety function of the MS isolation instrumentation is accomplished. STS footnote (i) was written for the standard MS system design which contains only one MS isolation valve per MS line, y CTS footnote (") to Table 3.3-3 also requires all MS isolation valves to be closed / However,1 ige w MS system has two isolation valves in each MS line. SNC has proposed to change STS footnote (i) to

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reflect the FNP design since closing one isolation valve in jisch MS line'sccon%Hahes the ,[ -

intended safety function. Footnote (d) to the FNP ITS Sedan 3.3.21iWill allow theEeucoption to' the Mode applicability for the MS isolation function when og lectation valve in each MS line is closed. The staff finds the proposed change acceptable. M%

A; L :p ITS 3.3.4 - Source Ranoe Neutron Flux Monitor for Reu E ShF " Artem Mi&z, CTS Table 3.3-9 does not include the source range M (SRM) fortis Remote Shutdown

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System. FNP uses a separate SRM to give remate imSostion.iSNC has, proposed that they

would submit a report to the NRC if they lost this SRM and faBed to,make it operable within 30 i

' days. The report would explain SNC's prepte6ned attainatisapStedef ensuring the reactor remains in the shutdown condition in the event of a siintrol roisa evicuation, the cause of the inoperability, and SNC plans and schedule for rest 6 ring the SRM to operable status. The STS requires the plant to be in shutdown condition stor 30 days.4 Since adding this SRM to the TS is more restrictive, and since this SRM providespleual indicati6n only and is not used in any automatic actuation signal or to'mohitor the operation N any component necessary to maintain j the unit in Mode 3 (hot standby), we find SNC's justification acceptable.

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ITS 3.3.5 -[ossiaf h%wer. Diedai$mrator Start Instrumentation SNC in their ITS conversion modified STS Section 3.3.5 to include a new degraded grid alarm for the " Loss of Power, DG Start instrumentation" section of the FNP ITS. SNC revised ITS Section 3.3.5 and the corresponding basis sections to incorporate this change since the required action and completion time for this alarm is different from other functions in this section.

This alarm does not exist in the CTS. SNC revised the ITS because of a commitment made in response to a finding documented in the NRC Inspection Report Numbers 50-348/92-17 and 50-364/92-17. The NRC staff accepted SNC's commitment in a SE dated November 21,1995.

Based on this, the staff Snds the change acceptable since it is more restrictive and meets the previous SNC commitment.

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ITS 3 41 '--

- RCS Pressure. Temperature. and Flow DNB Limits i

ITS SR 3.4.1.4 contains a note stating that the 18-month RCS minimum flow surveillance is not I required to be performed until 7 days after 2 90% RTP. This note is consistent with Westinghouse STS SR 3.4.1.4 except that the elapse time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> instead of 7 days in the STS. The intent of the note is to require performing the 18-month surveillance at the beginning of each fuel cycle with the reactor power as close to stable 100% power as possible, especially when performing a precision calorimetric heat balance measurement. FNP's CTS do not specify a time limit for performing the 18-month surveillance. With the added note, SNC will complete l

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I-the surveillance within 7 days once the unit reaches 90% RTP, in accordance with example 1.4-3 in the STS Section 1, "Use and Application." SNC indicated that the proposed 7-day limit is based on operating experience. SNC stated that 7 days provide enough time to set up for measurement. allow for typical instrumentation problems, and achieve stable conditions without FNP,lTS.

adversely affecting safety. The staff finds A the A7-day limit to be acceptabg

- ITS 3.4.2 - RCS Minimum Temperature for Criticality h7 g $

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- FNP ITS SR 3.4.2.1 contains a note stating that the surveillance verWying *RCS Tavg in each loop 2 541" F" is required only if the Lo-Lo Tavg alarm is not inest and any RCS loop Tavg is

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< 547 degrees F. Using the Lo-Lo Tavg alarm in this note as a premeuisite for the surveillance is inconsistent with the Tavg - Tref Deviation alarm specified in the M CTS, but is consistent

' alarm can be with the STS used. SNC providedin the which either reasons thethe for using Lo-Lo Tavg alarm,ortie 14La'Tavg;aParm over , Towg - Tref @trie'Tavg -

alarm. The Tavg - Tref Deviation alarm is calculated using:e median Tsvg and a default or low limited value for Tref of 547 degrees F priorID' turbine loading.iWith a' deviation alarm setpoint of15 degrees F, one loop Tavg could be < 541 degrees F priorto actuating the Tavg - Tref

. Deviation alarm. The Lo-Lo Tavg alarm provides,an alarm any time the temperature in any of the three loops is < 543 degrees F, a6d therefore its use is,more conservative. In addition, three Trip Status Light Box indications are available to the' operator to indicate individual loop Tavg < 543 degrees F, eliminating concerns c5er single failure of the alarm. The staff concludes the ITS to be acceptable. g b y; y ITS SR j 4.5 2-RCS Loops - Mode 3f 4

i SNC proposed ITS SR 3.4.5.2 requir'es verifying steam generator secondary side water levels are 2 74% (wide range) for RCS loops.'The CTS is 10% while the STS has a bracketed 17%.

In response to a staff question, SNC stated that the basis for the minimum steam generator level  ;

during Mode- 3 is to ensure that the steam generator tubes remain covered, thereby ensuring that the associated RCS loop is capable of providing the heat sink for decay heat removal. As a part of the conversion to the ITS, FNP requested Westinghouse to determine the steam generator level necessary to meet the basis stated for SR 3.4.5.2. Westinghouse determined that74% wide range steam generator level was the bounding minimum level. The staff find that this Westmghouse<:alculated bounding level is reasonably conservative and concludes that the SNC-proposed minimum level of 74% wide range steam generator level is acceptable.

w-CTS 3/4.10.4 - Reactor Coolant Loops CTS LCO 3/4.10.4 permits SNC to suspend CTS LCO 3.4.1.1 while performing start-up and physics test' uhen thermal power does not exceed the P-7 interlock setpoint (10% of rated thermal power). As a part of its ITS conversion, SNC proposed to delete CTS 3/4.10.4 from the CTS. SNC stated that this test exception was intended to allow reactor power operation up to 10% rated thermal power with no RCS loops in operation to circulate coolant. At FNP, this test exception served the single purpose of allowing SNC to accomplish natural circulation testing

c during the initial plant start-up test program. There are no current requirements to continue i natural circulation testing after the initial plant start-up test program, and SNC does not perform other testing at FNP requiring this test exception. The staff agrees with SNC's assessment and concludes that it is acceptable to delete CTS 3/4.10.4 from FMP's CTS ind that it should not appear in FNP's ITS. O 4s3, ,y h

ITS 3 4.15 - Reactor Coolant System Leakaoe Detection Instrumerdation Wg l -

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In CTS 3/4.4.7.1, the RCS leakage detection systems that are,,ioquired to be operable include a  !

containment atmosphere particulate radioactivity monitoring system,'.and either a containment 1 air cooler condensate level monitoring system or a containment atmosphere gaseous I radioactivity monitoring system. With only one of these required le'akage detection systems operable in Modes 1 through 4, FNP operation may continue.for up to'7(days provided SNC 1 obtains and analyzes containment atmosphere grab samples at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous or particulate radioactive monitoring system is inoperable. Otherwise,

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FNP must be in at least hot standby within th'e'next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> arsd in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. ,c / b- j 7s j jy For proposed ITS 3.4.15, SNC revised the applicable STS 3.'4.15 LCO for leakage detection instrumentation to be consistent,with FNP's CTS 3/4.4.7.1.5 The STS LCO requires one containment sump monitor, one containment atmosphere radioactivity monitor (gaseous or particulate); and one containment air cooler condensate flow rate monitor to be operable in Modes 1 through 4. S,NC would have 3_0 days to restore an inoperable containment sump monitor. SNC would have 30 days to restore an inoperable containment atmosphere radioactivity monitor or verify that the containment air cooler condensate flow rate monitor was operable. If the containment atmosphere radioactivity monitor and the containment ait cooler condensate flow rate monitor were inoperable, then SNC would also have 30 days to rectere either monitor.

. L Since the CTS does not include the requirement for a containment sump level or flow monitor, the actions related to the containment sump have been deleted from proposed ITS 3.4.15. In the original FNP SE Report (NUREG-75/034, Section 5.6), the staff found the FNP leakage detechon equipment and methods acceptable to satisfy the requirements of General Design Criterion 30 of 10 CFR Part 50, Appendix A without a containment sump monitor. FNP's containment sump design prevents it from being qualified as a leak detection system per RG 1.45.: c Proposed ITS 3.4.15 would extend the time allowed to restore an inoperable RCS leakage detection instrument to operable status to be consistent with the applicable STS. The CTS allows 7 days to restore the gaseous or particulate radioactive monitoring system, if inoperable.

The proposed ITS 3.4.15 change would allow 30 days to restore the containment atmcsphere particulate radioactivity monitor, if inoperable, or 30 days to restore either the containment atmosphere gaseous radioactivity monitor or the containment air cooler condensate flow rate monitor, if inoperable. The associated actions for an inoperable required leakage detection l

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instrument (s) under proposed ITS 3.4.15 includes analyzing containment atmosphere grab samples once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or performing an RCS water inventory balance once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. FNP would be required to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and Mode 5 (cold shutdown) in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> if SNC did not meet the required actions and associated completion times, as stated in the applicable STS.

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The staff has determined that for one inoperable leakage detection monitor, sufficient actions exist with one operable leakage detection monitor and theiemodel actions to analyze,,

containment atmosphere grab samples or perform an RCS'waterinventory balance once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the inoperable monitor, to provide the necessary in'viir.stbn to ensure that RCS leakage will not go undetected. Additionally, SNC stated # tat thi non-TS dew point temperature monitoring system should be available to provide further assurance'that SNC will detect RCS leakage in a timely manner. Therefore, the staff concludes'that exten' ding the allowed outage time for one inoperable leakage detection monitor'from 7 days to 30 days for proposed ITS 3.4.15, is acceptable.

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Il ITS 3.5.3 - ECCS Shutdown LCO y/ / $'

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SNC added a new Action to the ECCS-Shutdown LCO. The'new Action provides an allowed outage time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the required ECCS centrifugal charging subsystem to be inoperable provided the remaining operabis ECCS comp 6nents are capable of providing 100% of the ECCS flow, equivalent to a single operable ECCS subsystem. At FNP, the requirement to ensure that only one centrifugal pump is operable for overpressurization concerns is not applicable until the temperature of one or more of the RCS cold legs is s 180 degrees F.

Therefore, in Mode 4, two or more centnfugal charging pumps may be available and an Action similar to that in STS 3.5.2 (ECCS-Operating) may be applied in FNP ITS 3.5.3 (ECCS-Shutdown).

The proposed allowed outa"go time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for this condition is consistent with the time currently allowed for one train of ECCS to be inoperable in Modes 1-3. The exposure of the unit to the small probability of an event requiring ECCS during this time is considered insignificant and offset by the benefit gained through avoiding unnecessary plant transients.

Thi new Action provides 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of outage time for a ECCS centrifugal charging subsystem

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wheh in Mode 4,' hot shutdown. However, this new Action is permissible only if the remaining operable ECCS components are capable of providing 100% of the ECCS flow, equivalent to a single operable ECCS subsystem. Another Action requires that if 100% ECCS flow equivalent to a single operable ECCS train is not available, then the required action is for FNP to be in Mode 5 in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In combination with the above stated requirements, the staff finds the proposed new Action for a 72-hour outage time for the ECCS centrifugal charging subsystem while in Mode 4 to be acceptable.

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29-ITS 3.5.5 - RCP Seal iniection Flow .

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SNC proposed to change the way they confirm reactor coolant pump (RCP) sealinjection flow, in converting to ITS, SNC would use a graph to measure seellr$setion flowinstead of verifying a single operating point. SNC would determined appropriatejow based op te 4Elerence between the RCS pressure and the charging discharge healer pressure? The' points on the graph are based on FNP-specific safety analysis assumptions which^ provide the ' relationship header pressuri^over a between seal injection flow, RCS pressure, range of values for each of these parameters. Establishing and charging %'s fuference d allows SNC to more precisely and repeatedly verify seal injectleriSow an.d proper throttle valve position. The NRC approved this method of determining die seal flow limit for Vogtle.

In response to an RAl, SNC explained how they estM_the graptCh based the graph on a minimum differential pressure between the dierging header and the, pressurizer and verified total seal injection flow to be within the limits delegnined in accordance with the seal injection resistance assumed in the ECCS safety analjisesNRis24 gpm and 31 gpm points are based on the required flow and differential pressure determinMd lenaccordance with the conditions discussed above. The 27 gpm point is based on tais linear graph between the previous points, which is conservative as compared to the actual 27 gpm point that would be determined in accordance with the conditions discussed above. This information justifies the

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acceptability for using the grapdin Mace of the single ~cherating point. Therefore the staff finds this proposed TS change to be acceptable. (

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ITS 3.7.1 - Reducino the Power Ranom Neutron Flux Setooint The ITS retains the CTS requiremenito reduce the power range neutron flux high trip setpoint based on plant safety analysis and the Westinghouse Nuclear Safety Advisory Letter NSAL-94-001 recommendation of January 20,1994. The trip setpoint is reduced to mitigate the loss of turbine load / turbine trip transient, since the high pressurizer pressure and over temperature delta T trips may not act quickly enough to prevent over pressurizing the secondary system when the initial transient condition is partial power operation with inoperable main steam safety valves (MSSVs). FNP uses a note to revise the applicability of this action in the ITS that only requires reducing the power range neutron flux high trip setpoint in Mode 1, and not in Modes 2 and 3 as the CTS currently requires. In Modes 2 and 3 a rod withdrawal event is tbs onlyhnsientl as discussed in the FSAR, which would result in a significant power rise and potentially challenge MSSV relief capacity. The power range neutron low-flux trip and the source range neutron flux-high trip provide adequate protection for this transient in Modes 2 and

3. Therefore, this change in applicability of the action to reduce the power range neutron flux high trip setpoint is acceptable. -

ITS 3.7.2 - Main Steam isolation Valves SNC's March 12,1998, letter proposed to revise TS 3.7.1.5, " Main Steam Isolation Valves (MSIVs)," to take credit for redundant MSIVs in each eteam line at FNP, Units 1 and 2. CTS l

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3.7.1.5 and STS 3.7.2 for MSIVs apply to typical pressurized-water reactor designs which include' a single MSIV per steam line. In CTS 3.7.1.5, with one MSIV inoperable in Mode 1, SNC has to restore it to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in Mode 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ~ With one MSIV inoperable in Modes 2 and 3, SNC has to restore the vdve to ope'ralde status or close it within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering Mode 2; otherwise, be in Mode 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ~and Mode 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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At the FNP, there are two MSIVs in each steam line, whic redu'ndant. One M$Nin'esch {

steam line is designed to actuate on a train A ESF isolatiorEsignal and the other MSIV in each steam line is designed to actuate on a train B ESF isolation algnal.dhe two MSIVs are installed adjacent to each other on each steam line with no significent pipe volume between the valves.

The actuation of either MSIV in a steam line fulfills the isolation reqilleeseents,of the applicable both MSIVs to close safety in order toanalyses. There mitigate an event are Since at FNP. no design thereis abasis merenos accident betwee analyses that re@'n the F and the design the STS address, FNP chose to take credit for the two MSIVs in each steam line for proposed ITS 3.7.2.

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in proposed ITS 3.7.2, SNC includes two condition 5 (A and B) for inoperable MSIVs in Mode 1.

Condition A applies when one or more steam lines have one' inoperable MSIV in Mode 1. The required action is to restore the inopierable MSN to operatile status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72-hour completion time is based on completion times provided in the CTS and the STS for one inoperable train in redundant engmoered safety _ feature (ESF) systems. Also, there is a low probability of an accident occurving during this time that would require the MSIVs to close.

Condition B applies when one or more steam lines have two MSIVs inoperable in Mode 1. The required action is to restore one MSIV to operable status in the affected steam line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4-hour completion time is consistent with the CTS and the intent of the STS requirements for a loss of isolation function in plants designed with only one MSIV per line. If the required action and associated completion time of Condition A or B are not met, then Condition C would require the unit to be in Mode 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Condition C is consistent with CTS and STS actions and requires the same power reductions and completion times.

In proposed ITS 3.7.2, SNC includes two conditions (D and E) for inoperable MSIVs in Modes 2 and 3J Condition D applies when one or more steam lines have a single inoperable MSIV in Mode 2'or 3.tThe required action is to restore the inoperable MSIV to operable status or close at le'ast one MSN in the affected steam line within 7 days and verify that it is closed once every 7 days thstesfter. SNC proposed this 7-day completion time since this would occur in Modes 2 and 3 when MSIV testing and maintenance, including valve stroking, may be performed during hot conditions, and power is restricted to 5% reactor thermal power. Condition E applies when one or more steam lines have two MSIVs inoperable in Mode 2 or 3. The required action is to restore one MSIV to operable status or verify that one MSIV is closed in the affected steam line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and once every 7 days thereafter. The 4-hour completion time is consistent with j the CTS and the intent of the STS requirements for a loss of isolation function in plants designed with only one MSIV per line. If the required action and associated completion time of Condition D or E are not met, then Condition F requires the unit to be in Mode 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and l

I Mode _4 in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Condition F is consistent with the CTS and the STS actions and requires the same power reductions and completion times. j1 s /[

~n The staff concludes that SNC's proposal to take credit for reddulent MSfVs NITS 3.7.2 is acceptable based on the following: j;'

The action times allowed by the CTS and the STS for a , single inoperable MSIV do.not consider that a redundant MSIV remains operable in the anocted steam line and is fully capable of performing the intended safety function adin the FNP design f[MM

• There are no design basis accident analyses that would require tbth MSIVs to close in order to mitigate the event.

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ITS 3.7.8 - Service Water System /,h]Q!x '

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}lN W x2 SNC's March 12,1998, letter proposed to add an action to133.'7A,fService Water System (SWS)," to account for the redundant automatic turbirie buildiig; idolation valves in each service water train at the FNP, Units 1 and 2. [*

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CTS 3/4.7.4 and the associated ST8 do not account for a specific design feature at the FNP, which includes two redundant aiutorinatic turbiriis. bubdirig isolation valves in series in each SWS train. The valves close automatically on a safety ir(setion signal to isolate the non-safety turbine

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building SWS loads and ensure adequate SWS flow to essential components. When two of these isolation valves,' one in each SWS train ' become inoperable, the actions in both the CTS and the STS do not address a condihon that would be applicable. Therefore, FNP would enter LCO 3.0.3. SNC behoves that entering LCO 3.0.3 in this condition would be overly conservative since there are two automabc turbine building isolation valves in each SWS train, and one automatic valve in each SWS train would still remain fully operable. Therefore, two 100%

capacity SWS trains would % available to provide the required system safety function.

In proposed ITS 3.7.8, SNC added an action for one inoperable automatic turbine building isolation valve in each SWS train, which requires that both inoperable valves be restored to operable status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. SNC based the 72-hour completion time on the fact that the isolation function pedormed by these valves is not lost in either SWS train since one automatic turbliis building isolation valve in each train remains operable. Although the reliability of the isolation funchon performed by the automatic turbine building isolation valves is reduced, there ,

are still two 100% capacity SWS trains available to perform the required safety function of the system, and there is a low likelihood of an event occurring during this time that would require the  ;

isolation function provided by these valves.

Based on the above, the staff concludes that adding the action to ITS 3.7.8 at FNP is acceptable based on the redundant design of the automatic turbine building isolation function in each SWS train and to prevent an unwarranted entry into LCO 3.0.3 resulting in an unnecessary plant l shutdown.

l

g

-3 2 -

ITS 3.8.1 - Removal of Accelerated Testino Reauirements for Emeroency Diesel Generators e

CTS surveillance 4.8.1.1.2.a refers to accelerated testing requirements fo'r emergency diesel generators (EDGs). Table 4.8-1 of the CTS contain thesa accolorated testin0' requirements.

Table 3.8.1-1 in NUREG - 1431 (STS) also contains EDG accelerated testing requimments.

b, f%. Nc: .

For the FNP iTS, SNC proposes to delete the EDG accelerated testing requirements consistent with GL 94-01, " Removal of Accelerated Testing and Spedal Reporting Requirements for.

Emergency Diesel Generators" guidance. GL 94-01 contains guidance that allows utilities to remove the accelerated testing requirements specified in TS.provided the provisions of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Mainteristice at Nuclear Power Plants (maintenance rule), including the applicable regulatorypance wh au5 provide a program to assure EDG performances are implemented. Aftep,unplementing the performance program as required in the maintenance rule, GLM01 notes that utiliti5s can remove individual EDG accelerated testing requirements from TS.>Since FNP hos~ implemented the maintenance I rule, the proposed ITS do not include accelersted testirig requirements However, RG 1.160, ,

" Monitoring the Effectiveness of Maintenan'ce at Nuclear Pow'er Plants," addresses a docketed commitment to maintain selected target reliability values for EDGs. In this regard, the staff expressed concern regarding such a docketed cornmitment.4 fn response to this concem, SNC noted their commitment for an EDG target reliability value of .95. Thus, by implementing maintenance rule requirements [SNC effectively retainid the intent of the accelerated testing requirements to ensure EDG'reinbility and avetetdity, and as such the regulations continue to provide adequate assurance"of EDG performance. The frequency of ITS EDG testing is 31 days as determined from STS Table 3.8.1-1. This frequency is also consistent with GL 94-01 guidance. Based on the above, this change is acceptable.

b, ITS 3.8.2.1 - Eliminatina Raouired Surveillances for Modes 5 and 6 Plant Conditions That Demonstrate Caoabilities Not Raouired for These Modes CTS 4.,8.1.2 requires SNC t derform certain surveillances in Mode 5 and Mode 6 (refueling) that demonstrate capabilities that are not required for these Modes. Table 1 showr, these surveillance requirements (SRs) and the corresponding ITS SRs. CTS SR 4.8.1.1.1.b addresses transferring the unit power supply from the normal power circuit to the alternate circuit.. SR 4.8.1.1.2.c.B addresses a simulated safety injection signal overriding the EDG test )

mode! SR 4.8.1.1.2.d addresses simultaneously fast starting of each EDGs. SR 4.8.1.1.2.c.3 l addresses automatic fast starting of an EDG from a safety injection test signal (without loss of  ;

offsite power). SR 4.8.1.1.2.c.9 addresses EDG automatic load sequencer timers. SR 4.8.1.1.2.cA addresses simulating a loss of offsite power in conjunction with a safety injection test signal and automatic EDG start and load sequencing.

i

Table 1 z.o Surveillance Requirement CTS ITS s SR 4.8.1.1.1.b kSR3.8.11 p -

SR 4.8.1.1.2.c.8 $M 3.8.1.15 Y

SR 4.8.1.1.2.d ,AR 3.8.1.19, SR 4.8.1.1.2.c.3 [ SR 3.8Sd01

• SR 4.8.1.1.2.c.9 /$ Mgm 3.8.1.1k s 4

SR 4.8.1.1.2.c.4 &e SR 3.8.1.17 y .n pp pg ,[ jf""

The FNP ITS revise CTS 4.8.1.2 survaBances to aliminate the requirement to perform the above ITS Section 3.8.1 surveillances for Modes 5 and 6' plant conditions. Since the SRs for AC

^

sources - shutdown define and verifp the operability requirements of the AC sources required in

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Modes 5 and 6, the AC sources - shutdown survelNance requirement is revised in the ITS to more clearly identify the applicable operability requirements. The definition of operability refers to the system or' equipment being capable of performing its required safety function. The above surveillances proposed in the ITS as exceptions for Modes 5 and 6 do not demonstrate any capability related to requred safety func2on of an AC source for these Modes. The revised AC sources - shutdown surveillances in the ITS do not require SNC to meet or perform the excepted surveillances for Modes 5 and 6 plant conditions. SNC took exception to surveillances that require the following: i (1). ' demonstrating the pability to transfer offsite circuits (only one offsite circuit is required

~

in Modes 5 and 6) .

(2) ; demonstrating AC source response to an engineered safety features actuation (SI)

f : signal (SI is not a required safety function in Modes 5 and 6)

(3) -verifying EDG starting independence (only one EDG is required in Modes 5 and 6)

' ~

Except'for SR 3.8.1.10, eliminating the requirement to perform the above FNP ITS surveillances for Modes 5 and 6 plant conditions is consistent with NUREG -1431 guidance. SR 3.8.1.10 verifies that an SI signal fast starts each EDG, each EDG operates for2 5 minutes, and emergency loads are energized from the offsite power system. However. for Modes 5 and 6 plant conditions, operators defeat the SI signal. In this regard, requiring SNC to perform this surveillance for Modes 5 and 6 plant conditions demonstrates a capability that is not required for these Modes.

In addition, the FNP ITS Bases note that during plant shutdown Modes, and consistent with ITS

- Section 3.8.10 (Distribution Systems - Shutdown), portions of a second train of the electrical power distribution subsystems are required to be operable. Further, the FNP ITS Bases note the following: gy ]

(1) Required portior)s of the second train of altematinfhurrent pow dibnS5n' , f subsystems may be energized from the associated inverter Sonnected to the required direct current (DC) bus or the attemate Class 1E Msource consisting of the inverter static transfer switch and the associated constant % transformer, or by}ll,.

(2) - Required DC buses associated with the second train of Splitsution subsystems are energized from either an operable DC source, consisting of ephettery, one battery charger, and the corresponding control equgunent and intercognec8ng cabling associated with that train or a battery chasgeruglig9te corresponding control equipment and the interconnecting catWirig within the train.' .A

[ N N&'

The staff expressed concem regarding the reliability 5f a less9 ten fully complimented second Class 1E power train, since the second power train may not consist of a fully complimented Class 1E power train. In response to'this concers SNC agsed to include in the FNP ITS Bases Sections 3.8.8 (Inverters - Shutdown) and 3.8.10 the following:

, j. tF gymgy Class 1E power and distribution systems are normally used because these systems are

' available'and reliable. Housvar, due to events such as maintenance or modification, i portions of the. Class 1E sysfeni may be temporarily unavailable. In such an instance the plant staff assesses the alternateisystems to ensure that defense in depth is maintained and that risk is minimized. s y .

The FNP CTS do not requiroISNC to assess necessary second power trains during plant shutdown Modes to minimias dek. As such, these actions are considered voluntary. In addition, licensee voluntary actions be ond the CTS, which include safety planning and assessment in shutdown, were an important part of the Commission's decision to cancel the shutdown rule.

n Based on the above, we conclude that eliminating the requirement to perform the identified surieRanoes for Modes 5 and 6 plant conditions is consistent with the provided in NUREG - l 1431)didence and is acceptable. Further, we conclude that SNC's voluntary actions for the i necessary escond power train during plant shutdown Modes are consistent with the Commission's decision to cancel the shutdown rule and are acceptable.

ITS SR 3.8.2.1 - Addina a Note to Exclude Performina Specific Aoolicable Surveillances On Certain AC Sources for Plant Shutdown Conditions ITS SR 3.8.2.1 adds a note to revise CTS SR 4.8.1.2 for AC Sources During Shutdown. The note identifies ITS SR 3.8.1.8, SR 3.8.1.11, SR 3.8.1.12, SR 3.8.1.13, SR 3.8.1.14, and SR 3.8.1.18 as applicable but not required to be performed. SR 3.8.1.8 addresses each EDG 1 .--%.. , _ - - , , - . - - e e % _ ,...

,m . . . .

l rejecting a load greater than or equal to its associated single largest post-accident losd. SR 3.8.1.11 addresses bypassing EDG protective trips when receiving an by an Si signal. SR 3.8.1.12 addresses the EDG 24-hour endurance test. SR 3.8.1.13 addresses each EDG starting and achieving specified voltage and frequency valuesjn312 sscands within 10 minutes of shutting it down after operating it for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded at afteater thanbr equal to specified load. SR 3.8.1.14 addresses synchronizing a loaded EDG with the offsilei pode(source and .

transferring the EDG loads to the offsite power source. SR 3.8.1.18inddresses e'ach EDG~

rejecting a load of 21200 kW ands 2400 kW. Ect ~ b

\ :x Adding the note is intended to stop requiring the operable EDG to be paralleled with the offsite power network or otherwise rendered inoperable to perfosin a sOnegence. .Some of the CTS surveillances required to be performed involve tests that require paMcEng the EDG to the offsite power network. This condition, with the only reiq'uired EDG andte onip required offsite circuit connected, presents a risk of a single fault resulting in a station blackout. To address this concern and to avoid other conflicts with testing and opershWty, the STS' includes a note with SR 3.8.2.1 to exclude the requirement to perform certain sOrveillance tests. The exception provided by the note does not exclude the requirement for the EDG to be cap'a ble of performing the particular function but rather SNC doesirsot have to demonstrate the capability while the EDG source of power is being relied on to meet the LCO. Thus, adding the note to the FNP ITS SR 3.8.2.1 to exclude the requirement to perform the above surveillances is consistent with STS guidance and is acceptable. /

/ '

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ITS SR 3.8.2.1 - Addino a Ust of Soncific Sur9eillances Reauired to be Performed for Modes 5 and 6 Plant Conditions M FNP ITS SR 3.8.2,1 specifically notes that SR 3.8.1.1, SR 3.8.1.2, SR 3.8.1.4, SR 3.8.1.5, and SR 3.8.1.6 apply and SNC is to perform them for Mode 5 and 6 plant conditions. SR 3.8.1.1 l verifies correct breaker alignment and indicated power availability for offsite power circuits. l SR 3.8.1.2 verifies that each EDG starts from standby conditions and achieves specified steady- l state voltage and frequency values. SR 3.8.1.4 verifies that each EDG day tank contains a j

minimum specific number of gallons of fuel oil. SR 3.8.1.5 verif'3s that the fuel oil transfer '

system operates to transfer fuel oil from the storage tank to the day tank. SR 3.8.1.6 verifies thatleach EDG starts from a standby condition and achieves specified voltage and frequency ,

valpin512. seconds l The CTS SR 4.8.1.2 provides requirements for the AC Sources during shutdown by referencing SRs 4.8.1.1.1 and 4.8.1.1.2. The CTS SR 4.8.1.2 only provides a specific exception to the j referenced surveillances required for the shutdown AC sources but does not identify the specific i surveillances required to be performed. The STS provides additional exceptions, as identified and addressed above, but also does not identify the specific surveillances that are required to  ;

be performed for the shutdown AC sources. Proposed ITS SR 3.8.2.1 adds a specific list of surveillances that apply and which must be performed. This list of required surveillances is consistent with the FNP ITS lists of surveillances remaining after deleating exclusions which are not required to be performed. The added list of five required surveillances for ITS SR

i - 3.8.2.1 is also consistent with the five surveillances the STS require to be performed. Thus,

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adding this list of five required surveillances does not introduce a technical change to the SRs, but rather is an administrative change which is provided to clearty identify the surveillances actually required to be performed on the shutdown AC sources 3 Accordingy,pis change is only an administrative change that list the five specific surveillandes required to be and thus is acceptable. yp n g

Wh,ph.e A .

ITS 3.8.4 - Revisions to STS 3.8.4 Ar* ions to Retain An .". ^ (*) Footnote and CTS Battery Connection Resistance Reauirements V  ;

g*gf ITS 3.8.4 actions revise the STS 3.8.4 actions to be consstent wth9te FNP design and CTS requirements. ITS 3.8.'4 provides new action conditions besed on tkQ15h the FNP specific design which includes batteries dedicated to the SWOCl control powier.2 Proposed FNP

t. specific action conditions are necessary to retain tisii RP specific allowance provided by an asterisk (*) footnote associated with two of the. CTS surveEences and to address the fact that the inoperability of the service water intake structure bEttery aSects'enly the associated train of the SWS. # F  %

l

[ .h 8 An asterisk (*) footnote to the surveillances of both the auxillery building and service water intake structure battery CTS provides'en allowince to deliy declaring the battery inoperable due l to connection resistance not within Emit. ThisJootnote' establishes a 24-hour completion time to restore connechon resistance to within the required tolerance. The CTS allowance to restore

~

' battery connachon resistancia to within the required tolerance is supported by IEEE-450 which j . notes that connedion resistance is mensly an indication of conditions that can be easily corrected prior to the next inspection. In addition, IEEE-450 does not note that battery connection resistance is a besis on which to declare the battery inoperable. In this regard, ITS 3.8.4 includes a separate adion condition for connection resistance not within limits. The l completion time associated with this proposed ITS condition is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> which is the same as that specified in the CTS footnote.

- ITS 3.8.4 action conditions revise the STS 3.8.4 action conditions to incorporate a specific

. default action for the service water intake structure battery system. This change is necessary since this battery system supplies only one required safety system, which is the SWS. The l SWgjes a completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for an inoperable train. Each train of service water I intake ~ structure betteries supplies the associated service water DC control power. Considering that either unit or both units at the same time are permitted to have one train of service water  !

inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (as permitted by the CTS or STS) for reasons other than the DC l L control power, the completion time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with a plant shutdown required if the completion i l time is not met is very restrictive for an inoperable service water intake structure battery train. In this regard, ITS 3.8.4 provides separate action conditions for an inoperable service water intake j structure battery train which, if not met, default to a condition that requires the associated train of !

service water to be declared inoperable. This action condition format is consistent with that contained in the STS where similar support / supported system relationships exist. j i

r ITS 3.8.4 also _ revises STS 3.8.4 battery connection resistance SRs to be consistent with the

FNP CTS. The terminology used in the surveillance requirement and the seeistance values specified are revised to be consistent with the language and connection resistance limits used in the FNP CTS. The above changes are consistent with the FRCTS or'speallic design, are

,provided in a format and presentation consistent with that ined in the STS, and are acceptable.

fy? , u ta e v.

ITS 3.8.9 - Revi= ion to CTS Aden Statement for Service ^^^ ^ DC Distribution an'd n= Mary Systems -

Wm CTS 3/4.8.2.5 action statement for the ter LCOdelitbution for thesystem servicewa&rJh requires Nh that with one of the 125-Vdc distribution trains inoperab,ja.festore le distribution system to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in atjesWhet standby the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Prop'es8W3.8.9 revise's the CTS LCO action statement such that with one of the 125-volt distribution trains inoperable, restore the inoperable distribution system to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 1840urs tem discovery of failure to meet the LCO or declare the associated service water' train in'epefable immediately.

At the FNP, a separate DC distribution .h h battery' system that is indepe

[and distribution and battery system suppBei SWS DC controlpower. The primary purpose of the service water intake structure DC detribution sind belle'rp system is to supply DC control power to the associated service water train. 'The service water intake structure DC distribution and battery system does not supply any atter T84 elated or required loads. The CTS requirements that apply to an inoperable servloe water trair) aAow 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore that train to operable status before requiring a plant ' shutdown. As the service water intake structure DC distribution and battery systems are required in TS solely to support the associated SWS train, the CTS allowance of only 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore the servios water intake structure DC distribution and battery systems to operable

- status before requiring a plant shutdown is very conservative. If an entire service water train may be ipoperable for any reason for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, it is very restrictive to require a plant shutdown to

~

begin in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if the DC control power to the same service water train is inoperable. The 2-hour restoration time associated with distribution and battery systems is based on the fact that DC systems typically support many TS required engineered safety feature systems and the loss of the supporting distribution or battery system impacts many required systems. This is not the case for the {NP service water intake structure DC distribution and battery system. The proposed iTS

. actios statements' revise the CTS actions to be similar to other STS actions for support systems  :

when that support system becomes inoperable (that is, declare the supported system inoperable). l The current default actions to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> are  ;

replaced with an action to declare the associated service water train inoperable immediately, in  !

this regard, the completion time for the support system becomes more consistent with the i completion time for the supported system. The CTS action to restore the inoperable distribution or battery system to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is retained to ensure prompt attention to problems with the distribution or battery systems and also allow a reasonable time to restore minor problems j before requiring that the associated train of service water be declared inoperable. In addition, '

retention of the CTS 2-hour action statement maintains greater consistency with associated STS

38--

actions.for other distribution and battery systems in STS LCOs. On the basis of the above technical reasons being consistent with the guidance provided in the STSitio proposed action statement for ITS 3.8.9 is acceptable. A C .-

ITS 5.1.2 - Administrative Controls - Resoonsibility gafr? kfkh..

g# 4

. A b ~4 .,s SNC proposes to adopt less restrictive wording for a portio [n of TS 5.1,' ResponsiblWy contained in their submittal would substitute the generic titlhi Se6ior Reactor Operator.(SRO) for the ITS specific title of Shift Supervisor (SS). The specification seguires that an individual must have an SRO license to assume the specific command and_conkultmotion. An on-shift SRO can be designated to assume the command and control functi$n ESS,4 eaves the control room because the qualifications, as defined in ANSI 3.1,199A"Selectio 'on and Training of j Personnel for Nuclear Power Plants," for an on-shigSIO,and for the  ; designated as the Shift i Supervisor are the same. The change in wording is toes stenotive but not reduce the underlying requirement of the specification and,*thereforeilsacceptabd i 8 [ 3MSh ITS S 3.1 - Administrative Controls - UnitStaff QuaMcations$b' gb g ff ,

SNC proposes to adopt alternative wonfing to reder to the irMidual responsible for management of '

the radiological protection program for plant operations., SNC proposes to use the generic reference of " senior individual dchErge of HeWi Phyeles' in lieu of the ITS specific reference to the Radiation Protection Manager.1The generictWe'would apply to 5.3, Staff Qualifications. The use of a genenc reference iri teu of a specific title'does not change SNC's current commitment to an appropriate qualificahon standard as specified in TS 5.3. Although the wording proposed by SNC does not match the ITS speelfic reference, the individual with responsibility for the radiation protection program can be identified and'is required to be qualified to an appropriate level and, therefore, is acceptable.;

m ITS 5.5.7- RCP Fivwheel'inanaction Freauency

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We will provide the SE inpuIt a' t a later date.

ITS '5.8.7 - Revisions to the Emeroency Diesel Generator Failure Report The% w ,~-

annual w nh diesel generator reliability data report in CTS 6.9.1.12 is replaced in the ITS with STS 5.6.7,'EDG Failure Report. The content of the replacement report is modified to include additional j information to be supplied with each report consistent with the current FN_P practice. The CTS ,

annual EDG report requires that all tests and the number of failures to start on demand for each i EDG be submitted each year. In addition, the CTS requirement references RG 1.108 for report content. The STS EDG reporting requirement is based on the number of failures in the last 25 )

demands, ar.d a report is only required to be submitted when an individual EDG experiences four or i more valid failures in the last 25 demands. The EDG reporting requirement is revised in the ITS to correct the reference for the additional information to be included in the repot'. The additional i information to be provided in the EDG failure report is currently derived from the FNP EDG 1

1

reliability monitoring program, which provides more information than the CTS reference to RG 1.108. This FNP program is referenced in place of the STS references to RGs 1.9 and 1.108 for the additional information and exists to fulfill a commitment provided in response to the station .

blackout rule. The elements of the EDG monitoring program are consistent with the guidance provided in RG 1.155, RG 1.9, and Appendix D of NUMARC 87-00, Rev c 1. The FNP monitoring program ensures the data on all EDG demands is logged and evaluated and that EDG reliability performance is monitored in accordance with RG 1.155, RG 1.9, and Appendix D of , ..

NUMARC 87-00, Rev.1. In addition, this monitoring program specifies the actions to be taken if one or more of the EDG reliability performance indicators datated in the program reaches or exceeds the FNP reliability trigger values. In this regard, the STS EDG failure report is revised in the ITS to be consistent with the current FNP practice regarding'the addihonal information to be included in the report. The other aspects of the STS report require'rNat prov6de more relaxed requirements compared to the CTS requirement for a report to be subfetled ~each year regardless of the number of failures and all tests and all failures to be reported each year. However, the revised reporting requirement does require additionalirdormation such as a description of the failures, underlying causes, ar,J corrective action taken." This adelonel information is necessary to assess the EDG reliability and the overall effectiveness of the FNP EDG maintenance and testing program. Therefore, the STS reporting 'r equirement as modified in the ITS by reference to the FNP EDG reliability monitoring program is' acceptable'as it contirises to provide a means to monitor the FNP EDG reliability and allows for corrective minasures to be taken if required.

p p ,y n ITS 5.7.1.c - Administrative Controls - Hioh R=8=% Area w ~

SNC proposed to adopt alternative wording in TS 5.7.1.c, High Radiation Area. The ITS references the Radiation Protechon Manager as specifying the frequency of periodic radiation surveillances.

The existing FNP specificabon assighs that responsibility to the individual with the title of Health Physics Supervisor. In the conversion to ITS, SNC proposes to substitute the more generic reference of health physics supervisor rather than referencing a specific title. The requirements related to a high radiation area remain unchanged. The level of the individual required to establish the frequency of radiation surveillances also remains unchanged. SNC's proposed wording varies from the ITS specific reference but is consistent with the existing specification and does not change the requirements related to a high radiation area and, therefore, is acceptable.

p , g-IV. STATE CONSULTATION t" +

In aedordance with the Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendment. The State official for the State of Alabama had no comments.

V. ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21,51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the FEDERAL REGISTER on [date] (64 FR xxxxx).

Accordingly, based upon the environmental assessment, the Commission has determined that

r

-4 0 -

issuance of this amendment will not have a significant effect on the quality of the human environment. A m  ;

VI. CONCLUSION 46 N'Sg y J . ,y JF 4:: %. -,

The FNP ITS provide clearer, more readily understandable requirements to ensure sale operation of the plant. The NRC staff concludes that they satisfy the guidance in the Cominission's policy statement with regard to the content of TS, and conform t(See acidel provided in NUREG-1431

~

with appropriate modifications for plant-specific considerati6n's?1he NRC staff further concludes l that the FNP ITS satisfy Section 182a of the Atomic Energy Ad,;10 CFR 50.36 and other applicable

~

standards. On this basis, the NRC staff concludes that the proposed FNP ITS are acceptable.

fw Th The NRC staff has also reviewed the plant-specific changes to CTS as b+ d in this evaluation. l On the basis of the evaluations described herein tiifeech of the changes lthe NRC staff concludes that these changes are acceptable. si' p ,,,

1

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The Commission has concluded, based on the considerations' discussed above, that: (1) there is reasonable assurance that the health arif safety of the public'will not be endangered by operation in the proposed manner; (2) such activlbes will be conducted in compliance with the Commission's regulations; and, (3) the issuance of the amendments will not be inimical to the common defense and security or to the health anil safety of the public.

A  ;?L - ,

Principal Contributors: C. Schullen J A. Chu %g" H. Balukjian W' LM. Westonf "WiOrdaz M. Padovan R.Tjader TF. Ashe G. Hsii

/ TJ.-Luehman H. Garg. C. Liang R. Giardina M?.Ashley Date: August 6,1999 W ,-

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