ML20206U613

Forwards Draft SER on Proposed Conversion of Current TSs for Wolf Creek Generating Station to Improved Tss.Encl Draft SER Being Provided for Review to Verify Accuracy & to Prepare Certified Improved TSs

ML20206U613
Person / Time
Site:	Wolf Creek
Issue date:	02/02/1999
From:	Donohew J NRC (Affiliation Not Assigned)
To:	Maynard O WOLF CREEK NUCLEAR OPERATING CORP.
References
TAC-M98738, NUDOCS 9902160318
Download: ML20206U613 (195)

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W UNITED STATES s NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30086 0001 p

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Mr. Otto L. Maynard President and Chief Executive Officer l Wolf Creek Nuclear Operating Corporation  ;

l Post Office Box 411 l Burlington, Kansas 66839

SUBJECT:

DRAFT SAFETY EVALUATION REGARDING PROPOSED CONVERSION TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS FOR WOLF CREEK GENERATING STATION (TAC NO. M98738)

Dear Mr. Maynard:

Encum a 1 is the draft Safety Evaluation (SE) on your proposed conversion of the current )

L Technmal Specifications (CTSs) for the Wolf Creek Generating Station (WCGS) to the l Improved Technical Specifications (ITSs). The ITSs are based on the CTSs, NUREG-1431, l " Standard Technical Specifications, Westinghouse Plants" Revision 1, dated April 1995, and on i guidance provided in the Commission's ' Final Policy Statement on Technical Specifications I improvements for Nuclear Power Reactors," published in the Federal Reaister on July 22,1993 (58 FR 39132).

l The enclosed draft SE, including five tables attached to the SE that list the changes to the l CTSs, is based on the staff's review of your application dated May 15,1997 (ET 97-0050), as supplemented by letters in 1998 dated June 30 (ET 98-0049), August 5 (WO 98-0078),

August 28 (ET 98-0071), September 24 (ET 98-0078), October 16 (ET 98-0085), October 23 l (ET 98-0087), November 24 (WO 98-0105), December 2 (ET 98-0098), December 17 (ET 98-0102), and December 21 (ET 98-0107). These letters were your responses to the staff's requests for additional information (RAls) dated May 22, June 17, July 7, July 14, July 15, July 17, July 22, and August 14,1998. The staff has had additional RAls in its letters of l September 3 and October 7,1998, and there were also the meeting summaries that were issued on August 28, October 16, and November 6,1998.

I The enclosed draft SE is being provided for your review to verify its accuracy and to prepare the

, certified ITSs for WCGS to be submitted to NRC for issuance in the conversion amendment. i l The beyond scope issues are addressed in Section 3.G of the draft SE. You are requested to i provide your comments on the draft SE in writing, and a certified ITSs and Bases to the ITSs, I within 30 days of receipt of this letter or with the update to your application based on the  ;

responses to the staff's RAls, whichever date is later. After the staff has reviewed your {

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comments, it will incorporate changes, as appropriate, in the final SE before issuing the ITSs ,

and the final SE by amendment. The conclusions of the NRC staff in the enclosed draft SE are not valid until the final SE is issued. o\

The draft SE has additional information needed by the staff that is identified (by bold type) in the SE itself and in the attached tables. An electronic copy of the draft SE and the tables has been provided to your staff. You are also requested to provide the additional informat!on i needed by the staff with your comments on the draft SE.

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Mr. Otto L. Maynard February 2,1999 You are also requested to submit a license condition for Appendix D to the WCGS license to make enforceable the transfer of those requirernents in the CTSs being relocated into licensee-controlled documents (i.e., documents, such as the WCGS Updated Safety Analysis Report (USAR), for which changes to the documents by licensees are controlled by the regulations, in the case of the USAR,10 CFR 50.59, or the ITS Section 5.0) for the ITS conversion, as described in your letters and the enclosed draft SE. Enclosure 2 contains an acceptable license condition. A similar license condition to Enclosure 2 should also be submitted for (1) each commitment to complete a future action that you have included in your above letters on the ITSs for WCGS, and (2) the first performance of new and revised surveillance requirements (SRs) for the ITSs to be related to the implementation of the ITSs. An acceptable license condition for the new and revised SRs is provided in Section 6 of the enclosed draft SE and Enclosure 2.

Please do not hesitate to contact me at 301-415-1307 (or jnd@nrc. gov on the Internet) if you have any questions.

@ e ack N. Donohew, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation Docket No. 50-482 DISTRIBUTION:

Docket OGC

Enclosures:

1. Draft Safety Evaluation PUBLIC ACRS

2. Acceptable License Condition PDIV-2 Reading WBeckner EAdensam (EGA1) EPepon cc w/encis: See next page WBateman KThomas/CPoslusny JDonohew

/ WJohnson, RIV Document Name: LTR-DSE.CP / i OFC: PM/PD4-1 LNPD4 2 PM PDd PM PD41 BC-TSB nh PD/PDIV 1

) NAME: JDr EPeyton C sny TPohch WBeckn DATE: 1/ 99 1/ /99 1/ [/99 1 / I /99 hl S /99 h/b /99 conc: BSls BSic COPY YESWO YESNO YES/NO YESWO YESNO YES/NO OFC- BC HOMB \C HICB [ BC SRXB BC SCSB BC EELB NAME: J TCollins CBerlinger JCalvo DATE: f/ / /99 l/ 19 9 l /h/99 klj /99

[ 5-M CN 445-LS-31 CN 710-LS-9 N 2 A CN 2-27-M COPY: YE YES/NO YES/NO YES/NO YES/NO OFFICIAL RECORD COPY

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Mr. Otto L. Maynard -3 February 2,1999 cc w/encls:

Jay Silberg, Esq. Chief Operating Officer Shaw, Pittman, Potts & Trowbridge Wolf Creek Nuclear Operating Corporation 2300 N Street, NW P. O. Box 411 Washington, D.C. 20037 Burlington, Kansas 66839 Regional Administrator, Region IV Supervisor Licensing U.S. Nuc! ear Regulatory Commission Wolf Creek Nuclear Operating Corporation 611 Ryan Plaza Drive, Suite 1000 P.O. Box 411 Arlington, Texas 76011 Burlington, Kansas 66839 Senior Resident inspector U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident inspectors Office P. O. Box 311 8201 NRC Road Burlington, Kansas 66839 Steedman, Missouri 65077-1032 Chief Engineer Utilities Division Kansas Corporation Commistion 1500 SW Arrowhead Road Topeka, Kansas 66604-4027 Office of the Governor State of Kansas Topeka, Kansas 66612 Attorney General Judicial Center 301 S.W.10th 2nd Floor Topeka, Kansas 66612 County Clerk Coffey County Courthouse Burlington, Kansas 66839 Vick L. Cooper, Chief Radiation Controi Program Kansas Department of Health and Environment Bureau of Air and Radiation Forbes Field Building 283 Topeka, Kansas 66620

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001 g

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DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. XXX TO FACILITY OPERATING LICENSE NO. NPF-42 l WOLF CREEK NUCLEAR OPERATING CORPORATION l

WOLF CREEK GENERATING STATION l DOCKET NO. 50-482 1

1.0 INTRODUCTION

Wolf Creek Generating Stetion (WCGS) has been operating with Technical Specifications (TS) )

issued with the original operating license on June 4,1985, as amended from time to time. By application dated May 15,1997, as supplemented by letters dated June 30, August 5, August 28, September 24, October 16, October 23, November 24, December 2, December 17, and December 21,1998, Wolf Creek Nuclear Operating Corporation (the licensee) proposed to convert the current Technical Specifications (CTS) for both units to the improved Technical Specifications (ITS). The conversion is based upon:

NUREG-1431, " Standard Technical Specifications [STS), Westinghouse Plants,"

Revision 1, dated April 1995,

. Commission Final Policy Statement, "NRC Final Policy Statement on Technical Specification improvements for Nuclear Power Reactors," published on July 22, 1993 (58 FR 39132), and

. 10 CFR 50.36," Technical Specifications," as amended July 19,1995 (60 FR 36953).

The overall objective of the conversion, consistent with the Final Policy Statement, is to rewrite, l reformat, and streamline the TS for WCGS to be in accordance with 10 CFR 50.36. The NRC l staff acknowledges that, as indicated in the Final Policy Statement, the conversion to STS is a voluntary process. Therefore, it is acceptable that the ITS differs from STS, reflecting the current licensing basis for the WCGS. '

1 In addition to basing the ITS on the STS, the Commission's Final Policy Statement, and the l requirements in 10 CFR 50.36, the licensee retained portions of the CTS as a basis for the ITS.

Plant-spe,cific issues, including design features, requirements, and operating practices, were WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION l

l l

l discussed with the licensee during a series of conference calls and meetings. Meetings were l held with the licensee during the weeks of August 17, September 14, and October 12,1998, and meeting summaries were issued on August 28, October 16, and November 6,1998, respectively.

Based on these discussions, the licensee has proposed specifications that were not in the STS or the CTS. For proposed specifications that were generic to the STS, the NRC staff requested that the licensee submit the generic revised technical specifications as a proposed change to the STS through the NRC/ Nuclear Energy institute's Technical Specifications Task Force (TSTF). Proposed changes to the STS, or NUREG-1431, are identified by the acronym TSTF and a number as, for example, TSTF-111. For proposed specifications that were plant-specific, the changes are beyond scope issues for the conversion and are addressed separately in Section 3.G of this Safety Evaluation (SE). The licensee has identified several such generic and plant-specific changes in its application for the ITS conversion.

Consistent with the Final Policy Statement, the licensee also proposed transferring some CTS requirements to licensee-controlled documents (such as the updated safety analysis report (USAR) for the WCGS, for which changes by licensees to the documents are controlled by a regulation such as 10 CFR 50.59 and may be able to be changed without prior staff approval).

NRC-controlled documents, such as the TS, may not be changed by the licensee without prior staff approval. 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 appropriate scope of the ATS. Further, significant changes were proposed to the Bases to make each ITS requirement clearer and easier to understand.

Since the May 15,1997, application was submitted, Amendment Nos. XX through XX for WCGS were approved. The licensee has incorporated these amendments as appropriate into the ITS.

The NRC staff's evaluation of the application included the supplements listed above that resulted from staff requests for information (RAls) and discussions with the licensee during the NRC review. The staff issued RAls in the letters dated May 22, June 16, June 17, July 7 July 9, July 15, July 17, July 21, August 14, September 3, and October 7,1998, and the meeting summary issued October 16,1998.

The Commission's proposed action on the WCGS application for amendment dated May 15,1997, was published in a notice of consideration of issuance of amendment to the WCGS operating license in the Federal Register on October 5,1998 (63 FR 53471).

During its review, the NRC staff relied on the Final Policy Statement and the STS as guidance for acceptance of CTS requirements into the ITS. This SE provides a summary basis for the NRC staff conclusion that the licensee can develop an ITS for WCGS based on the STS, as modified by plant-specific changes, and that the use of the ITS is acceptable for continued operation of WCGS. These plant-specific changes serve to clarify the iTS with respect to the guidance in the Final Policy Statement and STS. The SE also explains the NRC staff's WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

conclusion that the ITS is consistent with the WCGS current licensing basis and the requirements of 10 CFR 50.36.

Hereafter, the proposed or improved TS for the WCGS are the ITS, the existing or current TS are the CTS, and the improved standard TS, NUREG-1431 for WCGS, are the STS or NUREG-1431. The corresponding TS Bases are ITS Bases, CTS Bases, and STS Bases, respectively.

2.0 BACKGROUND

Section 182a of the Atomic Energy Act requires that applicants for nuclear power plant operating licenses will state:

l (S]uch technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization . . . of special nuclear material will be 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 doing 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 are 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 related to station operation: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation  ;

(LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be inciuded in a plant's TS.

For several years, NRC and industry representatives have sought to develop guidelines for ,

improving the content and quality of nuclear power plant TS. On February 6,1987, the Commission issued an interim policy statement on TS improvements, " Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (52 FR 3788). During the period from 1989 to 1992, the utility Owners Groups and the NRC staff developed improved STS, such as NUREG-1431 for Westinghouse plants, that would establish models of the Commission's policy for each primary reactor type. In addition, the NRC staff, licensees, and Owners Groups developed generic administrative and editorial guidelines in the form of a

" Writer's Guide" for preparing TS, which gives greater consideration to human factors principles and was used throughout the development of licensee-specific ITS.

WOLF CREEK GENERATING STATION DRAFT SAFETY EvALutTioN l

In September 1992, the Commission issued NUREG-1431, which was developed using the guidance and criteria contained in the Commission's Interim Policy Statement. The STS in NUREG-1431 were established as a model for developing the STS for Westinghouse plants in general. The STS reflect the results of a detailed review of the application of the interim policy statement criteria to generic system functions, which were published in a " Split Report" issued to the nuclear steam system supplier (NSSS) owners groups in May 1988. The STS also reflect the results of extensive discussions concerning various drafts of STS, so that the application of the TS criteria and the Writer's Guide would consistently reflect detailed system configurations and operating characteristics for all NSSS designs. As such, the generic Bases presented in NUREG-1433 provide an abundance of information regarding the extent to which the STS present requirements that are necessary to protect public health and safety. The STS in NUREG-1431 apply to the WCGS.

On July 22,1993, the Commission issued its Final Policy Statement, expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36 (58 FR 39132). The Final Policy Statement described the safety benefits of the STS, and encouraged licensees to use the STS as the basis for plant-specific TS amendments, and for complete conversions to ITS based on the STS. Further, the Final Policy Statement gave guidance for evaluating the required scope of the TS and defined the guidance criteria to be used in determining which of the LCOs and associated surveillances should remain in the TS. The Commission noted that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TS, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland GeneralElectric Co. (Trojan Nuclear Plant), ALAB-531,9 NRC 263,273 (1979). There, the Appeal Board observed:

[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 legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern 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 necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

By this approach, existing LCO requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TS; those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The Commission codified the four criteria in 10 CFR 50.36 (60 FR 36953, July 19,1995). The four criteria are as follows:

J WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

5-Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

Section 4.0 of this SE explains the NRC staff's conclusion that the conversion of the CTS to the ITS based on STS, as modified by plant-specific changes, is consistent with the WCGS current licensing basis, and the requirements and guidance of the Commission's Final Policy Statement and 10 CFR 50.36.

3.0 UTILITIES JOINT EFFORT This conversion is a joint effort in concert with three other utilities: Pacific Gas & Electric Company for Diablo Canyon Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323); TU Electric for Comanche Peak Steam Electric Station, Units 1 and 2 (Docket Nos. 50-445 and 50-446); and Union Electric Company for Callaway Plant (Docket No. 50-483). It is a goal of the four utilities to make the ITS for their plants as similar as possible. This group of four utilities was designated the four loop owners group (FLOG).

This joint effort includes a common methodology for the licensees in marking-up the CTS, STS, and STS Bases, that has been accepted by the staff. This common methodology is discussed at the end of Enclosure 2," Mark-Up of Current TS"; Enclosure Sa," Mark-Up of NUREG-1431 Specifications"; and Enclosure 56," Mark-Up of NUREG-1431 Bases," for each of the 14 separate ITS sections that were submitted with the licensee's application. For each of the ITS l sections, there is also the following enclosures:

l l

l WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

= Enclosure 1, " Cross-Reference Tables," the cross-reference table connecting each CTS specification (i.e., LCO, required action, or SR) to the associated ITS specification, sorted by both CTS and ITS specifications.

- Enclosures 3A and 3B," Description of Changes to Current TS" and " Conversion Comparison Table," the description of the changes to the CTS section and the comparison table showing which plants (of the four licensees in the joint effort) that each change to the CTS applies to.

. Enclosure 4,"No Significant Hazards Considerations," the no significant hazards consideration (NSHC) of 10 CFR 50.91 for the changes to the CTS with generic NSHCs for administrative, more restrictive, to be relocated, and to be moved-out-of-CTS changes, and individual NSHCs for less restrictive changes.

. Enclosures 6A and 6B, " Differences From NUREG-1431" and " Conversion Comparison Table," the descriptions of the differences from NUREG-1431 Specifications and the comparison table showing which plants (of the four licensees in the joint effort) that each difference to the STS applies to.

The common methodology includes the convention that, if the words in an CTS specification are not the same as the words in the ITS specification, but the CTS words have the same meaning or have the same requirements as the words in the ITS specification, then the licensees do not have to indicate or describe a change to the CTS. In general, only technical changes have been identified; however, some non-technical changes have also been identified when the changes cannot be easily be determined. The portion of any specification which is being deleted is struck through (i.e., the deletion is annotated using the strike-out feature of the word processing computer program or crossed out by hand). Any text being added to a specification is shown by shading the text, placing a circle around the new text, or by writing the text in by hand. The text being struck through or added is shown in the marked-up CTS and STS pages in Enclosures 2 (CTS pages) and 5 (STS and STS Bases pages) for each ITS section attachment to the application. Another convention of the common methodology is that the technical justifications for the less restrictive changes are included in the NSHCs.

The changes to the CTS are identified by change numbers that are listed in Enclosure 3 and are determined by the convention discussed at the end of Enclosure 2. The change number is of the form 4-13-A, where the first number is a prefix number (i.e., the 4 of 4-13-A) assigned to each specification (or group of similar specifications) within an CTS section, as for example CTS 3/4.6, containment systems, such that it refers to the same specification for each utility regardless of the actual specification number in their individual plant CTS. The second number (i.e., the 13 of 4-13-A) identifies the change within the given specification or group of specifications (these are changes having the same prefix number); however, the second number does not denote the sequence of the changes within the given specification or group of specifications. For example, the change 4-03-X may not follow change 4-02-X in the CTS specification, or group of specifications, denoted by the prefix 4. The changes through the CTS specifications may not be in the same sequence r.s given by the second number. The letter suffix (i.e., the A of 4-13-A) identifies one of the following types of change:

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATloN

"A" for administrative changes.

. "M" for more restrictive changes. i

. "LS" or "TR" for CTS requirements that are relaxed or eliminated, or for which l

new operational flexibility is added to the ITS compared to the CTS. t "R" for changes to relocate CTS requirements, which do not meet the 10 CFR l 50.36a criteria, to appropriate licensee-control'ed documents.

"LG" for CTS descriptions and details, not requirements, that are relocated to appropriate licensee-controlled documents.

For the case where the same change to the CTS is being proposed by more than one of the licensees, then these licensees use the same change number to identify the change and the i other licensees, not proposing the change, list the change number but state "not applicable" in 1 the description of the change. For example, change 01-07 LG for ITS 3/4.2 is a change to relocate surveillance frequencies to licensee-controlled documents and is proposed by all the licensees (see Enclosure 3B of the licensee's application). For change 01-03-LG in the same ITS section, only the licensee is proposing the change and this change is "not applicable" to the other licensees. There may be cases, where most of the identified changes for an (TS section may not be applicable to a specific licensee because these changes do not need to be made to the CTS for that licensee.

i l

The licensee may have more than one less-restrictive change in the same ITS section with the same "LS" number or 'TR" number. Because these "LS" and "TR" numbers refer to specific NSHCs provided in Enclosure 4 to the application for an ITS section, these less-restrictive

! changes are the not same change, but they are the same type of "LS" or "TR" change and have the same NSHC.

As a result of differences between the individual CTS for the FLOG, and because of changes to the CTS that may occur after the initial assignments of change identifiers, the change numbers may not appear sequentially in the CTS markup. Also, the second number is assigned sequentially independent of the type of change that is identified. Therefore, change 4-12 M may be listed before 4-13-A and after 4-11-LG.

The type of change also identifies the type of NSHC provided in Enclosure 4. The NSHCs for the A, M, R, and LG changes are generic and only one NSHC is provided for each of these types of changes in Enclosure 4. The NSHCs for LS and LG changes are individual and a suffix number is assigned for each such change, for example,4-13-LS-1 or 4-13-LG-2, where the first LS change or second LG change is identified. The change number listed in Enclosure 3 that was assigned to these LS and LG would also include the suffix number, as change 4 LS-1 or change 4-13-LG-2. These change numbers are included in the tables attached to this SE to identify the changes described in the tables. There are tables for each type of change listed above.

WOLF CREEK GENERATING sTATloN DRAFT SAFETY EVALUATION

4.0 EVALUATION The NRC staff's ITS review evaluates changes to CTS that fall into five categories defined by the licensee and includes an evaluation of whether existing regulatory requirements are adequate for controlling future changes to requirements removed from the CTS and placed in licensee-controlled documents. This evaluation also discusses the NRC staff's plans for monitoring the licensee's implementation of these controls at WCGS.

The NRC staff review also identified the need for clarifications and additions to the application in order to establish an appropriate regulatory basis for translation of CTS requirements into the ITS. Each change proposed in the amendment request is identified as either (1) a description of change (DOC), identified by a change number (CN), to the CTS or (2) a difference from NUREG-1431, which is a justification for deviation (JFD) from the STS. The NRC staff comments were documented as RAls and issued in letters or meeting summaries to the licensee. These comments were intended to clarify the licensee's basis for translating the CTS requirements into ITS. The NRC staff finds that the licensee's submittals including responses to RAls provide sufficient detail to allow the staff to reach a conclusion regarding the adequacy of the licensee's proposed changes to the CTS.

The license amendment application was organized such that changes were included in each of the following CTS change categories, as appropriate:

(1) Administrative Changes, (A), i.e., non-technical changes in the presentation of CTS requirements; (2) Technical Changes - More Restrictive, (M), i.e., new or additional TS requirements; (3) Technical Changes - Less Restrictive (specific), (LS and TR), i.e., changes, deletions, and relaxations of CTS requirements; (4) Technical Changes - Less Restrictive (generic), (LG), i.e., deletion of CTS details by the relocation of information and requirements from existing specifications (that are otherwise being retained) to licensee-controlled documents, including the ITS Bases; and (5) Relocated Technical Specifications, (R), i.e., relaxations in which whole CTS specifications (the LCO, and associated action and SR) are removed from the CTS (an NRC-controlled document) and placed in licensee-controlled documents.

The changes that are in the ITS conversion for the WCGS for each of the above categories are listed in the following five tables attached to this SE:

Table A of Administrative Changes to Current Technical Specifications Table M of More Restrictive Changes to Current Technical Specifications wolf CREEK GENERATING STATION DRAFT SAFETY EVALUATION l

Table LS of Less Restrictive Changes to Current Technical Specifications (that also includes the TR changes)

Table LG of Details Relocated from Current Technical Specifications Table R of Relocated Current Technical Specifications These general categories of changes to the licensee's CTS requirements and STS differences may be better understood as follows:

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 editorialin 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 the licensee have used the STS as guidance to reformat and make other administrative changes.

Among the changes proposed by the licensee and found acceptable by the NRC staff are:

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

(2) identifying plant-specific wording for system names, etc.;

(3) changing the wording of specification titles in STS to conform to existing plant practices; (4) splitting up requirements currently grouped under a single current specification to more appropriate locations in two or more specifications of ITS; (5) combining related requirements currently presented in separate specifications of the CTS into a single specification of ITS; (6) presentation changes that involve rewording or reformatting for clarity (including moving an existing requirement to another location within the TS) but which do not involve a change in requirements; (7) wording changes and additions that are consistent with CTS interpretation and practice, and that more clearly or explicitly state existing requirements; (8) deletion of TS whose applicability has expired; and (9) deletion of redundant TS requirements that exist elsewhere in the TS.

Table A of WCGS administrative changes lists the administrative changes being made in the ITS conversion. TaNe A is organized in ITS order by each A-type DOC to the CTS, and provides a summary description of the administrative change that was made, and CTS and ITS references. The NRC staff reviewed all of the administrative and editorial changes proposed by wolf CREEK GENERATING STATION oRAFT SAFETY EVALUATION

the licensee and finds them acceptable because they are compatible with the Writer's Guide and the STS, do not result in any change in operating requirements, and are consistent with the Commission's regulations.

B. Technical Changes - More Restrictive The licensee, in electing to implement the specifications of the STS, proposed a number of requirements more restrictive than those in the CTS. The ITS requirements in this category include requirements that are either new, more conservative than corresponding requirements in the CTS, or that have additional restrictions that are not in the CTS but are in the STS.

Examples of more restrictive requirements are placing an LCO on plant equipment which is not required by the CTS to be operable, more restrictive requirements to restore inoperable equipment, and more restrictive SRs. Table M of WCGS more restrictive changes lists the more restrictive changes being made in the ITS conversion. Table M is organized in ITS order by each M-type DOC to the CTS and provides a summary description of the more restrictive change that was adopted, and the CTS and ITS references. These changes are additional restrictions on plant operation that enhance safety and are acceptable.

C. Technical Changes - Less Restrictive (Specific)

Less restrictive requirements include changes, deletions and relaxations to portions of the CTS requirements that are not being retained in ITS. When requirements have been shown to give little or no safety 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 NRC staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the Owners Groups comments on the 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 WCGS design was also reviewed to determine if the specific design basis and licensing basis for the WCGS are consistent with the technical basis for the model requirements in the STS, and thus provide a basis for the ITS.

A significant number of changes to the CTS involved changes, deletions and relaxations to portions of the CTS requirements evaluated in Categories I through Vill as follows:

Category 1 -

Relaxation of CTS LCO Applicability Category II -

Relaxation of CTS Surveillance Frequency Category 111 - Relaxation of CTS Action Requirements CategoryIV - Relaxation of CTS Required Action Completion Time l

Category V -

Relaxation of CTS Surveillance Requirement Acceptance Criteria Category VI - Relaxation of CTS Action Entry to Perform SRs l WOLF CREEK GENERATING sTATloN DRAFT SAFETY EVALUATION l

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Category Vil - Deletion of Requirements Contained in Regulations and of Explicit Post Maintenance SRs  !

Category Vill - Relaxation of LCO Requirements I'

The following discussions address why various specifications within each of these eight categories of information or specific requirements are not required to be included in ITS.

1 Relaxation of CTS LCO Acolicability (CategoryI)

Reactor operating conditions are used in the CTS to define when the LCO is required to be met. The LCO applicabilities can be specifically defined terms of reactor modes of '

operation: hot shutdown, cold shutdown, reactor critical, or power operating condition.

Applicabilities can also be more general. Depending on the circumstances, CTS may I require that the LCO be maintained within limits in "all modes" or "any operating mode."

However, generalized applicability conditions are not contained in the STS, therefore the I

ITS eliminate the 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 required features.

In another application of this type of change, CTS requirements may be eliminated during conditions 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 can be satisfied by exiting the applicability thus taking the plant out of the conditions inat require the safety system to be operable. Therefore, changes falling within Category I are acceptable.

. Relaxation of CTS Surveillance Freauency (Categoryll) 1 CTS and ITS surveillance frequencies specify time interval requirements for performing surveillance testing. Increasing the time interval between surveillance tests in the ITS results in decreased equipment unavailability because of testing. In general, the STS contain surveillance 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 other TS requirements which provide regular checks to ensure limits are met.

Reduced testing can enhance safety because it reduces system unavailability from testing; in turn, reliability of the affected structure, system or component should remain constant, or may increase because of fewer testing challenges to the system. Reduced testing is acceptable where operating experience, industry practice, or industry standards, such as manufacturers' recommendations, have shown that components WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

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usually pass the surveillance when performed at the specified interval. Therefore, the i frequency is acceptable from a reliability standpoint. Surveillance frequency changes to l incorporate alternate train testing has been shown to be acceptable where other qualitative or quantitative test requirements are required which are established predictors of system performance (e.g., a 31-day air flow test is an indicator that positive i pressure in a controlled space will be maintained because the test would use the same fans as the less frequent ITS 36-month pressurization test and industry experience shows that components usually pass the pressurization test). Additionally, surveillance j frequency relaxation can be based on staff approved topical reports. The NRC staff has accepted topical report analyses that bound the plant-specific design and component reliability assumptions. Therefore, changes falling within Category ll are acceptable.

. Relaxation of CTS Action Reauirements (Category lli) l Upon discovery of a failure to meet an LCO, the STS specify required actions to complete for the associated TS conditions. Required actions of the associated conditions are used to establish remedial measures that must be taken in response to the degraded conditions. Adopting required actions from the STS is acceptable because STS-required actions take into account the operability status of redundant systems of TS-required features, the capacity and capability of the remaining features, and the compensatory attributes of the required actions as compared to the LCO requirements. In conjunction with the relaxation of the applicability of several CTS specifications (Type I changes), the associated action requirements to exit the applicability are also relaxed. Such relaxations of action requirements are acceptable because they are commensurate with industry standards for reductions in thermal power in an orderly fashion without compromising safe operation of the plant. Therefore, changes falling within Category lll are acceptable.

. Relaxation of CTS Reouired Action Comotetion Time (Category IV) l Upon discovery of a failure to meet an LCO, the STS specify times for completing l required actions of the associated TS conditions. Required actions of the associated conditions are used to establish remedial measures that must be taken within specified completion times. These times define limits during which operation in a degraded ,

condition is permitted. Adopting completion times from the STS is acceptable because completion times take into account the operability status of the redundant systems of TS-required features, the capacity 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. Therefore, changes falling within Category IV are acceptable.

. Relaxation of CTS Surveillance Reauirement Acceptance Criteria (Category V)

The CTS require safety systems to be tested and verified operable prior to entering applicable conditions. The ITS provide the additional requirement to verify operability by actual or test conditions. Adopting the STS allowance for " actual" conditions is WOLF CREEK GENERATING STATION ORAFT SAFETY EVALUATloN

l l acceptable because TS-required features cannot distinguish between an " actual" 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 the format preference in the STS is acceptable because SRs that remain include testing of all previous features required to be verified operable.

The identification of the specific signal for safety system testing may be listed in the l CTS; however, this detail is not necessary for inclusion in the TS to ensure operability of l the associated systems. This detail will be relocated to the ITS Bases where changes i are controlled by the ITS Bases control program in ITS 5.5.14. The ITS require that changes to the Bases may be without prior staff approval only if the changes meet the criteria in 10 CFR 50.59, which is the same criteria used to control changes to the i description of the plant in the WCGS USAR. The ITS Bases is an acceptable licensee-controlled document for this detail. Therefore, changes falling within Category V are acceptable.

. Relaxation of CTS Action Entry to Perform SRs (Category VI)

The STS allows an instrument channel to be placed in an inoperable status solely for the performance of required surveillance testing, without entering the associated Conditions and Required Actions, provided the associated function maintains trip capability. This allowance is generclly six hours, during which time the functional capability is maintained. This relaxation could be in accordance with approved Topical Reports that apply to WCGS. Adopting this STS approach to action entry during surveillance testing is acceptable because it takes into account the capability of the specified function, time for required test completion, and the extremely low probability of a design basis event occurring during the test period. Therefore, changes falling within Category VI are acceptable.

• Deletion of Reauirements Contained in Reculations and of Exolicit Post Maintenance SRs (Category Vil)

Some requirements contained in the regulations have also been included in plant TS. If these requirements are in the regulations, they will apply to the licensee's operation of the plant whether or not they are in the TS. Therefore, these requirements do not need to be included in the TS. Also, plant TS have included specific requirements on performing surveillances prior to returning equipment or systems to service following maintenance, repair or replacement. Explicit post-maintenance TS surveillances requirements do not have to be included in the TS because these requirements are adequately addressed by administrative post-maintenance programs and the definition of operability. These deletions are acceptable because they are not important to ensure the iTS's effectiveness. In addition, omitting this information from the ITS is acceptable because it will continue to be contained in appropriate station procedures required by ITS 5.4.1. Therefore, changes falling within Category Vil are acceptable.

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

Table LS of WCGS less restrictive changes is organized in ITS order by each LS-type or TR-type CN to the CTS, and provides a summary description of the less restrictive change that was made, the CTS and ITS references, and a reference to the applicable change categories as discussed above (if applicable). If the change category does not apply, the word " unique"is given. For ease of reference, the 8 less restrictive change categories are listed at the bottom of each page of Table LS.

The changes identified as unique are evaluated separately below. Each evaluation below is preceded by the ITS section or specification and the CN identifier (e.g., LS-1 or TR-1) associated with the change. All of these changes to the CTS are consistent with the STS and, l therefore, are not beyond-scope issues for the ITS conversion. The changes that are beyond-scope issues for the ITS conversion are addressed in Section 3.G of this SE.

ITS Section 1.0 LS-1 The STS definition of core alterations is proposed for the ITS and is less restrictive than the corresponding CTS definition because it will or.ly apply to those activities that create the potential for a reactivity excursion and thus warrant special precautions or controls in the ITS. The ITS definition will apply to fewer activities. The iTS definition will restrict core alterations to the movement of fuel, sources, or reactivity control components which may cause significant reactivity changes in the core. Under the revised definition, in-vessel movement of instruments, cameras, lights, tools, etc., will not be considered to be core alterations. This change is acceptable because special controls on components other than fuel, sources, or reactivity contro! components to prevent reactivity excursions are not warranted. In addition, the proposed definition adds an allowance that suspension of core alterations shall not preclude completion of movement of a component to a safe position. This is acceptable because it is not desirable to immediately stop moving a component (e.g., stop the movement with the component suspended from the refueling grapple over the core).

LS-2 The CTS Table 1.2 is proposed to be revised in the following manner: (1) notation "NA" replaced "O" under % rated thermal power (RTP) for Modes 3, 4,5, and 6, (2) notation

• NA" replaced the reactor coolant temperature for Modes 1,2, and 6, (3) notation "NA" replaced the reactivity condition for Mode 6, (4) a new note b has been added to Modes WOLF CREEK GENERATING sTATloN DRAFT SAFETY EVALUATION

i 4 and 5 stating that the required reactor vessel head closure bolts are fully tensioned, and (5) a new note c replaced the note applied to Mode 6 and stating that the required reactor vessel head closure bolts are less than fully tensioned. These changes are administrative, resulting in no technical changes, except the new notes b and c which i relaxes the definition of Mode 6. The CTS table is revised such that the required reactor vessel head closure bolt requirements for Modes 4,5 and 6 are clarified. Currently a footnote applicable only to Mode 6 defines that mode, in part, by reference to " vessel head closure bolts less than fully tensioned." That footnote does not specify the ,

transition point between Modes 5 and 6 with regard to the number of vessel head '

closure bolts that must be fully tensioned, leaving the issue open to interpretation. The proposed change provides the necessary clarification by adding a footnote to Modes 4 and 5, consistent with the approach used in NUREG-1431, to define those modes as having the required number of reactor vessel head closure bolts fuliy tensioned. The transition point between Modes 5 and 6 would also be clarified as occurring when the i required reactor vessel head closure bolts are less than fully tensioned. The required l number of closure bolts, which may be less than the total number, is established by analysis that demonstrates adequate O-ring compression to prevent leakage and ensures that ASME Section lli stress limits for affected components are not exceeded.

Therefore, the proposed changes are acceptable.  ;

iTS Section 3.0 LS-2 The STS LCO 3.0.5 is proposed to be added to the ITS to provide an exception to ITS LCO 3.0.2. ITS LCO 3.0.2 states that, upon discovery of a failure to met an LCO (i.e.,

equipment is inoperable), the required actions of the LCO shall be met. The LCO 3.0.5 exception is for instances where restoration of the inoperable equipment to an operable status could not be performed while continuing to comply with the required actions for an LCO. Many LCO actions require an inoperable component to be removed from service and an exception to these actions is necessary to allow the performance of SRs to either demonstrate the operability of the equipment being returned to service or to demonstrate the operability of other equipment.

LCO 3.0.5 is necessary to establish an allowance that is not formally recognized in the CTS. Without this allowance, certain components could not be restored to operable status and a station shutdown would ensue. Clearly, it is not the intent or desire that the TS preclude the return to service of a component to confirm its operability. This allowance is deemed to represent a more stable, safe operation than requiring a station shutdown to complete the restoration and confirmatory testing. The time during which the equipment is returned to service is very small, therefore, the probability of an accident during that time period is also very small and insignificant. Therefore, the proposed STS LCO 3.0.5 is acceptable.

ITS Specification 3.2 LS-9 The proposed change will delete the requirement to reduce the power range neutron flux - high reactor trip setpoints when AFD is outside its limits. With the AFD not within wolf CREEK GENERATING STATION DRAFT SAFETY EVALuATloN

the limits specified in the core operating limits report (COLR), the CTS required that the thermal power be reduced to less than 50% RTP and that the power range neutron flux

- high reactor trip setpoints be reduced to less than or equal to 55% RTP within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The requirement to reduce the high neutron flux reactor trip setpoints is proposed to be deleted to be consistent with NUREG 1431. The proposed change is considered to be acceptable because reducing the power level to < 50% RTP maintains the plant in a relatively benign condition where the axial flux distribution is not a significant accident analysis input and the probability of an accident or transient that would cause an increase in reactor power, for which reduced trip setpoints would provide additional protection, is very low. There are no AFD limits below 50% RTP. A rise to a 50% RTP with AFD outside the limits does not immediately create an unacceptable situation, and a reactor trip, or timely operator action depending on the I rapidity of the power transient, would successfully terminate the event. Rapid power excursions resulting from events that are peaking factor limiting would likely be terminated by normal reactor trip signals or reactor trip on safety injection. Slower ,

transients would be terminated by the operator or, for large AFD deviations, the  !

overtemperature 4T reactor trip. The overtemperature 6T reactor trip setpoint is l automatically reduced when axial flux distributions deviate sufficiently from the required l operating area. Thus, reducing the high neutron flux setpoint provides an additional level of accident mitigation which would not be necessary in most cases. Based on this, l the proposed change is acceptable.

LS-13 The proposed change will delete the CTS actions for the quadrant power tilt ratio (OPTR) being outside its limit, and the ratio being either below or above 1.09. The ITS conditions would be only for the OPTR being outside its limit, to be consistent with NUREG-1431. Actions involving OPTRs of 1.09 would be eliminated in conformance with NUREG-1431. While the requirements in CTS regarding OPTRs in excess of 1.09 due to misalignment of control rods would be addressed by the ITS requirements associated with rod group misalignment limits, the CTS Actions regarding OPTRs in excess of 1.09 due to other causes would be replaced by less restrictive requirements.

The CTS require that the OPTR be calculated once per hour and that power be reduced to less than 50% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the power range neutron flux high trip setpoint be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In addition, the CTS require identification and correction of the cause of the tilt condition and periodic verification that OPTR is within limits during any subsequent ascension to RTP. The ITS would require (1) that the OPTR be calculated only once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, (2) only a 3% RTP reduction for each 1%

of OPTR in excess of 1.0 and no reduction in flux trip setpoints, and (3) verification of peaking factors prior to and following power ascension and reevaluation of safety analyses prior to power ascension. The licensee stated that the proposed change are acceptable because:

(1) The OPTR would be expected to change slowly over time so a less frequent calculation of OPTR would be acceptable; (2) Once the operating staff commences a power reduction,in accordance with ITS WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

1 requirements, the effect of any flux tilt will tend to be mitigated by reducing the flux and establishing greater margin to fuel design limits, the reduction of power required by the ITS would result in a plant transient that generally would be less severe than the reduction to less than 50% as required by CTS, and eliminating the trip setpoint reduction is acceptable because a OPTR in excess of limits does not necessarily imply that accident analyses assumptions have been violated; and (3) The ITS Required Actions prior to and subsequent to power ascension provide assurance that power operation at or near RTP will be in accordance with the safety analyses.

Based on this, the proposed change is acceptable. l LS-14 The proposed change will (1) extend the completion time from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for resetting the power range neutron flux - high trip setpoints during the power reduction following QPTR measurements, (2) clarify that power reductions must be completed l within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after each OPTR measurement and that equilibrium conditions must be  ;

achieved for measuring peaking factors, and (3) delete the action to requiring the OPTR l to be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee stated that the CTS require a power I reduction within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of determining OPTR out of limit and a high flux trip setpoint reduction within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The completion time for resetting the high flux trip setpoints would be changed to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after determining QPTR is out of limit. This change would extend the allowed time to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for reducing power range neutron flux-high trip setpoints when peaking factors are out of limit. With the OPTR not within limit, the CTS require the thermal power and the power range neutron flux-high reactor trip setpoints to be reduced. Two hours are allowed for power reduction, and an additional four hours are allowed for the completion of the setpoint reduction. A completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to reduce the trip setpoints, as proposed, will allow time to reduce reactor power, perform the required OPTR determination, and permit an orderly resetting of the high flux trip channel setpoints while reducing the chances of an inadvertent reactor trip during these evolutions. During the trip setpoint change, there is increased potential for human error resulting in a plant transient. In addition, the reactor power would be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; this would provide additional margin to fuel design limits. The CTS actions requiring OPTR to be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce power to < 50% RTP and requiring verification of OPTR during return to full power operation would be eliminated in accordance with NUREG-1431. Also, the requirement to reset the power range neutron high-flux trip setpoints after a required reduction to s 50% RTP would be eliminated. The ITS would add new required actions for OPTR out of limit including requirements for measuring Fo(Z) and FL and performing safety analyses to verify peaking factors are acceptable prior to and following a return to power. The ITS would require a re-evaluation of the safety analyses prior to increasing reactor power above the reduced power required by the OPTR limit. Finally, the ITS would require a confirmation that Fo(Z) and FL are within limit following the power increase. The ITS focus on maintaining the peaking factors Fo(Z) and FL within limits rather than the OPTR. This is appropriate because OPTR is a monitored parameter WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

that is indicative of peaking factor problems. The ITS would require verification that Fo(Z) and FL are within limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions by performing SRs that can directly measure flux shapes in the core, if Fo(Z) or FL are not within limits, the Conditions for those TS will specify additional required actions. Since the peaking factors are of prime importance, the ITS will ensure that the power distribution remains consistent with the initial conditions assumed in safety analyses.

The ITS would retain the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> requirement to reduce power proportionally to the percent that OPTR exceeds its limit. This would result in a power reduction that would provide additional margin to fuel design limits during a flux tilt condition to assure that design limits are not challenged by local flux peaking. These design margins are set conservatively and provide further assurance that operation in accordance with the required actions during or beyond the 24-hour period would not challenge fuel design limits. The proposed change also would eliminate the requirement to reduce the setpoints to s 55% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of reaching 50% RTP would be eliminated. This change is acceptable on the basis that the ITS would still required the power range neutron flux - high trip setpoints to be reduced during the power reduction as discussed above; and the acceptablility of changing the completion time from four to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> has been addressed previously. Based on this, the proposed change is acceptable.

ITS Snecification 3.3 LS-1 The proposed change is exchange the active verb in the CTS SRs from " demonstrated" to " verified" to allow reactor trip system (RTS) and engineered safety feature actuation gwm (ESFAS) sensor response time surverflance to be performed in accordance with approved WCAP-13632-P-A, Revision 2, and eliminate pressure sensor response time testing. The licensee stated that the applicability of the generic analysis of the WCAP report has been confirmed for Wolf Creek and that the specific transmitters installed at WCGS that require RTT are included in Table 9-1 of WCAP-13632. In addition, the licensee provided the following discussion that addressed the four actions raised in the NRC SER dated September 5,1995, that approved the WCAP report (a) A hydraul:c response time test will be performed on any new or refurbished transmitter, prior to declaring the affected channel operable, to determine an initial sensor-specific response time value.

(b) A hydraulic response time test will be performed on units that use capillary tubes after initialinstallation of replacement transmitters or following any maintenance or modification activity that could damage the capillary tubing or degrade the response time characteristics of installed sensors.

(c) WCGS does not utilize pressure sensors that incorporate a variable damping feature in the RTS or ESFAS channels that are required to have their response times verified.

(d) WCGS uses pressure sensors manufactured by Rosemount in applications that are required to be response time tested. The licensee's actions in response to WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

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NRC Bulletin 90-01 and its Supplement 1 have been completed and accepted by NRC in the SER dated May 15,1994. Licensee engineering, operations, and instrumentation and control (l&C) personnel are aware of the loss of fill oil phenomena applicable to Rosemount transmitters manufactured prior to July 11, 1989.

The licensee further stated that the CTS are revised to indicate that the system response time shall be verified using a sensor response time justified by the methodology described in WCAP-13632 P A Revision 2. Based on this, the proposed change is acceptable.

LS-8 The proposed change will add the option to reduce power to less than P-7 within 12 l hours, for the number of operable channels less than those required, which is not in the l CTS. The new footnote c is added to the applicable modes for the functions so that the l applicable modes are consistent with the added option. The change reflects a revision i to CTS Action 6. If the requirements in the action not met, LCO 3.0.3 would be entered.

This action is proposed to be revised to state that, if the action requirements are not met, thermal power must be reduced to below the P-7 interlock setpoint within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Most of the functional units that have Action 6, the pressurizer pressure -low, ,

pressurizer water level - high, reactor coolant flow - low, two loops (above P-7 and below P-8), RCP undervoltage, and RCP underfrequency, are automatically blocked below P-7 and an applicability note has been added accordingly. The reactor coolant flow - low (single loop) reactor trip function does not have to be operable below the P-8 setpoint; however, the acton must take the plant below the P-7 setpoint, if an inoperable channel is not tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, due to the shared components between this function and the reactor coolant flow - low (two loops) trip function. Based on this, the proposed change is acceptable. 4 LS-11 The proposed change will revise CTS notes e and g in Table 3.3-2, for functions #4.a.2,

1. 4.b, #4.c, #4.d, #4.e, #5.a, and #5.b. The condition of being deactivated is proposed to be added to note c for the steam line isolation functional units to state that the LCO requirements are not applicable in Modes 2 and 3 when the main steam isolation valves I (MSIVs) are closed "and deactivated." Note g is proposed to be added for the feedwater isolation and turbine trip function units to state that the LCO requirements are not applicable when all main feedwater isolation valves (FIVs) and bypass valves are closed and deactivated, or isolated by a closed manual valve. When these valves are closed and deactivated or isolated, they are performing their safety function. These safety functions are accomplished when the associated valves are closed and deactivated or isolated, whether that closure is as a result of automatic isolation circuitry or operator action. Operability requirements on actuation circuitry are not applicable if the valves are closed and deactivated or isolated. The proposed change will not affect any of the analysis assumptions for any of the accidents previously evaluated. The proposed change will not affect the probability of any event initiators nor will the proposed change affect the ability of any safety-related equipment to perform its intended function. There will be no degradation in the performance of nor an increase in the number of WOLF CREEK GENERATING STATION oRAFT SAFETY EVALUATION

challenges imposed on safety-related equipment assumed to function during an accident situation. Based on this, the proposed change is acceptable.

LS-31 The proposed change will relax the CTS action by providing an option for multiple  ;

inoperable channels on one bus to continue operating for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> which is not in the i CTS. For multiple inoperable channels on one bus, the new action would require restoration of all but one channel per Function (loss of voltage and/or degraded voltage) in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. For multiple inoperable channels on two buses, the new action would require restoration of all but one channel per function (loss of voltage and/or degraded voltage) on one bus in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, which does involve any relaxation in dhe allowed outage time currently required by entry into LCO 3.0.3. If the completion time for multiple channels on one bus inoperable were not met in Modes 1 to 4, the plant must be shutdown to 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 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. If the completion time were not met et other times in the applicability, the associated DG and offsite circuit would be immediately declared inoperable and corresponding actions would be taken. Since LCO 3.0.3 applies only in Modes 1 to 4, a relaxation of 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> would be involved for multiple inoperable channels on one bus as compared to immediate LCO 3.0.3 entry.

For the special case of tie breaker closure for repairs, affecting one bus only with an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> AOT in CTS Action 19, the relaxation would be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Based on this, the proposed change is acceptable.

LS-40 The proposed change deletes the requirement in Table 4.3-1, in footnote d for the function 15, to verify the setpoint during the quarterly trip actuating device operational test (TADOT) for reactor coolant pump (RCP) underfrequency and undervoltage. The licensee stated that the setpoint is adequately confirmed during the 18-month channel calibration. The functions are not an initiator for an accident previously analyzed but the functions are used to provide a reactor trip signal to mitigate certain accidents and the setpoints were selected such that the allowable value is not expected to be exceeded 1 during the 18 months between calibrations. Based on this, the proposed change is {

acceptable. )

I LS-43 The proposed change to CTS SR 4.3.3.2.1.a will limit the channel check to each required instrument channel"that is normally energized." The revised SR will exempt instrumentation that is not normally energized. The CTS require that channel calibrations are performed for instrumentation used in the post-accident monitoring and remote shutdown systems on an 18-month basis. Some of these instruments are then de-energized and remain in this state until re-energized for use in the management of plant events or for the performance of the channel checks. Channel checks are performed more frequently than channel calibrations for the purpose of detecting gross channel failures or excessive drift of one channel relative to other channels monitoring the same process variable. During the period that the channelis de-energized,it is not subject to the failure mechanisms or conditions that typically lead to instrument failure or excessive drift. Because de-energized channels are not subjected to the same failure mechanisms as energized channels, it was proposed to exempt instrumentation that is not normally energized from the performance of the periodic channel checks. Based on this, the proposed change is acceptable.

WOLF CREEK GENERATING STATION oRAFT SAFETY EVALUATION l

ITS Specification 3.4 LS-2 The proposed change adds an additional specific relaxation to allow the use of an operating RCS loop in lieu of an operating RHR loop in Mode 5 during planned heatup in preparation to enter Mode 4. The proposed change will relax CTS LCO 3.4.1.4.1, footnote "", by allowing the use of an operating RCS loop in lieu of an operating residual heat removal (RHR) loop in Mode 5 during planned heatup in preparation to enter Mode 4. The primary functions of the operating RHR loop in Mode 5 are to remove decay heat and to prevent boron stratification in the RCS. These functions can also be performed by an operating RCS loop which is a normal method of accomplishing these same functions when in Mode 4. In addition, at least one RHR loop must remain operable during the transition to Mode 4. The proposed change does not reduce the heat removal / boron mixing capability or system reliability when the RCS loop is performing these functions. Based on this, the proposed change is acceptable.

LS-12 The proposed change will delete the requirement in CTS SRs 4.4.5.2.1.a and 4.4.5.2.1.b to monitor the RCS leakage detection system once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Because the ITS LCO 3.4.15 requires that a channel check be performed on the containment radioactivity monitor channels on the same frequency as the CTS and the containment st.mp level and flow monitoring system and the condensate flow monitor are continuously monitored from the control room via available alarms and indications, such monitoring is unnecessary. Leak detection provides information that may indicate degradation of the RCS pressure boundary; however, the RCS leakage detection system is not credited in any safety analyses. Nevertheless, the continued operation of the leakage detection function is assured by the diverse means of leakage detection that have been provided within the system and by the requirement that a RCS water inventory balance be performed every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Because leakage information is available from diverse sources, which are checked by an RCS water inventory balance, the deletion of the surveillance does not negatively impact RCS leak detection. Based on this, the proposed change is acceptable.

LS-14 The proposed change will delete the requirement in CTS SR 4.4.5.2.1.e for monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Flange leakoff does  ;

not provide an indicator of pressure boundary integrity. Reactor head leakage, which is collected in the reactor coolant drain tan:t, is quantified as identified leakage which is determined by performance of a RCS water inventory balance and limited to a maximum value by the ITS. The initial RCS water inventory balance is required within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following RCS steady state operation and every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter. The flange leakage by itself is not an initial assumption in the accident analyses. Because reactor head flange leakage is accounted for by RCS inventory balance and can be detected by the various leakage monitoring systems, the proposed change is acceptable.

LS-15 The proposed change will relax the criteria, in CTS SR 4.4.6.2.2.b, for PlV testing following operation in Mode 5. The criterion that testing be performed after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i was extended to 7 days. Allowing additional time in Mode 5 before testing is required WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

i j

22 -

will have little or no impact on the pressure retaining capability of the isolation valves.

Based on this, the proposed change is acceptable.

LS-20 The proposed change will revise CTS SR 4.4.9.3.1.a to allow performance of the COT on the PORV actuation channels within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering Mode 4. Previously, such testing was required to be performed prior to entry into the mode. The proposed change will not affect the ability of the PORVs to perform their intended function as part of the cold overpressure mitigation system because the channel calibration is still required to be maintained current. In addition, entry into Mode 4 (from Mode 3) occurs at 350 F at which time the RHR system is normally in operation providing relief capability via the RHR relief valves. Based on this, the proposed change is acceptable.

LS-22 The proposed change will relax the CTS by requiring only two RCS loops to be operable when the control rod system is capable of rod withdrawal and one RCS loop to be operable when the control rod system is not capable of rod withdrawal. The LCO and Action b of Specification 3.4.1.2," Reactor Coolant System, Hot Standby," would be revised to require that two reactor coolant loops be operable. Loop operation requirements would also be revised to be contingent on rod control system status. The requirement to have a third operable reactor coolant loop would be deleted. The decay heat removal in Mode 3 is sufficiently low that a single RCS loop with one RCP running is adequate to remove core decay heat. A second RCS loop ensures redundant capability for decay heat removal. When the rod control system is capable of rod withdrawal, two loops must be in operation to ensure accident analysis assumptions are satisfied. When rod withdrawal is precluded, only one loop is required to be in operation to satisfy Mode 3 accident analyses. Based on this, the proposed change is acceptable.

LS-24 The proposed change will add 4 notes to the CTS LCO to reflect CTS SR 4.5.3.2, LCO 3.5.4 actions, LCO 3.5.4 applicability notes, and the accumulator action added in CN 9-10-M for CTS 3/4.4. Note 1 on centrifugal charging pump (CCP) swap operations is a relaxation of the CTS because it allows both CCPs to be capable of injecting into the RCS for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> whenever low temperature overpressure protection (LTOP) is required. Overall performance of the protection system will remain within the bounds of the previously performed accident analyses since no hardware changes are proposed.

The initial conditions and assumptions for the LTOP mass addition and heat injection transients will be unchanged. Actions will be taken to ensure that only one CCP is capable of injecting into the RCS during the LTOP applicability. The proposed change will not affect the probability of any event initiators nor will the proposed change affect the ability of any safety-related equipment to perform its intended function. There will be no degradation in the performance of nor an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation.

Based on this, the proposed change is acceptable.

LS-26 The proposed change will relax the limit for operational leakage from PlVs. This LCO has been modified to change the allowed leakage limit for reactor coolant system (RCS) pressure isolation valves. The licensee stated that the RCS pressure isolation valve LCO permits system operation in the presence of leakage through valves in amounts wolf CREEK GENERATIMG STATION oRAFT SAFETY Evaluation

i l

l which do not compromise safety. The RCS is isolated from other systems by valves.

During plant life these interfaces can produce varying amounts of reactor coolant I

leakage through either normal operational wear or mechanical deterioration. Increasing allowed leakage limits from 1 gpm up to 5 gpm for the pressure isolation valves will not challenge the pressure relief capacity of interfacing systems. This amount of leakage is I considered negligible when compared with the capacity of the pressure relief valves.

Pressure isolation valve leakage limits apply to leakage rates for individual valves. The basis for this LCO is that potentialintersystem loss of coolant accidents (LOCAs) are a l significant contributor to the risk of core melt. The study documented in NUREG-0677, l "The Probability of Intersystem LOCA: Impact Due to Leak Testing and Operational I Changes," dated May 1980, evaluated various pressure isolation valve configurations to i determine the probability of intersystem LOCAs. This study concluded that periodic leak l testing of the pressure isolation valves can substantially reduce intersystem LOCA probability. The previous criteria of 1 gpm for all valve sizes is not an indicator of imminent accelerated deterioration or potential valve failure. The licensee stated that another study documented in EG&G Report, EGG-NTAP-6175," Inservice Leak Testing of Primary Pressure Isolation Valves," dated February 1983, concluded allowable leak rates based on valve size was superior to a single allowable value. The single value imposes an unjustified penalty on the larger valves without providing information on potential valve degradation. In addition, enforcing the single value criteria resulted in higher personnel radiation exposures because larger valves must be repaired in place.

Based on this, the proposed change is acceptable.

LS-30 The proposed change will relax the CTS SR for performing an RCS water inventory balance by allowing deferral of the balance until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. An RCS water inventory balance cannot be meaningfully performed unless the plant is operating at steady state conditions. Therefore, CTS SR 4.4.6.2.1.d would be revised to allow deferring the RCS inventory balance in the event of a transient until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state conditions. Based on this, the proposed change is acceptable.

LS-34 The proposed change will limit the applicability of CTS SRs 4.4.4.1 and 4.4.4.2, footnote

1. , to perform the 92 day surveillance of the pressurizer PORV block valves and ths 18 month surveillance of the pressurizer PORVs (i.e., perform one complete cycle of each valve) to Modes 1 and 2. The licensee stated that the proposed change is consistent with the recommendations of Generic Letter (GL) 90-06," Resolution of Generic issue 70,' Power-Operated Relief Valve and Block Valve Reliability,' and Generic issue 94,

' Additional Low-Temperature Overpressure Protection for Light-Water Reactors,'

Pursuant to 10 CFR 50.54(f)," dated June 25,1990. The PORVs and block valves are required to be operable for manual operation to mitigate the effects associated with a SGTR event. The SGTR event is primarily a concern during Modes 1 and 2 where the consequences of an accident would be most severe. Based on this, the proposed change is acceptable.

LS-36 The proposed change willlimit CTS SR 4.4.4.2 to perform the 92 day surveillance of the pressurizer PORV block valves (i.e., perform one complete cycle of each block valve) so WOLF CREEK GENERATING STATION DRAFT SAFETY EvALUATloN

1 l

24-that it is not required to be performed if the block valve is closed to meet Action a. I Credit is taken only for the manual operation of the PORVs during the SGTR accident; however, the capability to manually cycle the PORVs will be unaffected by the proposed l change. This change will not affect the ability of the block valve to open, if closed to meet Action a,in the mitigation of an SGTR. Deferral of the block valve cycling surveillance will not diminish the design capability of the block valve to open against differential pressures that would be present after an SGTR since the block valves are capable of opening against 2485 psig, the safety valve lift pressure, whereas pressurizer pressure decreases after an SGTR. The lack of quarterly block valve cycling, which ,

could extend to a complete cycle since Action a allows continued operation with the block valves closed, does not decrease the likelihood of successful pressurizer relief since power remains available to the block valve motor operator (s) and the surveillance frequency for the PORVs can be as long as 18 months. Based on this, the proposed change is acceptable.

ITS Specification 3.5 I LS-4 (CN 3-02 LS-4) The proposed change revises the CTS prescriptive wording related to pump operability, in footnote

• to SR 4.5.3.2, to specifically address the ECCS pump capability to inject into the RCS. This change involves the configuration of the centrifugal charging and safety injection pumps. The LTOP limitations on ECCS pumps, )

and related surveillances, are relocated to ITS 3.4.12. The prescriptive wording related 1 to pump operability is changed to wording specifically addressing the pumps' capability  !

to inject into the RCS. The requirement for having the charging pumps / safety injection pumps ' inoperable' has been revised to preclude injection into the RCS. This change is consistent with the cold overpressure analysis requirements. The intent of specifying that the required number of centrifugal charging pumps / safety injection pumps be inoperable is to preclude the possibility of injecting flow into the RCS in excess of that analyzed for the LTOP system. This change results in the operability statements being revised and allows deletion of the notes which were in place for testing or accumulator filling. The change does not result in a less conservative operational position as flow to the RCS is still precluded. Based on this, the proposed change is acceptable.

LS-4 (CN 4-01 LS-4) The proposed change will (1) revise the CTS LCO 3.5.4 Action b and  ;

SR 4.5.4.2 (the footnote) to satisfy LTOP analysis assumptions on ECCS injection i sources by rendering pumps inoperable to preclude those pumps from injecting into the RCS, and (2) delete the note dealing with testing and accumulator filling. The LCO requirement to satisfy cold overpressure analysis assumptions on ECCS Injection sources by rendering pumps inoperable has been revised to preclude those pumps from injecting into the RCS. The change does not result in a less conservative operational position as flow to the RCS is still precluded. The intent of specifying that the required number of centrifugal charging pumps / safety injection pumps be inoperable is to meclude the possibility of injecting flow into the RCS in excess of that analyzed for the LTOP system. Based on this, the proposed change is acceptable.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

f LS-6 The proposed change revises the CTS requirement to SR 4.5.3.1 to demonstrate ECCS I train operability in Mode 4 to delete (1) the 31-day surveillance to verify the correct position of each valve in the ECCS flow path which is not already locked in place, and (2) the 18-month surveillance to verify automatic actuation of ECCS pumps and I automatic valves. Due to the stable conditions associated with operation in Mode 4 and the reduced probability of occurrence of a DBA, the emergency core cooling system (ECCS) operational requirements are reduced. In this Mode, there is sufficient time for manual actuation of the required ECCS to mitigate the consequences of a DBA. Based on this, the proposed change is acceptable.

ITS Specification 3.6 L-1 The change added a statement to CTS SRs 4.6.1.1.a and 4.6.1.7.1 (footnotes " and +)

and LCO 3.6.3 (action footnote +) on valve and blind flange surveillance requirements that allows verification of valves, flanges and isolation devices located in high radiation areas to be verified by use of administrative means. This adds an exception for valves, blind flanges, and deactivated automatic valves which are located inside containment and are locked, sealed, or otherwise secured in the closed position. The note allows verification of valves, flanges and isolation devices located in high radiation areas to be verified by use of administrative means. These valves shall be verified closed during each cold shutdown; however, under the CTS, if an area outside of containment became a high radiation area, entry into the area would still be required to verify the closed positions. The ITS would allow verification of all areas that are high radiation areas or become high radiation areas by administrative means once they had been verified to be in the proper position. This change is consistent with restricting access to these areas to maintain occupational exposure as low as is reasonably achievable (ALARA), as required by 10 CFR Part 20. The probability of misalignment of these devices, once they have been initially verified in the proper position, is small because these valves would be under administrative control. Based on this, the proposed change is acceptable.  !

LS-2 The proposed change added "and within 10 days from discovery of failure to meet the LCO" to the CTS LCOs 3.6.2.1 and 3.6.2.3 action completion times, which extends the overall AOT allowed by the CTS. The containment spray and containment cooling systems require restoring the inoperable system to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in the CTS. The CTS limits the inoperability of any combination of these two systems to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or it provides a maximum of 7 days for restoring one group of cooling fans to operable status when everything else is operable. The STS provides a maximum of 10 days for not meeting the LCO. The 10 day provision in the completion time is considered appropriate based upon engineering judgment considering the low probability of coincident entry into two conditions in this specification coupled with the low probability of an accident occurring during this time. Based on this, the proposed change is acceptable.

LS-9 The proposed change will reduce requirements by adding a note to CTS SRs 4.6.1.7.2 and 4.6.1.7.4, to state that leakage rate testing is not required for containment purge WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

valves with resilient seals when the penetration flow path is isolated by a leak tested blank flange. The purpose of the leak testing requirement is to ensure containment leakage Integrity during an accident, and thereby limit potential accident consequences.

Isolation of the flow path with a leak tested blind flange accomplishes this safety function and additional leak testing of the valves in the flow path is redundant and unnecessary.

The required action for a containment ventilation isolation valve (CVI) r ct 'within its leakage limit is revised to allow the penetration to be isolated using a closed and deactivated automatic valve, a closed manual valve or a blind flange and does not require the isolation valve to be restored to operable status. This is an option not explicitly available in the CTS. The completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> remains the same as in the CTS. If valves with resilient seals are used to isolate the flow path, the leak rate of these valves must be verified at least every 92 days. If a leak tested blind flange is used to isolate the penetration flow path, the valves with resilient seals whose flow is isolated by the blind flange are not required to be leak rate tested. Isolation of the flow path with a leak tested blind flange provides the required leak barrier and additional leak testing of the valves in the flow path is redundant and unnecessary. Based on this, the proposed change is acceptable, i LS-11 The proposed change to CTS SRs 4.6.1.7.2 and 4.6.1.7.4, for containment isolation valves with resilient seats, will (1) extend the leakage rate testing frequency for 18-inch pressure relief valves from once per 92 days to once per 184 day, (2) delete the requirement to test on a staggered test basis (STB) for the 48-inch and 12-inch purge valves, and (3) add the requirement to perform a leakage test within 92 days of opening the valves for both sets of valves. The leakage rate testing frequency for containment isolation valves with resilient seals is revised to 184 days based on the NRC resolution of Multi-Plant Action (MPA) B-20," Containment Leakage Due to Seal Deterioration."

The NRC evaluation for WCGS was issued [date). Testing on a STB is no longer necessary and is not included in the STS. A new requirement has also been added to j perform a leakage test within 92 days of opening the valves to acknowledge that cycling i the valve could introduce additional seal degradation beyond that occurring to a valve that has not been opened. Decreasing the interval (from 184 days to 92 days) is a prudent measure after a valve has been opened. On this basis, the proposed change is acceptable.

LS-19 The proposed change relaxes CTS requirements that are in a footnote to CTS SR 4.6.1.1.a and in a footnote to LCO 3.6.3. The revised footnote and a footnote added to the SR state that only containment isolation valves that are not locked, sealed, or otherwise secured are required to be verified closed at the once per cold shutdown frequency. The CTS SR requires g!! penetrations not capable of being closed by an operable containment automatic isolation valve (and required to be closed for accident conditions) be verified closed on a 31-day frequency, except for valves, blind flanges, and deactivated automatic valves that are located inside containment and are locked closed, sealed or otherwise secured in a closed position. These excepted penetrations are verified closed during each cold shutdown but not more often than once per 92 days. Penetrations (inside or outside containment) which are isolated by manual valves and blind flanges that are locked, sealed or otherwise secured are not required to be wolf CREEK GENERATING STATION oRAFT SAFETY EVALUATION

verified closed, since they are verified to be in the correct position prior to locking and securing the valve and are under administrative control. Based on this, the proposed change is acceptable.

LS-25 The proposed change will delete the CTS requirement in SR 4.6.1.7.1 to blank flange the containment shutdown purge supply and exhaust (CSDPSE) valves and revised the CTS SR for verification of the closed CSDPSE valves and flanges inside containment by adding the statemerit "if not completed in the previous 92 days." CTS 3.6.1.7 for the containment ventilation rystem requires the CSDPSE valves to be closed and blank flanged. In the event one containment isolation valve in one or more penetration flow paths is inoperable, the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed and deactivated power-operated containment isolation valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured. The requirement to blank flange the containment shutdown purge supply and exhaust isolation valves was removed because these valves are isolation barriers that meet this criterion. A provision "if not performed within the previous 92 days" is also added to the footnote to CTS Surveillance Requirement 4.6.1.7.1. The footnote allows the CSDPSE valves and flanges, located inside containment, to be verified closed (or flanges installed) prior to entering Mode 4 following each cold shutdown. The provision "if not performed within the previous 92 days"is reasonable because of the licensee's administrative controls that will ensure that isolation device misalignment is an unlikely possibility. Based on this, the proposed change is acceptable.

ITS Specification 3.7 LS-8 The proposed change will delete the CTS SR in item 1 of Table 4.7-1 to determine gross radioactivity. The consequences of secondary system releases are limited by radioiodines and their resultant thyroid exposures, not the whole-body exposures received for the noble gases and the primary to secondary leakage limits and dose equivalent I-131 limits ensure the dose analyses in the USAR remain valid. The CTS require that the gross radioactivity of the secondary system coolant be determined every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, but this surveillance is basically only an indicator of the potential offsite whole body dose. Since the radiciodines and the resulting thyroid dose are limiting, the 72-hour gross radioactivity surveillance requirement is deleted as being unnecessary.

Because the limits on primary to secondary leakage and dose equivalent 1-13T assure that the dose analyses in the USAR remain valid, the revised surveillance is more appropriate. The ITS will also require that the surveillance for verification of I-131 activity be performed every 31 days on an unconditional basis, which is more restrictive than the CTS. This change will only delete gross radioactivity sampling where results are bounded by the primary to secondary leakage and dose equivalent 1-131 limits.

Based on this, the proposed change is acceptable.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION l

LS-13 The proposed change will delete the CTS SR acceptance criteria that the laboratory analyses for the CR emergency filtration / pressurization system be completed "within 31 days after removal." This requirement is not contained in NRC Regulatory Guide (RG) 1.52," Design, Testing, and Maintenance Cnteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," the applicable American Nuclear Standards Institute (ANSI) standards, or the STS. The licensee stated that the failure to complete such an analysis within 31 days has insignificant safety consequences because the results would be available within approximately the same time period and it is very unlikely that the charcoal would be degraded to the extent that there would be a complete loss of a safety function. The licensee has a ventilation filter testing program in ITS 5.5.11 that includes the testing of this ventilation system. Based on this, the proposed change is acceptable.

LS-22 The proposed change will replace the specific CTS SR 4.7.1.3.2 to periodically verify the essential service water (ESW) system is in operation whenever the system is the supply source for the AFW pumps by a general statement to verify operability of the backup water supply. The periodicity of the verification does not change; however, the verification is relaxed to allow administrative means. The CTS require that when the condensate storage tank (CST) contained water volume is not within limits that the ESW system be demonstrated operable and by " verifying that the ESWS is in operation." The ITS will revise this action to " verify by administrative means" the operability of the ESW system as a backup supply to the auxiliary feedwater (AFW) pumps. The result of this change is that the ESW system would not have to be physically started should the contained water volume for the CST fall below the limit. Instead the ESWS would be required to be verified operable by administrative means. This change would include verification that the flow paths from the ESWS to the AFW pumps are operable, that the required volume of water is available, and that the pump meets its operability requirements. This is normal case for other specifications in the CTS. Based on this, the proposed change is acceptable.

LS 25 The proposed change will delete the CTS LCO 3.7.1.6, Actions a, b, and c, for atmospheric relief valves being inoperable due to sealleakage. Sealleakage is no longer a condition of operability because the atmospheric relief valves can perform their required function when they can be fully opened or closed on demand and can provide controlled relief of steam. Based on this the proposed change is acceptable.

ITS Specification 3.8 LS-4 For CN 1-47 LS-4, the proposed change will revise CTS SRs required for the CTS SR 4.8.1.2 en AC sources operability in Modes 5 and 6 to include only those SRs which are applicable for operability.

For CN 2-15 LS-4, the proposed change will add a note to the CTS SR allowing certain parts of the battery SR 4.8.2.2 to not have to be performed for the DC source operability in Modes 5 and 6. The licensee stated that the note does not delete the requirement WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

that the battery be capable of performing these functions, just that the capacity need not be demonstrated while that battery is relied on to meet the LCO.

i I

The revisions deleted certain CTS SRs that are not applicable because they depend on ESF actuation signals (which are not required to be operational during Modes 5 and 6) and automatic load sequencing (most of these loads are not required in Modes 5 and 6).

The SRs required for AC sources operability in Modes 5 and 6 would be revised to include only those SRs which are applicable. SRs that are not applicable are those that depend on ESF actuation signals (which are not required to be operational during Modes 5 and 6) and automatic load sequencing (most of these loads are not required in Modes 5 and 6). The 10-year simultaneous auto-start of all DGs is also not applicable 1 to Modes 5 and 6. The note listing exceptions to SR required for Modes 5 and 6 in CTS i 4.8.1.2 would be revised to include additional SRs. SRs that are applicable but not required to be performed are those that place a DG in parallel with offsite power which increases the probability of a station blackout. The licensee stated that the change assures the performance of SRs that are necessary and safe to perform for the plant conditions. The SRs required for AC sources and DC sources operability in Modes 5 and 6 would be revised to include only those which are applicable. In addition, notes would be added stating the SRs that are not required to be performed for operability in the modes governed by shutdown for the AC and DC sources LCOs. SRs were not listed as applicable for shutdown because (1) the SR is only required when DGs are required to be operable, (2) the SR is only required when the safety injection (SI) signal is operable, or (3) the SR is only required when the sequencers are required to be operable.

l For AC sources at shutdown, many of the CTS SRs involve tests that would require the  ;

one required DG to be paralleled to offsite power; this condition presents a significant i risk of a single fault resulting in a station blackout. Other tests, such as load rejection j tests, put the availability of the operable DG at risk during the test. To address this concern and to avoid potential conflicting TS, a note is added to not require that these i surveillances be performed in Modes 5 and 6.

For DC sources at shutdown, a note would be added stating which CTS SRs are not required to be performed for the DC source operability in Modes 5 and 6. Certain of the currently required SRs involve tests that would cause the battery to be rendered inoperable. If the only required operable battery were inoperable due to testing, the risk of an event occurring that would require battery operation, would present an additional risk. The exception provided by the note does not exempt for the battery from the requirement to be capable of performing the particular function, only that the capability need not be demonstrated while that source of power is being relied upon to support meeting the LCO.

The proposed SRs would continue to provide adequate assurance of the operability of the required AC and DC source functions. The changes would delete the requirement to meet SRs that verify functions which are not required in the applicable modes of the ITS.

wolf CREEK GENERATING STATION ORAFT SAFETY EVALUATION

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Based on this, the proposed changes are acceptable.

l i

! LS-12 The proposed change will add a footnote to CTS SR 4.6.1.1.2.g.7 stating that momentary transients outside the load and/or power factor range do not invalidate the SR tests. This is not allowed in the CTS. The licensee states that a footnote will be added stating that momentary transients outside the load range do not invalidate the test, since DG loading could change during this test due to changing bus conditions.

Some load fluctuation is expected and should not invalidate this test. The current  !

practice of monitoring and recording load every 15 minutes during the overload part of  !

the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> load test and once every hour for the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> is sufficient to ensure the DG load is within the load range. DG load found out of the load range and j immediately returned to within the band would not invalidate an DG load test. Based on '

this, the proposed change is acceptable. I LS-23 The proposed change will relax the CTS SR 4.8.2.1.e on battery capacity by also -

allowing a modified performance discharge test for verifying battery capacity and will l revise the CTS allowance that the performance discharge test may be performed in '

place of the battery service test of CTS SR 4.8. 2.1.d to (1) only allow the " modified" performance test replace the service test, and (2) delete the restriction that the discharge test could replace the service test only once per 60 months. CTS SR 4.8.2.1.e allows the performance of a modified performance discharge test in lieu of a service test only once per 60 months. This proposed change omits the "once per 60 months" limitation on use of a modified performance discharge test is lieu of a service test. lEEE-450-1996, Section 5.4 places no such limitation on use of a discharge test in  ;

place of a service test since the discharge rate is required to envelope the duty cycle of the service test. A modified performance discharge test is a test of the battery's ability to provide a high-rate, short-duration load. This will often confirm the battery's capability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a modified performance test. IEEE-450-1995, Section 5.4 states that,"A modified performance discharge test can be used in lieu of a service test at any time." Based on this, the proposed change is acceptable.

LS-26 The proposed change will restrict CTS LCO 3.8.3.2, for onsite power distribution in shutdown, to "the necessary portion of" electrical buses shall be energized "to support equipment required to be operable." Only the portions of these distribution subsystems necessary to supply AC and DC power to equipment required to be operable in shutdown mutt be operable. The change revises the requirement for operable onsite shutdown power. The CTS requires that one train (subsystem) of the various power supplies and busses be operable. The change requires that only the necessary portions of these subsystems be operable. The necessary portions are those portions required to support the equipment in that train which is required to be operable in the existing shutdown conditions. There is no reason to have portions of the power systems operable that are not supporting components which are being credited in the safety analyses for shutdown events. Based on this, the proposed change is acceptable.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

ITS Specification 3.9 LS-2 The proposed change will delete CTS SR 4.9.1.1 to verify reactivity conditions in the LCO for Mode 6 prior to (1) removing or unbolting the reactor vessel head, and (2) withdrawal of any control rod greater than 3 feet from its fully inserted position. The first of these requirements is redundant to the requirement imposed by the applicability note in ITS LCO 3.9.1 to meet the LCO prior to entering Mode 6 from Mode 5. Compliance with the LCO is assured by verifying boron concentration in accordance with ITS SR 3.9.1.1. In this case, unbolting the vessel head in preparation for removal is part of the definition of Mode 6. Therefore, this requirement is redundant to the requirement to verify boron concentration prior to entry into Mode 6. The second requirement that involves withdrawal of control rods is redundant because the analysis used to determine the boron concentration limit specified in the COLR considers the most adverse conditions of fuel assembly and control rod position. The boron concertration is sufficient to maintain k, s 0.95 with the most reactive rod control cluster assembly completely removed from its fuel assembly. Based on this, the proposed change is acceptable.

LS-3 The proposed change, for CTS SRs 4.9.2.b and 4.9.2.c and a new SR, will (1) delete the analog COT requirements, and (2) add a channel calibration, for source range neutron flux monitors in Mode 6. In Mode 6, the source range monitors are required for indication only and there are no precise setpoints associated with these instruments. In this capacity, the source range instrumentation is typically used to read a relative change in count rate and is monitored for significant changes in count rate which are important to evaluate the change in core status. In the STS, indicating instruments only require channel checks and channel calibrations. The more frequent ACOTs are applied only to those channels with operational interlocks or other setpoint actuations.

Therefore, the Mode 6 channel checks and channel calibration requirements for the source range monitors are adequate to assure their operability, considering the more frequent ACOTs performed on this instrumentation in other Modes, the effectiveness of these surveillance requirements in maintaining other indicating instruments operable, and the accuracy required of these instruments in Mode 6. Based on this, the proposed change is acceptable.

LS-4 The proposed change will delete CTS SR 4.9.4 to perform verification within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of core alteration or movement of irradiated fuel. The purpose of the CTS SR is to assure the operability of the containment penetrations that must be closed or capable of closing to prevent the release of radioactivity in the event of a fuel handling accident (FHA). The SR is intended to assure that mitigation features are available and has no impact on the probability of an accident occurring. The applicability statement for this LCO is "During CORE ALTERATIONS or movement of irradiated fuel within the containment." Therefore, the requirement to verify the LCO is met within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of starting the evolutions for which the LCO is applicable is redundant; because the LCO must be met at the time that the evolutions occur and only the timing (7 day frequency versus 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) is different. The 7-day frequency is adequate surveillance. Based on this, the proposed change is acceptable.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALuATloN

1 I

1 LS-6 The proposed change will relax CTS requirements in LCO 3.9.8.1 by allowing the removal of the RHR loop from operation for additional purposes other than the performance of core alterations in the vicinity of the hot legs. The licensee stated that the change will allow the removal of the RHR loop frorn < Jeration for additional purposes other than the performance of core alterations in the vicinity of the hot legs. ,

Therefore, this will allow increased flexibility for core mapping and isclation valve testing. l No operations are permitted that would result in the RCS boron concentration being outside the limits in the COLR because the operations are controlled by LCO 3.9.1. l Based on this, the proposed change is acceptable.

l LS-7 The proposed change will delete CTS LCO 3.9.9 Action a to close each purge valve when the containment ventilation system is inoperable. The containment ventilation TS requirements would be integrated into ITS 3.9.4 which has the effect of changing the actions required when the ventilation system is inoperable from closing the purge valves j to suspending core alterations and irradiated fuel movement. The applicability of the {

LCO and required actions for both the current TS and ITS 3.9.4 are identical, i.e., During j core alterations or movement of irradiated fuel assemblies within containment.

Therefore, neither of these LCOs would be in effect if core alterations or movement of irradiated fuel were suspended. The function of the purge valves is to close following a  ;

FHA to prevent the escape of radioactivity from containment. Therefore, the proposed change would have no effect on the probability of a FHA. In addition, the change from requiring the valves to be closed to prevent radioactivity release to suspending activities which could lead to a FHA (and radioactivity release) would have the same effect with regard to consequences of the accident. Based on this, the proposed change is acceptable.

LS-9 The proposed change will delete the "within 31 days after removal" requirement in CTS SRs 4.9.13.b.2 and 4.9.13.c for completion of laboratory analyses for the emergency exhaust system (EES) carbon sample. To assure charcoal adsorber operability, the j CTS SR requires that a laboratory analysis be performed and the results obtained within 31 days of removing the charcoal sample. The sample must be sent to an offsite laboratory for this analysis. It is proposed that the time requirement of "within 31 days after removal" for completion of laboratory analyses be deleted. This requirement is intended to avoid extended plant operation with degraded charcoal filters. This requirement is not contained in the ITS nor is it contained in the Regulatory Guide 1.52 or the applicable ANSI standards. There is no safety significant basis for maintaining this time limit as a CTS requirement. Laboratory analyses are performed under contract with a laboratory on a prompt basis, and it is not necessary to prescribe a time limit within CTS for completing the analysis. Failure to complete an analysis within 31 days has insignificant safety consequences because the results would be available within approximately the same time period and it is very unlikely that the charcoal would be degraded to the extent that there would be a complete loss of a safety function. Based on this, the proposed change is acceptable.

LS-11 The proposed change will delete the restrictions in CTS LCO 3.9.12.b, while refueling operations arc in progress, on placing spent fuel into region 2 of the spent fuel pool and WOLF CRCK GENERATING STATION DRAFT SAFETY EVALUATION

changing storage location designations from region 1 to region 2. ITS SR 3.7.17.1 requires that the spent fuel be verified as acceptable for placing in Region 2 prior to placing the fuelin Region 2. Also, any changes of storage locations from Region 1 to ,

Region 2 are controlled by plant procedure. Based on this, the proposed change is l acceptable.

LS-20 The proposed change will add a footnote to CTS LCO 3.9.4.c stating that penetration flow paths that provide direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls. The note would allow these penetrations to be unisolated during core alterations and movement of irradiated fuel assemblies within containment provided that specified administrative controls were ,

eniployed. The proposed note requires administrative controls that consist of written j procedures that require designated personnel having knowledge of the open status of I the valves in question and specified persons designated and readily available to isolate i the open penetration in the event of a fuel handling accident. These administrative controls provide equivalent protection to that afforded by the administrative controls used to establish containment closure for a containment personnel airlock. The NRC has allowed changes to the requirements for airlocks that allow both doors of an airlock to be open during core alterations and during movement of irradiated fuel inside containment provided that administrative controls are in place to quickly close one door and establish containment closure. The change would ensure the isolation vaWes, or functional equivalent, will perform their required containment closure function and will serve to limit the consequences of a FHA such that the results of the safety analysis in the USAR is unchanged. In considering the consequences of a design basis FHA inside containment, the assumptions in the analysis take no credit for the containment as a barrier to prevent the postulated release of radioactivity. For events that would occur during core alterations or movement of irradiated fuel assemblies, containment closure is considered a defense-in-depth boundary to prevent uncontrolled release of radioactivity. Based on this, the proposed change is acceptable.

LS-21 Tha proposed change will delete the CTS LCO 3.9.2 requirement related to indication provided by the source range detectors for refueling operations instrumentation. The change would eliminate requirements associated with indication channels that are not required to mitigate boron dilution events. The requirements for visual indication for plants that do not rely on a boron dilution analysis would be discussed in the ITS Bases I and the requirements for audible indication would be eliminated. In Mode 6, the source range monitors are required for indication only and there are no precise setpoints associated with these instruments. In this capacity, the source range instrumentation is ,

typically used to read a relative change in count rate. The source range instrumentation l is monitored for significant changes in count rate which are important to evaluate the change in core status. The accepted convention for defining criticality does not require precise or specific setpoints or indication, but only requires verification of a slowly increasing count rate. The ITS requirements consist of maintaining two source range neutron flux monitors operable to ensure that redundant monitoring capability is available to detect changes in core reactivity. There is no requirement for an audible signal or alarm to initiate operator responso because in Mode 6 reactivity changes would WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

l be slow and a boron dilution accident is not postulated. The occurrence of a boron dilution event is precluded by maintaining the isolation valves from unborated water sources secured in the closed position in accordance with ITS 3.9.2. During refueling, the source range monitors are designed to provide visual and audible indication of l neutron count rate to plant operators. The proposed deletion of audible indication for these channels would not affect the availability of visual indication. There are no alarms, .

interlocks, or trip setpoints associated with these channels that are required to be l operable during Mode 6. In addition,in Mode 6 the source range instruments provide no automatic actuation function used for mitigation of accidents, and they would have no effect on the outcome of an accident. Based on this, the proposed change is l acceptable.

LS-22 The proposed change will delete (1) CTS SR 4.9.10.1 to verify water level within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> i prior to the start of movement of fuel assemblies and (2) deleted descriptive detail not '

needed in the SR. CTS LCO 3.9.9.1 on the required water level must be met at the time that movement of fuel assemblies is performed because tne LCO is applicable for this movement. The SR for level verification within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to irradiated fuel movement l is not needed because the SR for verifying refueling pool level every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is retained in the CTS and is sufficient for ensuring that the water level over the core is at an acceptable level. The descriptive detail deleted from the SR does not meet the criteria in the Commission's Final Policy Statement for inclusion in TS. Based on this, the proposed change is acceptable.

LS-26 The proposed change will establish a new CTS LCO 3.9.13 Action c and completion times for degradation of the fuel building pressure envelope. The change provides specific required actions for failed surveillances designed to detect ventilation system envelope degradation. These surveillances require a positive or negative pressure limit be satisfied in the area with the associated required ventilation train operating. While other surveillances in the same specification test the operability of the ventilation train, these surveillances ensure the envelope leak tightness is adequate to meet the design assumptions. However, there are no corresponding conditions, required actions, or completion times associated with these surveillances. The licensee stated that the proposed change would allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore the capability to maintain the proper pressure by allowing for routine repairs before requiring the unit to perform an orderly shutdown. Based on this, the proposed change is acceptable.

ITS Section 5.0 LS-2 The proposed change will extend the time to complete the analysis of the fuel oil from 30 days to 31 days. The licensee stated that the surveillance interval for verifying that other properties are with limits for ASTM 2D fuel oil will be changed from "within 30 days" to "within 31 days" after obtaining a sample. The fuel properties that can have an immediate detrimentalimpact on diesel combustion, (i.e., API gravity, kinematic viscosity, flash point and appearance) are verified prior to addition to the storage tank.

The "other properties" may be analyzed after addition to the tank. The licensee stated that the 31-day verification interval for these properties is acceptable because the fuel WOLF CREEK GENERATING STATION oRAFT SAFETY EVALUATION

properties of interest, even if they are not within their stated limits, would not have an immediate effect on diesel generator operation. The CTS 30-day verification interval was probably chosen because it was a convenient time interval for sending the sample and receiving the results from the laboratory selected for testing and NUREG-1431 has selected a 31-day testing interval. The 1-day increase in the interval would not have a significant effect on the acceptability of the diesel fuel oil. Based on this, the proposed change is acceptable.

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

D. Relocated CTS Details (Not Entire Specifications)

When requirements in the TS have been shown to give little or no safety benefit, their removal from the TS may be appropriate. This includes details that are not necessary for inclusion in the TS because, for example, the details do not support the safety analyses or do not support the operability of systems. This section discusses the relocation of details within the CTS to licensee-controlled documents. The relocation of entire specifications from the CTS to licensee-controlled documents is discussed in Section 3.E below. In most cases, relaxations previously granted to licensees on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and oper8 ting experience, or (3) resolution of the Owners Groups comments on the STS (i.e., the TSTF process). 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 WCGS design was also reviewed to determine if the specific design basis and licensing basis of the WCGS were consistent with the technical basis for the model requirements in the STS, and thus provide a basis for the proposed ITS. A significant number of changes to the CTS involved the removal of specific requirements and detailed information from individual specifications evaluated to be Types 1 through 4 that follow:

Type 1 Details of System Design Type 2 Descriptions of System Operation Type 3 Procedural Details for Meeting TS Requirements Type 4 Requirements Redundant to Regulations Type 5 Requirements Not Supporting Safety Analyses The following discussions address why each of the four types of information or specific requirements are not required to be included in ITS.

WOLF CREEK GENERATING STATION oRAFT SAFETY EVALUATION

Details of System Desion (Type 1)

The design of the facility is required to be described in the USAR by 10 CFR 50.34. In addition, the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50 require that station design be documented in controlled procedures and drawings, and maintained in accordance with an NPC-approved QA plan (referenced in the USAR). In 10 CFR 50.59 controls are specified for changing the facility as described in the USAR, and in 10 CFR 50.54(a) criteria are specified for changing the QA plan. The ITS Bases also contain descriptions of system design and ITS 5.5.10 specifies 10 CFR 50.59 controls for changing the Bases. Removing descriptive details of system design from the CTS is acceptable because this information will be adequately controlled in the USAR, controlled design documents and drawings, or the TS Bases, as appropriate.

Cycle-specific design limits are moved from the CTS to the core operating limits report (COLR) in accordance with NRC GL 88-16. ITS 5.6.5 has the programmatic requirements for the COLR.

. Descriotions of System Ooeration (Type 2)

The plans for the normal and emergency operation of the facility are required to be described in the USAR by 10 CFR 50.34. Controls specified in 10 CFR 50.59 apply to changes in procedures as described in the USAR. Controls specified in 10 CFR 50.54(a) apply to changes to the OA Program. The ITS Bases also contain descriptions of system operation and ITS 5.5.10 specifies that 10 CFR 50.59 will be used for making changes to the Bases. It is acceptable to remove details of system operation from the TS because this type of information will be adequately controlled in the USAR, OA program, station operating procedures described in the USAR, and the ITS Bases, as appropriate.

. Procedural Details for Meetina TS Reauirements (Type 0)

Details for performing action and surveillance requirements are more appropriately specified in the USAR, station procedures required by ITS 5.4.1, the ITS Bases, the technical requirements manual (TRM), or in programmatic documents, such as the offsite dose calculation manual (ODCM), which are required by ITS 5.5. Typically, details for performing action and surveillance requirements are already contained in the station procedures required by ITS 5.4.1. ITS 5.4.1.a requires written procedures to be established, implemented, and maintained for station operating procedures including ,

procedures recommended in NRC RG 1.33, Revision 2, Appendix A, February 1978. l These procedures ensure proper implementation of action and surveillance requirements. For example, control of the station 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 GL 91-04," Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," allowing this procedural controlis consistent with the vast majority of other SRs that do not dictate station conditions for surveillances. Prescriptive procedural information in an action requirement is unlikely to contain all procedural considerations WOLF CREEK GENERATING STATION oRAFT SAFETY EVALUATION l

necessary for the station operators to complete the actions required, and referral to station procedures is, therefore, required in any event.

Removing procedural details for meeting TS requirements from the TS is acceptable because locating such details in the USAR, the ITS Bases, the TRM, or in programmatic documents required by ITS Section 5.5, as appropriate, will maintain an effective level of regulatory control while providing for a more appropriate change control process, such l as 10 CFR 50.59 and ITS 5.5.10, Bases Control Program. Similarly, deleting reporting requirements in the CTS is appropriate because ITS Section 5.6," Reporting Requirements, 10 CFR 50.36 and 10 CFR 50.73 adequately cover the reports deemed to be necessary. i

\

Reauirements Redundant to Reaulations (Type 4)

Certain CTS administrative requirements are redundant to regulations and thus are relocated to the USAR or other appropriate licensee-controlled documents. The Final Policy Statement allows licensees to relocate to licensee-controlled documents CTS requirements that do not 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) governs changes to the OA plan, and ITS .

5.5.1 governs changes to the USAR. Therefore, relocation of the administrative details  !

identified above, is acceptable.

. Reauirements Not Sucoortina Safety Analyses (Type 5)

The TS rule,10 CFR 50.36, provides criteria for determining what requirements should be specified in the TS LCOs. These criteria are based on meeting the safety analyses i for the plant. In some cases, while a TS LCO may support the safety analyses, certain other requirements within the specification, such as a SR, may not. Since the l Commission's Final Policy Statement a%s ficensees to relocate CTS LCOs that do not I meet any of the 10 CFR 50.36 criteria to licensee-controlled documents, it is also acceptable to allow licensees to also relocate certain requirements within LCOs ,to licensee-controlled documents, when these requirements do not support the safety analyses for the plant.

Table LG lists the requirements and detailed information in the CTS that are being relocated to licensee-controlled documents and not retained in the ITS. Table LG is organized in CTS order by each LG-type DOC to the CTS. It includes the following: (1) the sequential number within the ITS section or specification designation, as appropriate, followed by the DOC identifier (e.g.,

1-10 followed by LG); (2) the CTS reference where the detail was located; (3) a summary description of the relocated details; (4) the document to contain the relocated details or requirements (i.e., the new location); (5) the regulation or ITS section for controlling future changes to the relocated detail or requirement (i.e., the control process); (6) a characterization WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

l of the change; and (7) a reference to the specific change type, as discussed above, for not including the information or specific requirements in the ITS (i.e., Type 1,2, 3, or 4).

The NRC staff has concluded that these types of detailed information and specific requirements do not need to be included in the ITS to ensure the effectiveness of ITS to adequately protect the health and safety of the public. Accordingly, these requirements may be moved to one of the following licensee-controlled documents for which changes are adequately govemed by a regulatory or TS requirement:

TS Bases controlled in accordance with ITS 5.5.14, " Technical Specifications Bases Control Program."

Documents that have controls established by the Administrative Controls section of the ITS (e.g., ODCM in ITS 5.5.1, inservice inspection program in ITS 5.5.8, I explosive gas and storage tank radioactivity monitoring program in ITS 5.5.12, diesel fuel oil testing program in ITS 5.5.13, and core operating limits report in ITS 5.6.5).

. USAR (which includes the TRM by reference) controlled by 10 CFR 50.59.

OA plan, as approved by the NRC and referenced in the USAR, controlled by 10 CFR Pert 50, Appendix B, and 10 CFR 50.54(a).

The above is not a complete list of the acceptable licensee-controlled documents that could be i used to incorporate relocated CTS requirements. Table Table LG of details relocated from CTS, Table R of relocated CTS, and Table LS of less restrictive change to CTS list the licensee-controlled documents that are acceptable for relocated WCGS CTS requirements.

To the extent that requirements and information have been relocated to licensee-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, the NRC staff has concluded that they do not fall within any of the four criteria contained in 10 CFR 50.36 and discussed in the Final Policy Statement (see Section 2.0 of this SE). Accordingly, existing detailed information and specific requirements, such as generally described above, may be removed from the CTS and not included in the ITS.

E. Relocated Entire CTS Specifications The Commission's Final Policy Statement states that LCOs and associated requirements that do not satisfy or fall within any of the four specified criteria (now contained in 10 CFR 50.36) may be relocated from the CTS (an NRC-controlled document) to appropriate licensee-controlled documents. This section of the SE discusses the relocation of entire specifications in the CTS to licensee-controlled documents. These specifications include the LCOs, action j statements (i.e., LCO actions), and associated SRs. In its application and its supplements, the i licensee proposed relocating such specifications from the CTS to the TRM, which is  !

incorporated in the USAR by reference. The staff has reviewed the licensee's submittals, and i finds that relocation of these requirements to the TRM is acceptable, in that changes to the I TRM will be adequately controlled by 10 CFR 50.59. These provisions will continue to be WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION l

j

i l

implemented by appropriate station procedures (i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and work control procedures).

The licensee, in electing to implement the specifications of the STS, also proposed, in accordance with the criteria in the Final Policy Statement and 10 CFR 50.36, to entirely remove certain specifications from the CTS and place them in licensee-controlled documents. Table R lists all specifications that are being relocated from the CTS to licensee-controlled documents.

Table R is organized by each R-type DOC to the CTS, in a manner consistent with the organization of requirements in the CTS. Table R has the following: (1) the CN, (2) a references to the relocated CTS requirements, (2) summary descriptions of the reloc:ded CTS requirements, (3) name of the document that will contain the relocated requirements (i.e., the new location); and (4) the method for controlling future changes to the relocated requirements (i.e., the control process).

The NRC staff's evaluation of each relocated specification listed in Table R is provided below, in order of the ITS section and then the CN number.

l

1. CN 1-38-R CTS Table 3.3-1, Functions 6.a and Action 5, Source Range Neutron l

Flux, Reactor Trip and Indication, Shutdown There are requirements for the source range neutron flux, reactor trip and indication in shutdown specified in CTS Table 3.3-1. In Modes 3,4, and 5, when all control rods are fully inserted and the rod control system is incapable of rod withdrawal, the source range neutron flux function does not provide input to any reactor trip function nor is it credited for mitigation of any DBA. At one time, the source range neutron flux signal was credited in the analysis of the inadvertent boron dilution event; however, as described in a previous license amendments 20  ;

and 6 for WCGS, this analysis was revised to take credit for other alarms and not the source i range neutron flux function. Based on this and because the source range neutron flux function in these conditions does not satisfy any of the four criteria of 10 CFR 50.36(c)(2)(ii), it is being relocated from the TS to the USAR. The operation of this function will continue to be l maintained thrt gh plant procedures as it does provide important alternate indication to the reactor operators. The USAR is an acceptable licensee-controlled document for this information because changes to the USAR are controlled by 10 CFR 50.59. This relocation is i

acceptable.

2. CN 7-04-R CTS 3.6.1.7, Action b, Containment Ventilation Valves l The time limit restrictions on opening the 18-inch containment mini-purge supply and exhaust valves and the requirements to periodically accumulate the time that the valves have been open l would be relocated to the USAR. These requirements do not represent initial condition assumptions of any accident analysis and do not meet the criteria in the Commission's Final Policy Statement to be included in TS. They are being relocated to the USAR which is an acceptable licensee-controlled document for such requirements. Based on this, the relocation is acceptable.

l WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION l

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1 l

3. CN 3-01-R CTS 3/4.9.3, Reactor Decay Time ,

1 The requirements in CTS 3/4.9.3 on the decay time that the reactor core must be subcritical before there is movement of irradiated fuel in the reactor core are being relocated to the TRM.

This LCO requires 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to elapse to allow the radioactive decay of the short-lived fission '

l products. The screening criteria for including the requirements in the ITS have been satisfied for Criterion 2 since decay time is consistent with the assumptions used in an accident analysis; however, the activities necessary to be performed at WCGS before commencing movement of irradiated fuel ensure that 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of suberiticality will elapse before there is movement of irradiated fuel in the core. Therefore, the decay time LCO and SRs may be relocated to a l licensee-controlled document outside TS, because the Final Policy Statement criteria have not been satisfied. The TRM is an acceptable licensee-controlled document and the relocation is j acceptable.

l l The three relocated specifications from the CTS discussed above are not required to be in the ITS because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. In addition, the NRC staff finds that sufficient regulatory controls exist under the regulations cited above to maintain j the effect of the provisions in these specifications. The NRC staff has conciuded that '

appropriate controls have been established for all of the current specificat:ons, information, and requirements that are being moved to the TRM.

This relocation is the subject of a license condition discussed in Section 6.0 of this SE. Until incorporated in these licensee-controlled documents, changes to these specifications, .

information, and requirements will be controlled in accordance with the current applicable )

procedures that control these documents. Following implementation of the ITS and incorporation of these relocated requirements, the NRC will audit the removed provisions to ensure that an appropriate level of control has been achieved. The NRC staff has concluded that, in accordance with the Commission's Final Policy Statement, sufficient regulatory controls exist under the regulations, particularly 10 CFR 50.59. Accordingly, these specifications, information, and requirements, as described in detail in this SE, may be relocated from the CTS and placed in the identified licensee-controlled documents as specified in the licensee's letters.

F. Control of Specifications, Requirements, and Information Relocated from the CTS in the ITS conversion, the licensee will be relocating specifications, requirements, and detailed information from the CTS to licensee-controlled documents outside the CTS. This is discussed I

in Sections 3.D and 3.E above. The facility and procedures described in the TRM, incorporated into the USAR 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 the 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 WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

1 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 the licensee in accordance with the record retention requirements specified in the licensee's OA plan for WCGS and such applicable regulations as 10 CFR 50.59.

The license condition for the relocation of requirements from the CTS, discussed in Section 6.0 I

of this SE, will address the implementation of the ITS conversion, and when the relocation of the CTS requirements into licensee-controlled documents will be completed. The relocations to the TRM may be included in the next required update of this document.

l G. Evaluation of Other TS Changes included in the Application for Conversion to ITS This section addresses the beyond-scope issues (BSis) in which the licensee proposed changes to both the CTS and STS. The staff has published a notice of consideration for these BSis in the Federal Register (63 FR 53471); however, some of the notices issued for the proposed amendments were provided for changes to the CTS that are now not considered beyond the scope of the conversion in that they are now not considered changes to both the CTS and STS.

The changes discussed below are listed in the order of the applicable ITS specification or section, as appropriate (from ITS 3.4.5 to ITS 3.8.6).

1. ITS 3.4.5 CTS 3.4.1.2. Establishes Temperature Restrictions on Startina an Idle Reactor i Coolant Pumo (RCP) When Below the LTOP Armina Temperature of 368 decrees F. (CN 1-05-M for CTS 3/4.4)

A note is proposed to be added to CTS LCO 3.4.1.2 and ITS 3.4.5 to establish temperature I restrictions that must be met before starting an idle reactor coolant pump when below the LTOP system arming temperature of 368F. The proposed change is not in the STS.

CTS LCO 3.4.1.2 is applicable in Mode 3. The LTOP arming temperature of 368*F is in Mode 3. The note being added to CTS LCO 3.4.1.2 would prevent starting a reactor coolant pump with the RCS cold leg temperature less than or equal to 368 F (LTOP arming temperature), unless the secondary side water temperature in the steam generator is less than or equal to 50*F above the cold leg RCS temperature. This is an assumption in the WCGS LTOP analysis that prevents a low temperature overpressure event due to a thermal transient when a reactor coolant pump is started. Adding this note constitutes an additional restriction not found in the CTS or ITS. The staff finds this note acceptable in that it provides additional control to help ensure the assumption in the LTOP analysis remains valid.

Similar notes are included in CTS LCO's 3.4.1.3 and 3.4.1.4.1 (ITS 3.4.6 and 3.4.7). These notes state that a reactor coolant pump shall not be started unless the secondary side water l temperature in the steam generator is less than or equal to 50*F above the cold leg RCS temperature. The notes currently do not identify that the restriction on starting a reactor coolant pump is applicable only when RCS cold leg temperature is less than or equal to 368oF. This WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION l

restriction is added to these notes. Because the proposed note is consistent with the WCGS LTOP analysis, the proposed change is acceptable.

2. ITS SRs 3.4.5.2 and 3.4.6.2. CTS SRs 4.4.1.2.2 and 4.4.1.3.2. Reolace Steam and LCO 3.4.7 Generator Secondary Side Water Level (SGSDWL) Of

" wide rance . . areater than or eaual to 10%" with " narrow rance . . areater than or eaual to 6%." LCO 3.4.1.4.1.b.

Reolace SGSSWL Of "10% of the wide ranae" with "66%

of the narrow ranae." (CN 1-15-M for CTS 3/4.4)

CTS SR's 4.4.1.2.2 and 4.4.1.3.2 require steam generator (SG) levels to be periodically verified to be greater than or equal to 10% wide range water level The proposal is to change this level value to 6% narrow range water level. The proposed change is not in the STS.

The CTS value of 10% wide range does not ensure all SG tubes are covered. The licensee stated that the proposed 6% level value includes uncertainties, and corresponds to a SG level approximately 100 inches above the top of the highest SG tube. This levelis sufficient to ensure the tubes remain covered and that the SGs provide an adequate heat sink for removal for decay heat. Additionally the proposed value of 6% narrow range level is used in the Wolf  ;

Creek emergency operating procedures. Because the proposed SG level value of 6% narrow range will ensure SG tubes are maintained covered to provide an adequate heat sink for decay heat removal, the proposed change is acceptable.

A similar change is proposed for CTS LCO 3.4.1.4.b. This LCO currently requires that in Mode 5, with the reactor coolant loops filled and one RHR loop operable and in service, the secondary side water level of at least two SGs be maintained " greater than 10% of the wide range." The proposed change is not in the STS.

This CTS level value of 10% wide range does not ensure all SG tubes are maintained covered with water. The proposal is to increase this value to " greater than 66% of the wide range."

The licensee stated that, for Mode 5 conditions, the 66% wide range level corresponds to the top of the highest SG tube, with margins added for instrument loop errors and readability. The wide range instrumentation is calibrated for cold conditions. Because this value will ensure SG tubes remain covered in Mode 5 when SGs are required to be operable, the proposed change is acceptable.

3. ITS LCO 3.4.11. Condition D. and CTS LCO 3.4.4. Actions a. b. c. and d. as.d CTS LCO 3.4.16. Conditions B and C LCO 3.4.8. Actions a and b. Reduce LCO Action Reauirements To Prevent Plant Enterina into LTOP Conditions with inocerable Power-Ocerated Relief Valves. (CN 4-05-LS 31 for CTS 3/4.4)

The proposed changes will (1) replace the CTS LCO 3.4.4 actions for inoperable power-operated relief valves (PORVs), to eventually place the plant in hot shutdown, by the requirement to reduce Tmto <500 F and (2), for consistency, revise the actions of CTS LCO 3.4.8 in a sirr!!ar manner. The proposed changes would also revise and extend the time to WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

1 i

reach the required T,y by 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The changes would relax the action requirements for both LCOs. The proposed changes are not in the STS.

l With inoperable PORVs, the PORVs would not be able to perform their overpressure protection function of CTS LCOs 3.4.4 and 3.4.9.3. No credit is taken for the automatic actuation of the l PORVs in Modes 1,2, or 3. Credit is taken for the manual operation of the PORVs during a l SGTR. The capability to manually cycle the PORVs will be unaffected by this change.

The requirement of the LCO 3.4.4 actions to have the plant shut down to Mode 4 would result in l

the plant being within the applicability of LCO 3.4.9.3 which requires the PORVs to be operable.

It is not prudent to intentionally enter the LCO 3.4.9.3 applicability with inoperable PORVs and  ;

thus reduced overpressure protection. The licensee proposed a change to the LCO 3.4.4 actions to require shutdown to Mode 3 with RCS Tavg <500*F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, where the function of the PORVs to mitigate an SGTR is no longer needed. The proposed change would also include a continuing requirement to undertake actions to restore inoperable valve (s). The licensee stated that the risks of entering the applicability of LCO 3.4.9.3 with inoperable PORVs is greater than allowing the shutdown of the plant to terminate at Mode 3 with Tavg <500*F.

The LTOP arming temperature for WCGS is 368'F. Requiring the plant to shut down to Mode 4 (i.e., RCS temperature less than 350*F) would place the reactor vessel in a condition where a low temperature overpressure event is possible and the LTOP system is unavailable due to inoperable PORVs. The proposed change would allow the RCS temperature to be above the LTOP arming temperature.

The actions for CTS LCO 3.4.8 on the maximum allowable value of the specific radioactivity in

, the RCS are also proposed to be revised because the maximum allowable value would I determine the release of radioactivity in an SGTR avent. The release of radioactivity in the SGTR event with RCS Tavg <500*F is unlikely since the saturation pressure of the reactor coolant would be less than the lift pressure of the main steam safety and SG atmospheric relief valves. Since the Bases for CTS LCO 3.4.8 are also associated with SGTR, shutdown requirements would be revised to c equire Mode 3 be reached within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and T,y < 500*F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed change to the action termination point of CTS 3.4.4 and the increased completion time of CTS 3.4.8 will not affect the probability of any event initiators nor will the proposed change affect the ability of any safety-related equipment to perform its l intended function. The proposed change in the action statements will not affect any of the l analysis assumptions for any of the accidents previously evaluated. There will be no degradation in the performance of nor an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation.

The proposed change is still under staff review.

4. ITS 3.6.3.7 CTS SRs 4.6.1.7.2 and 4.6.1.7.4. Clarifv When Leakaae Rate Testino is Not Reauired (CN 7-10-LS-9 for CTS 3/4.6)

The proposed change will reduce requirements by adding a note, to CTS SRs 4.6.1.7.2 and 4.6.1.7.4, to state that leakage rate testing is not required for containment purge valves with WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

I i

i I

resilient seals when the penetration flow path is isolated by a leak tested blank flange. This is included in ITS SR 3.6.3.7. This is a change to the CTS and the STS.

The purpose of the leak testing requirement is to ensure containment leakage integrity during an accident, and thereby limit potential accident consequences. Isolation of the flow path with a l leak tested blind flange accomplishes this safety function and additionalleak testing of the valves in the flow path is redundant and unnecessary. The required action for a containment ventilation isolation valve (CVI) not within its leakage limit is revised to allow the penetration to be isolated using a closed and deactivated automatic valve, a closed manual valve or a blind flange and does not require the isolation valve to be restored to operable status. This is an l option not explicitly available in the CTS. The completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> remains the same as l in the CTS. If valves with resilient seals are used to isolate the flow path, the leakrate of these l valves must be verified at least every 92 days. If a leak tested blind flange is used to isolate the penetration flow path, the valves with resilient seals whose flow is isolated by the blind flange are not required to be leakrate tested. Isolation of the flow path with a leak tested blind flange provides the required leak barrier and additional leak testing of the valves in the flow path is redundant and unnecessary. Based on this, the proposed change is acceptable.

5. ITS SR 3.8.4.8 CTS SR 4.8.2.1.e. Revise Batterv Capacity Surveillance Test Acceptance Criteria. (CN 2-27-M for CTS 3/4.8)

CTS SR 4.8.2.1.e requires periodic completion of a battery performance discharge test. The performance test detects changes in battery capacity over time and trends overall battery degradation due to age and usage. The acceptance criteria in CTS 4.8.2.1.e for a performance discharge test is "80% of the manufacturer's rating." The proposed change is to revise this acceptance criteria to "85% of the manufacturer's rating"in corresponding ITS SR 3.8.4.8.

The proposed change is not in the STS. l i

The licensee proposed this change to reflect a recent design modification made by Wolf Creek that replaced the Gould manufactured square cell batteries with AT&T manufactured round cell  !

batteries. The AT&T round cell battery performance characteristics differ from the square cell batteries that were replaced. The square cell batteries are designed to exhibit a relatively l stable discharge capacity throughout their design life. The round cell battery capacity is designed to exhibit a gradual increase in capacity over time. Based on this, the expected l gradual decline in discharge capacity for square batteries would be abnormal for the installed round cell batteries. Therefore the CTS SR 4.8.2.1.e acceptance criteria of 80% of the manufacturer's rating is not appropriate for the round cell batteries, in that a reduction in l capacity would be indicative of abnormal battery performance.

The licensee's procurement requirements for the round cell batteries included a manufacturer test which demonstrated that the as-built batteries were at least 95% of the manufacturer's rating. The staff questioned why the discharge capacity acceptance criteria in CTS SR

! 4.8.2.1.e should not be revised to this value, i.e.,95% of the of the manufacturer's rating, since a decrease from this capacity value would be abnormal for the installed round cells. In response to the staff's concern, the licensee proposed that the CTS SR 4.8.2.1.e acceptance criteria be revised to "85% of the manufacturer's rating." This acceptance criteria is based on WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

consideration of the following: industry standards (IEEE 450 and IEEE 485), manufacturer's recommendations, margin in the plant specific Wolf Creek design, degradation mechanisms associated with round cells, industry experience with round cells, extensive manufacturer testing, other CTS surveillance test requirements, and the original procurement acceptance criteria.

This proposed change is still under staff review.

6. ITS SR 3.8.4.1 CTS SR 4.8.2.1.a.2. Decrease Minimum Batterv Terminal Voltaae. and ITS Table 3.8.6-1 CTS Table 4.8-2. Increase Minimum Float Voltaaes.

(CN 2-20-A for CTS 3/4.8)

The proposed change would increase the minimum battery cell float voltages for DC sources in CTS Table 4.8-2 by 0.01 to 0.02 volts. A corresponding change would be made to decrease the total required battery terminal voltage for a DC subsystem from 130.2 to 128.4 volts on float charge. The proposed changes are not in the STS.

Verifying the battery cell float voltages and total terminal voltage while on float charge helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended safety function. These proposed changes in minimum cell float voltage and corresponding total required battery voltage would reflect a recent design modification made by the licensee that replaced the Gould manufactured square cell batteries with AT&T manufactured round cell batteries. These proposed values are in accordance with IEEE 450, 1995 edition and the manufacturer's recommendations for float voltage. The proposed voltage values are based on the nominal design voltage of the batteries and are consistent with the initial voltages assumed in the battery sizing calculations. Based on this, the proposed changes i are acceptable.

l l 7. Additional BSis Not Yet Evaluated I

There are 15 additional beyond-scope issues (BSis) that have been recently identified by the  !

l licensee and the NRC staff, but have not yet been evaluated by the NRC staff. These j additional BSis resulted from the licensee's proposed changes to the STS (i.e., TSTFs) that will l not be approved by the staff before the licensee's conversion amendment may be issued by the

staff.

i These additional BSis are addressed with the justifications for the proposed changes to the CTS in the licensee's responses to requests for additional information (RAls) from the NRC staff. The additional BSis are listed in the table below with the associated change number, RAI number, RAI response submittal date, and description of the change.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

l Change RAI Licensee Description i Number / Question RAI Response ITS Section Date 4-05-LS-31 Q3.4.11-3 December 21, The actions of CTS LCO 3.4.4, for ITS 3/4.4 1998 inoperable power-operated relief valves and their associated block valves, to be in hot shudown were replaced by the requirement to reduce T,o to <500 F. For consistency, the actions of CTS LCO 3.4.7, for specific activity of the reactor coolant, was similarly revised and the time to reach the required T , was extended by 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

1-22-M Q3.3-49 November 24, The change is given in the application ITS 3/4.3 1998 Quarterly COTS have been added to CTS Table 4.3-1 for the power range neutron flux-low, intermediate range neutron flux, and source range flux trip functions. The CTS only require a COT prior to startup for these functions. New Note 19 is added to require that the new quarterly COT be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10 for the power range and intermediate range instrumentation (P-10 is the dividing point marking the Applicability for these trip functions), if not performed within the previous 92 days. New Note 20 i.s added such that the P-6 and P 10 interlocks are verified to be in their required state during all COTS on the power range neutron flux-low and intermediate range neutron flux trip functions.

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(

Change RAI Licensee Description Number / Question RAI Response ITS Section Date 1 LS-3 03.3-107 November 24, The changes are given in the application ITS 3/4.3 1998 and would (1) extended the completion time for CTS Action 3.b from no time specified to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for channel restoration or l changing the power level to either below P-6 or above P-10, (2) reduced the l

applicability of the intermediate range  !

neutron flux channels and deleted CTS Action 3.a as being outside the revised applicability, and (3) added a less restrictive new action that requires immediate suspension of operations irivolving positive reactivity additions and a power reduction below P-6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but no longer requires a reduction to Mode 3.

1-9-A Q5.2-1 September 24, New administrative change added to the ITS 5.0 1998 application. The CTS 6.2.2.e requirements concerning overtime would be replaced by a reference to administrative procedures for the control of working hours.

1-15-A 05.2-1 September 24, New administrative change added to the ITS 5.0 1998 application. The purposed change would revise CTS 6.2.2.G'to eliminate the title of Shift Technical Advisor. The engineering expertise is maintained on shift, but a separate individual would not be required as allowed by a Commission Policy Statement.

2-18-A 05.2-1 September 24, The proposed change is a revision to that ITS 5.0 1998 submitted in the application. The dose rate limits in the Radioactive Effluent Controls Program for releases to areas beyond the site boundary would be revised to reflect 10 CFR Part 20 requirements.

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WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

Change RAI Licensee Description Number / Question RAI Response l ITS Section Date 2-22-A 05.2-1 September 24, New administrative change added to the ITS 5.0 1998 application. The Radioactive Effluents Controls Program would be revised to include clarification statements denoting that the provisions of CTS 4.0.2 and 4.0.3, which allow extensions to surveillance frequencies, are applicable to these activities.

3-11-A 05.2-1 September 24, The proposed change is a revision to that ITS 5.0 1998 submitted in the application. CTS 6.12, which provides high radiation area access control alternatives pursuant to 10 CFR I

20.203(c)(2), would be revised to meet the current requirements in 10 CFR Part 20 and the guidance in NRC RG 8.3.8, on such access controls.

3-18-LS-5 05.2-1 September 24, Proposed change 3-18-A submitted in the ITS 5.0 1998 application was revised to be a new less restrictive change. The CTS 6.9.1.8 requirement to provide documentation of all challenges to the power operated relief valves (PORVs) and safety valves on the reactor coolant system would be deleted.

This is based on NRC Generic Letter 97-02 which reduced requirements for submitting such information to the NRC and did not '

include these valves for information to be submitted.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

i l

l Change RAI Licensee Description Number / Question RAI Response ITS Section Date i

9-17-LS-24 03.4.12-5 September 24, The proposed change is given in the l 1998 application and would add 4 notes to the I CTS LCO 3.4.9.3, to reflect CTS SR 4.5.3.2, LCO 3.5.4 actions, LCO 3.5.4 l

applicability notes, and the accumulator i action added in CN 9-10-M for CTS 3/4.4.

l Note 1 on centrifugal charging pump (CCP) l swap operations is a relaxation of the CTS l because it allows both CCPs to be capable of injecting into the RCS for up to 4 hourc throughout low temperature overpressure l protection (LTOP) applicability.

l 10 20-LS-39 03.7.10-14 October 16, The proposed change is given in the l ITS 3/4.7 1988 application and would revise and add an action to CTS LCOs 3.7.6 and 3.7.7, for ventilation system pressure envelope degradation, that allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore the control room pressure envelope through i repairs before requiring the unit to perform l an orderly shutdown. The new action has a longer allowed outage time than LCO 3.0.4 j which the CTS would require to be entered <

immediately. This change reconizes that the ventilation trains associated with the pressure envelope would still be operable. )

i 4-8-LS-34 0 3.4.11-2 September 24, The proposed change is given in the ITS 3/4.4 1998 application and would limit the CTS Srs 4.4.4.1 and 4.4.4.2 requirements to perform the 92 day surveillance of the pressurizer PORV block valves and the 18 month surveillance of the pressurizer PORVs (i.e., perforrn one complete cycle of each valve) to only Modes 1 and 2.

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50-Change RAI Licensee Description Number / Question RAI Response ITS Section Date 4-9-LS-36 Q3.4.11-4 September 24, Proposed Change in the application is ITS 3/4.4 1998 revised to add a note to Action d for CTS LCO 3.4.4 that would state that the action does not apply when the PORV block valves are inoperable as a result of power being removed from the valves in accordance Action b or c for an inoperable PORV.

1-60-A TR3.3-0073.3 December 21, A new administrative change is being ITS 3/4.3 1998 added to the application. The frequency for conducting the trip actuating device operational test (TADOT) for the turbine trip i of the reactor trip instrumentation surveillance requirements in CTS Table 4.3-1 would be changed from " prior to reactor startup" to " prior to exceeding the P-9 interlock whenever the unit has been in Mode 3."

1-70-M Q3.8.2-04 December 17, A new more restrictive change that is being ITS 3/4.8 1998 added to the application. The change will add shutdown requirements (including actions) for the load shedder and emergency load sequencer (LSELS) to CTS LCO 3.8.1.2 and surveillance requirements in SR 4.8.1.2 These requirements reflect current practice.

These additional BSis will be evaluated by the NRC staff and addressed in the final SE that will be issued by the NRC staff on the proposed conversion amendment.

5.0 COMMITMENTS RELIED UPON in reviewing the proposed ITS conversion for the WCGS, the staff has relied upon the licensee commitment to relocate certain requirements from the CTS to licensee-controlled documents as described in Table LG of Details Relocated from Current Technical Specifications and Table R of Relocated Current Technical Specifications attached to this SE. These tables reflect the relocations described in the licensee's submittals on the conversion. The licensee has been requested to submit a license condition to make this comrnitment enforceable. Such a commitment from the licensee is important to the ITS conversion because the acceptability of removing certain requirements from the TS is based on those requirements being relocated to WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

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licensee-controlled documents where further changes to the requirements will be controlled by the regulations (e.g., in accordance with 10 CFR 50.59).

6.0 LICENSE CONDITIONS There are problems with the first performance of the SRs in the ITS that will be new or revised l compared to the SRs in the CTS. The licensee should propose a license condition to define the schedule to begin performing the new and revised SRs during or after the implementation of the ITS. The staff has reviewed the following schedule for the licensee to begin performing the new and revised SRs and concludes that it is an acceptable schedule:

- For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of this amendment.

For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of this amendment.

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of this amendment.

For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the implementation of this amendment.

The licensee should also propose a license condition that will enforce the relocation of requirements from the CTS to licensee-controlled documents. The relocations are provided in Table LG of details relocated from CTS, Table R of relocated CTS, and Table LS of less restrictive changes to CTS. Table LS also contains relocations as, for example, the TR-1 changes that relocated the specific signals used to actuate the pumps and valves to the ITS Bases. The license condition should state that the relocations would be completed, during the implementation of the ITS.

7.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the ITS conversion amendment for the WCGS.

8.0 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 December XX,1998 (63 FR XXXXX), for the proposed conversion from the CTS to the ITS for the WCGS. Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.

WOLF CREEK GENERATING STATION DRAFT SAFETY EVALUATION

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l With respect to other changes included in the application for conversion to improved Technical Specifications, the items change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment required by these other changes involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. In ... sets of notices, the Commission issued proposed findings that the amendment required by these other changes involve no significant hazards consideration, and there has been no public comment on these findings published at: (a) 63 FR 53471 (October 5,1997), )

and . .. Accordingly, these changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the implementation of these changes.

9.0 CONCLUSION

l The NRC staff approves the licensee's changes to the WCGS CTS with modifications documented in the revised submittals. For the reasons stated infra in this SE, the NRC staff finds that the ITS issued with this license smendment comply with Section 182a of the Atomic Energy Act,10 CFR 50.36, and the guidance in the Final Policy Statement, and that they are in accord with the common defense and security and provide adequate protection of the health i and safety of the public.

The WCGS ITS provides clearer, more readily understandable requirements to ensure safer 1 operation of the station. The NRC staff concludes that the ITS satisfy the guidance in the Commission's Final Policy Statement, on technical specification improvements for nuclear power reactors, with regard to the content of TS, and conform to the STS provided in NUREG-1431 with appropriate modifications for plant-specific considerations. The NRC staff further concludes that the ITS satisfy Section 182a of the Atomic Energy Act,10 CFR 50.36, and other applicable standards. On this basis, the NRC staff concludes that the proposed ITS for the WCGS are acceptable.

The staff has also reviewed the plant-specific changes to the CTS as described in this SE. On the basis of the evaluations described herein for each of the changes, the NRC staff also concludes that these changes are acceptable.

The Commission has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security, or to the health and safety of the public.

Attachments: 1. Table A of Administrative Changes to Current Technical Specifications

2. Table M of More Restrictive Changes to Current Technical Specifications

3. Table LS of Less Restrictive Change to Current Technical Specifications WOLF CREEK GENERATING STATION oRAFT SAFETY EVALUATION

4. Table LG of Details Relocated from Current Technical Specifications

5. Table R of Relocated Current Technical Specifications Principal Contributors: N. Gilles C. Shiraki R. Tjader C. Schulten T.Liu R. Giardina J. Luehman A.Chu M.Reardon J,Donohew Date:

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