ML20236F802

Forwards Draft Safety Evaluation on Proposed Conversion of Current TS for CNS to Improved TS (Its).Se Based on Staff Review 970327 Application

ML20236F802
Person / Time
Site:	Cooper
Issue date:	06/24/1998
From:	Donohew J NRC (Affiliation Not Assigned)
To:	Horn G NEBRASKA PUBLIC POWER DISTRICT
References
TAC-M98317, NUDOCS 9807020353
Download: ML20236F802 (79)

Text

City -

l p+ 1 UNITED STATES sY t

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30086 0001

\*****/ June 24, 1998 Mr. G. R. Hom Sr. Vice President of Energy Supply Nebraska Public Power District 141415th Street Columbus, NE 68601

SUBJECT:

DRAFT SAFETY EVALUATION REGARDING PROPOSED CONVERSION TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS FOR THE COOPER NUCLEAR STATION (TAC NO. M98317)

Dear Mr. Hom:

Enclosure 1 is the draft Safety Evaluation (SE) on your proposed conversion of the current Technical Specifications (CTSs) for the Cooper Nuclear Station (CNS) to the improved Technical Specifications (ITSs). The ITSs are based on NUREG-1433, " Standard Technical Specifications, General Electric [GE] Plants, BWR/4," Revision 1, dated April 1995, and on guidance provided in the Commission's " Final Policy Statement on Technical Specifications improvements for Nuclear Power Reactors," published on July 22,1993 (58 FR 39132). The enclosed draft SE is based on the staff's review of your application dated March 27,1997 (NLS970002), as supplemented by the letters dated September 29,1997 (NLS970124), December 22,1997 (NLS970225), February 9,1998 (NLS980029), Mamh 13,1998 (NLS980046), March 26,1998 (NLS980057), April 16, 1998 (NLS980069), and May 6,1998 (NLS980065). The two letters dated September 29,1997, and April 16,1998, were concemed with the use of the GE Setpoint Methodology for the ITS and are addressed in Section Ill.G.2 of the draft SE on ITS Section 3.3. The staff issued requests for additional information (RAls) dated November 6,1997, December 22,1997, and March 19,1998.

The enclosed draft SE is for your review to verify its accuracy and to prepare the certified ITSs for CNS. You are requested to provide your comments on the draft SE in writing, and a certified ITSs and Bases to the ITSs, within 30 days of receipt of this letter. Your submittal should also address the two open items in Sections Ill.G.5 and Ill.G.7 of the draft SE, and include the changes to the ITSs that have been discussed with the staff since the last letter dated May 6, 1998. After the staff has reviev,ed your comments, it will incorporate changes, as appropriate, in the final SE before issuing the ITSs and the final SE. The conclusions of the NRC staffin the enclosed draft SE are not valid until the final SE is issued.

You are also requested to submit a license condition for an Appendix C to the CNS license to make enforceable the transfer of those requirements in the CTSs being transferred into licensee-controlled documents (i.e., documents, such as the Updated Safety Analysis Report (USAR), for which changes to the documents by licensees are controlled by the regulations, in this case 10 CFR 50.59) for the ITS conversion, as described in your letters and the enclosed draft SE.

Enclosure 2 is an acceptable license coridition. A similar license condition to Enclosure 2 should also be submitted for (1) each commitment to complete a future action that you included in your above lotters on the ITSs for CNS (e.g. the commitment to complete ITS setpoint calculations in accordance with the GE setpoint methodology), 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 V of the enclosed draft SE.

1 .

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

Sincerely, I

OdSp/ lo/24/'l?

( Jack N. Donohew, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects ill/IV Office of Nuclear Reactor Regulation l

Docket No. 50-298

Enclosures:

1. Draft Safety Evaluation

2. Acceptable License Condition cc w/encis: See next page DISTRIBUTION:

Docket File PUBLIC PD4-1 r/f T. Collins C.Hawes ACRS OGC (15B18) C. Miller T. Gwynn, RIV J. Hall E. Adensam (EGA1) W. Beckner J.Hannon J.Donohew L. Marsh J. Wermiel Document Name: COOes3ir.tra 4 cenCs uEuo oATEo uaY 29.189.>

orc puso41 m twD4.1 seNB BC:SRXB BC HICB , BC TSB _ PD/PDfV ,1, NAME X CHawes W TComne C J ff WB JHannon oare 6,1%, b i6. 6 as T < . 1, m. 4 1 / 91., tai 2 % Ci23, CONC ITS 5.5 9 M4 ITS 3.1.8 ITS 3.3.3. 2 LA1

1. D1 ITil 3.3.2.1 LA2 ITS 3.5.1 JrD1 [

YES h COPY YESNO YESNO YESINO YESNO YESNO YESNO

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/f6l oaTE ll% i CONC ITS 5 7 COPY b ' * }

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OFFICIAL RECORD COPY 1

-- - - - - - - -- -- - - - - - - - _ - - - -- - A

. Mr. G. R. Hom Please do not hesitate to contact me at 301-415-1307 (orjnd@nrc. gov on the Intemet) if you have any questions.

Sincerely,

{

. ecemn Jack N. Donohew, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects ill/IV J Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

1. Draft Safety Evaluation

2. Acceptable License Condition ec w/encis: See next page 1

l

Mr. G. R. Hom

- Nebraska Public Power District Cooper Nuclear Station cc:

Mr. John R McPhail, General Counsel Lincoln Electric System Nebraska Public Power District ATTN: Mr. Ron Stoddard P. O. Box 499 1040 O Street Columbus, NE 68602-0499 Box 80869 )

Lincoln, NE 68501 Nebraska Public Power District ATTN: Mr. J. H. Swailes MidAmerican Energy Vice President of Nuclear Energy ATTN: Dr. William D. Leech, Manager-Nuclear P. O. Box 98 907 Walnut Street I

Brownville, NE 68321 P. O. Box 657 Des Moines, IA 50303-0657 Randolph Wood, Director Nebraska Department of Environmental Nebraska Public Power District Control ATTN: Mr. B. L Houston, Nuclear '

P. O. Box 98922 Licensing & Safety Manager Lincoln, NE 68509-8922 P. O. Box 98 Brownville, NE 68321 Mr. Larry Bohlken, Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Aubum, NE 68305 Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 218 Brownville, NE 68321 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington,TX 76011 Ms. Cheryl Rogers, LLRW Program Manager Division of Radiological Health Nebraska Department of Health 301 Centennial Mall, South P. O. Box 95007 Lincoln, NE 68509-5007 1 i

Mr. Ronald A. Kucera, Department Director i ofIntergovemmental Cooperation  ;

Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 f

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DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. TO FACILITY OPERATING LICENSE DPR-46  !

l COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298

1. INTRODUCTION Cooper Nuclear Station (CNS) has been operating with Technical Specifications (TS) issued with the original operating license on January 18,1974, as amended from time to time. By lettet dated March 27,1997, as supplemented by letters dated September 29,1997, December 22,1997, February 9,1998, March 13,1998, March 26,1998, April 16,1998, and May 6,1998, Nebraska Public Power District (NPPD or the licensee) proposed to convert the current Technical Specifications (CTS) to the improved Technical Specifications (ITS). The conversion is based upon NUREG-1433, " Standard Technical Specifications for General Electric IGE] Plants, BWR/4," Revision 1, dated April 1995, and upon guidance in the "NRC Final Policy Statement on Technical Specification improvements for Nuclear Power Reactors" (Final Policy Statement), published on July 22,1993 (58 FR 39132), and 10 CFR 50.36, as amended July 19,1995 (60 FR 36953). The overall objective of the proposed amendment, consistent with the Final Policy Statement, was to rewrite, reformat, and streamline the TS for CNS to be in accordance with 10 CFR 50.36, " Technical Specifications." The two letters dated September 29,1997, and April 16,1998, were concerned with the use of the General Electric Setpoint Methodology for the ITS trip setpoints and allowable values and are addressed in Section 3 G on ITS Section 3.3.

In the letter of July XX,1998, the licensee proposed license conditions for the implementation of the ITS conversion. Also, in this letter, the licensee submitted revised ITS pages. The license conditions and revised ITS pages do not change the notice in the FederaI Reaister on March 17,1998 (63 FR 13074), for the conversion from the CTS to the ITS for the CNS. In addition to this notice, there were the five notices in the Federa/ Reaister for the beyond scope issues associated with the conversion that are discussed in Section Ill.G of this safety evaluation (SE).

There was one notice on April 22,1998 (63 FR 19971) and four notices on May 6,1998 (two notices on 63 FR 25111, and two on 63 FR 25112).

Hereafter, the proposed or improved TS for the CNS are the ITS, the existing TS are the CTS, and the improved standard TS, such as in NUREG-1433 for CNS, are the STS. The corresponding TS Bases are ITS Bases, CTS Bases, and STS Bases, respectively.

ENCLOSURE 1 Cooper Nuclear Station Draft Safety Evaluation i

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- 2-In addition to basing the ITS on the STS, the Commission's Final Policy Statement, and the

. requirements in 10 CFR 50.36, the licensee retained portions of the CTS as a basis for the ITS.

Plant-specific issues, including design features, requirements, and operating practices, were discussed with the licensee during a series of conference calls and meetings that concluded on April 7,1998 (the meeting summaries were issued on June XX,1998). Based on these

! discussions, the licensee proposed matters of a generic nature that were not in STS. The NRC i

staff requested that the licensee submit such generic issues as a proposed change to STS through the NRC/ Nuclear Energy institute's Technical Specifications Task Force (TSTF). ,

These generic issues were considered for specific applications in the CNS ITS. Consistent with

( the Final Policy Statement, the licensee proposed transferring some CTS requirements to l licensee-controlled documents (i.e., documents, such as the Updated Safety Analysis Report (USAR) for the CNS, 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),

l whereas 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 ITS. Further, significant changes were proposed to the CTS Bases to L

make each ITS requirement clearer and easier to understand.

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1 The Commission's proposed action on the CNS application for an amendment dated i March 27,1997, was published in the Federa/ Register on March 17,1998 (63 FR 13074). The {

Staff's evaluation of the application, including the supplements listed above, that resulted from NRC requests for information (RAls) and discussions with the licensee during the NRC staff l- review, is presented in this SE. The staff issued RAls dated November 6,1997, December 22, ,

j 1997, and March 19,1998. These plant-specific changes serve to clarify the ITS with respect {

! to the guidance in the Final Policy Statement and STS. Therefore, the changes are within the l scope of the action described in the Federal Register notice, except for the beyond scope issues that were the subject of separate notices discussed above.

In addition, since the application for an amendment was published, Amendment No.s 176 and 177 to the CNS operating license were approved on May 9,1997, and June XX,1998, respectively. These amendments (1) relocated the surveillance requirements for the standby liquid control (SLC) system relief valves from the CNS TS to the USAR and the CNS inservice testing program and (2) incorporated thermal hydraulic stability changes to the CTS. The licensee has incorporated these amendments as appropriate into the ITS.

l During its review, the NRC staff relied on the Final Policy Statement and the STS as guidance for acceptance of CTS changes. This SE provides a summary basis for the NRC staff conclusion that CNS can develop ITS based on STS, as modified by plant-specific changes, and that the use of the ITS is acceptable for continued operation. The NRC staff also acknowledges that, as indicated in the Final Policy Statement, the conversion to STS is a voluntary process. Therefore, it is acceptable that the ITS differs from STS, reflecting the current licensing basis for the CNS. The NRC staff approves the licensee's changes to the CTS with modifications documented in the revised submittals.

Cooper Nuclear Station Draft Safety Evaluation

For the reasons stated infra in this SE, the NRC staff finds that the ITS issued with this license amendment 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 and safety of the public.

11. BACKGROUND l

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

! [S]uch technical specifications, including information of the amount, kind, and l 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 i 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 l

items in the following five specific categories: (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 included 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-1433 for GE BWR/4's, that would establish mode's of the Commission's l 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 l preparing technical specifications, which gives greater consideration to human factors principles and was used throughout the development of licensee-specific ITS.

In September 1992, the Commission issued NUREG-1433, which was developed using the guidance and criteria contained in the Commission's Interim Policy Statement. The STS in NUREG-1433 were established as a model for developing the STS for GE BWR/4 plants in Cooper Nuclear Station Draft Safety Evaluation l 1

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! , 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. STS also reflect the results of extensive discussions conceming 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-1433 apply to the CNS.

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 enhe STS, and encouraged licensees to use the STS as the basis for plant-specific TS amendd.ents, .

and for complete conversions to ITS based on the STS. Further, the Final Policy Statement

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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 surveillance 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 Porfland Genera / Electric 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 I 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 discem it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigio 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 Final Policy Statement criteria are as follows:

Criterion 1 l

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

Cooper Nuclear Station Draft Safety Evaluation l

, 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 .

l- 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 r

" either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

I Part til of this SE explains the NRC staff conclusion that the conversion of the CNS CTS to those based on STS, as modified by plant-specific changes, is consistent with the CNS current licensing basis and the requirements and guidance of the Final Policy Statement and 10 CFR l 50.36.

Ill. EVALUATION l The NRC stats ITS review evaluates changes to CTS that fall into five categories defined by I

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 staffs plans for monitoring the licensee's implementation of these controls at CNS.

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 y

~ ITS. Each change proposed in the amendment request is identified as either a discussion of 1 l . change (DOC) to the CTS or a justification for deviation (JFD) from the STS. The NRC staff comments were documented as RAI and forwarded in letters dated November 6,1997, l December 22,1997, and March 19,1998. The licensee provided responses in letters dated l December 22,1997, February 9,1998, March 26.1998, and May,6,1998. The letters clarified and revised the licensee 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.-

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

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Cooper Nuclear Station Draft Safety Evaluation

, (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 CTS requirements; (3) Technic al Changes - Less Restrictive (specific), (L), i.e., changes, deletions and relaxations of CTS requirements; (4) Technical Changes - Less Restrictive (generic), (l.A), i.e., deletion of CTS requirements by movement of information and requirements from existing specifications (that are otherwise being retained) to licensee-controlled documents, including the ITS Bases; and (5). Relocated Specifications, (R), i.e., relaxations in which whole specifications (the LCO, and associated action and SR) are removed from the CTS (an NRC-controlled document) and placed in licensee-controlled documents.

Tnese 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 editorial in nature or involve the reorganization or reformatting of CTS requirements without affecting technical content or operational restrictions. Every section of the ITS reflects this type of change. In order to ensure consistency, the NRC staff and 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-l.pecific 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; Cooper Nuclear Station Draft Safety Evaluation

, (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 current interpretation and practice, and that more clearly or explicitly state existing requirements; and (8) deletion of redundant TS requirements that exist elsewhere in TS.

Table A of CNS Administrative Changes lists the administrative changes proposed in ITS.

Table 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 the licensee and finds them acceptable, because they are compatible with the Writer's Guide and STS, do not result in any change in operating requireme 1ts and are consistent with the Commission's j

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 l equipment, and more restrictive SRs. Table M of CNS More Restrictive Changes lists the more l restrictive changes proposed for the ITS. 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) i Less restrictive requirements include changes, deletions and relaxations to portions of the CTS l requirements that are not being retained in ITS. When requirements have been shown to give l little or no safety benefit, their removal from the TS may be appropriate. In most cases, j 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 l j acceptable because they are consistent with current licensing practices and the Commission's regulations. The CNS design was also reviewed to determine if the specific design basis and i licensing basis for the CNS are consistent with the technical basis for the model requirements in the STS, and thus provide a basis for the ITS.

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. l 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:

i Category l Relaxation of CTS LCO Applicability l

l Category 11 Relaxation of CTS Surveillance Frequency Category lli Relaxation of CTS Required Actions l

l Category IV Relaxation of CTS Required Action Completion Time Category V Relaxation of CTS Surveillance Requirement Acceptance Criteria Category VI Relaxation of CTS Setpoints to Allowable Values Category Vil Relaxation of CTS Action Entry to Perform Surveillance Requirements l 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.

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Relaxation of CTS LCO Analimhilitv (Category l}

Reactor operating conditions are used in the CTS to define when the LCO features are required to be Operable. The LCO applicabilities can be specifically defined terms of

! reactor conditions: hot shutdown, cold shutdown, reactor critical, or power operating l condition. Applicabilities can also be more general. Depending on the circumstances, i CTS may 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 ITS eliminate the CTS requirements such as "all modes" or "any  !

l operating mode," replacing them with ITS defined modes or applicable conditions that

! are consistent with the application of the plant safety analysis assumptions for l operability of the required features.

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l In another application of this type of change, CTS requirements may be eliminated i

! 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 indeterminant or which are inconsistent with application of i accident analyses assumptions is acceptable because when LCOs cannot be met, the TS can be satisfied by exiting the applicabitty thus taking the plant out of the conditions that require the safety system to be ope'able. Therefore, changes falling within i Category I are acceptable.

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. _ _ _ _ _ _ . __ _ .-. - ____ _ ___ _=____-___ __ _ _ _ __--__--

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, Relaxation of CTS Surveillance Freauency (Category ll) l CTS and ITS surveillance frequencies specify time interval requirements for performing surveillance requirement testing. Increasing the time interval between surveillance tests in the ITS results in decreased equipment unavailability because of testing which increases equipment availability. 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-l acceptable based on similar design, like-component testing for the system application, j and the availability of other TS requirements which provide regular checks to ensure l . limits are met.

f i_ Reduced testing can enhance safety because it reduces system unavailability from l testing; in turn, reliability of the affected structure, system or component should remain constant or increase. Reduced testing is acceptable where operating experience, l

industry practice, or industry standards, such as manufacturers' recommendations, have shown that components usually pass the surveillance when performed at the specified interval. Therefore, the frequency is acceptable from a reliability standpoint.

l Surveillance frequency changes to incorporate attemate train testing has been shown to i

be acceptable where other qualitative or quantitative test requirements are required i-which are established predictors of system performance (e.g., a 31-day air flow test is an indicator that positive 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).

l Additionally, surveillance 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 11 are acceptable.

Relaxation of CTS Reauired Actions (Category ill) {

Upon discovery of a failure to meet an LCO, the STS specify required actions to i 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 lli are acceptable.

Cooper Nuclear Station Draft Safety Evaluation l

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Relaxation of CTS Reauired Action Comoletion Time (Category IV)  !

L Upon discovery of a failure to meet an LCO, the STS specify times for completing .

required actions of the associated TS conditions. Required actions cf 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 l condition is permitted. Adopting completion times from the STS is acceptable because i 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 I i basis accident (DBA) occurring during the repair period. Therefore, changes falling j within Category IV are acceptable.

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Relaxation of CTS Surveillance Requirement Accentance 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 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 l

replaced in the ITS with separate and distinct testing requirements which, when l

combined, include operability verification of all TS-required components for the features specified in the CTS. Adopting this format preference in the STS is acceptable because SRs that remain include testing of all previous features required to be verified operable. j Therefore, changes falling within Category V are acceptable. ,

1 l Relaxation of CTS Setooints to Allowable Values (Category Vi )

l l Allowable values for the tnp settings of instrumentation establish operability limits for '

i that instrumentation. These allowable values have been established with the methods  !

described in General Electric (GE) Instrument Setpoint Methodology (NEDC-31336), GE '  ;

Document EDE-38-1090 (Setpoint Calculation Guidelines for the Cooper Nuclear l

Station), and CNS procedure 3.26.3, " Instrument Setpoint and Channel Error Calculation Methodology." The allowable value determinations were performed uring vendor documented performance specifications, where applicable and available. This is discussed ir Section Ill.G of this SE for ITS Section 3.3.

Where vendor documented performance specifications for drift were not available or applicable, the allowable value was determined using plant specific operating and surveillance trend data or an allowance as provided by the CNS Instrument Setpoint Methodology. The allowable value verification employed actual plant-specific trend information to ensure the validity of the results. All changes to safety analysis limits  ;

applied in the methodology were evaluated and confirmed to maintain the licensing acceptance limits. All design limits applied in the methodologies were confirmed as ,

ensuring that applicable design requirements of the systems are maintained. Plant Cooper Nuclear Station Draft Safety Evaluation

, calibration procedures ensure that the assumptions regarding calibration accuracy, measurement and test equipment accuracy, and setting tolerances are maintained.

Setpoints for each design or safety analysis limit have been established by accounting for the applicable instrument accuracy, calibration and drift uncertainties, environmental effects, power supply fluctuations, as well as uncertainties related to process and primary element measurement accuracy using CNS Instrument Setpoint methodology.

The allowable values have been established from each design or safety analysis limit by combining the errors associated with channel / instrumentation calibration (e.g. device accuracy, setting tolerance, drift) with the calculated Nominal Trip Setpoint using CNS Instrument Setpoint methodology. As a result, the revised allowable values ensure that the design basis and associated safety limits will not be exceeded during plant operation. Therefore, changes falling within Category VI are acceptable.

Relaxation of CTS Action Entry to Perform SRs (Category Vil)

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 generally six hours, during which time the functional capability is maintained. This relaxation is in accordance with approved Topical Reports that apply to CNS. 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 Vil are acceptable.

Table L of CNS Less Restrictive Changes is organized in iTS order by each L-type DOC 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). For ease of reference, the seven less restrictive change categories are listed at the bottom of each page of Table L.

Additionally, in electing to implement the specifications of STS, the licensee also proposed a number ofless restrictive changes to the CTS which do not apply to the above categories of changes, deletions and relaxations of CTS requirements. These changes are characterized as unique in Table L and are evaluated below. Each evaluation is preceded by the ITS section or specification and the DOC identifier (e.g., L.1) associated with the change. All of these changes to the CTS were consistent with the STS and, therefore, were not beyond the scope of the ITS conversion. The changes that were beyond the scope of the conversion are addressed in Section Ill.G of this SE.

ITS Section 1.0 L.1 The STS definition of Core Alterations is proposed for the iTS and is less restrictive than the corresponding CTS definition of Alteration of the Reactor Core, because it will only Cooper Nuclear Station Draft Safety Evaluation

, apply to those activities that create the potential for a reactivity excursion and thus warrant special precautions or controls in the ITS. The CTS requirement that 'the act of moving any component in the region above the core support plate, below the upper grid and within the shroud" is a core alteration will be deleted because this definition is too encompassing. 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 control 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).

However, movement of control rods with the control rod hydraulic system, that is exempted from the current definition of Core Alteration, does create the potential for a reactivity excursion and is an activity that warrants special precautions. Appropriately, normal control rod movement with the head removed is included in the proposed definition of Core Alterations because the poteratial for a reactivity excursion exists.

Control rod movement is not considered a Core Alteration provided that the four fuel bundles surrounding an associated control rod are removed. This is acceptable because removal of these four fuel bundles reduces the reactivity worth of the associated control rod such that complete removal of the control rod no longer has the potential to cause a reactivity excursion. The design of the control rod velocity limiter precludes complete removal of a control rod prior to the removal of the four associated fuel bundles. Therefore, the inclusion of normal control rod movement with the head removed in the definition of Core Alterations is acceptable.

L.2 The temperature and pressure requirements contained in the CTS 1.0.G definition of hot standby condition is proposed to be omitted from the ITS definition of Mode 2, startup with the mode switch in startup / hot standby, in ITS Table 1.1-1,

• MODES." This change is consistent with the STS and is acceptable because pressure and temperature will be adequately controlled by other ITS requirements. Reactor coolant system (RCS) temperature will be controlled via the RCS pressure / temperature (P/T) limits curve which is addressed by ITS 3.4.9. And omission of the CTS definition's requirement to maintain pressure below 1000 psig is acceptable because ITS 3.4.10 will limit reactor steam dome pressure to 1020 psig in Mode 2. This specification and the 1045 psig trip setpoint of the reactor protection system (RPS) high pressure trip, which would trip the reactor if pressure reached 1045 psig, will ensure that reactor vessel pressure does not exceed the maximum pressure assumed in the safety analysis.

L.5 The requirement contained in the CTS definition of cold shutdown that the reactor vessel be vented is proposed to be omitted from the ITS definition of Mode 4, cold Cooper Nuclear Station Draft Safety Evaluation

l 1.

l l . 13-i shutdown, in ITS Table 1.1-1. This is consistent with the STS and is acceptable because the requirements of ITS 3.4.9, RCS P/T Limits, will preclude the reactor vessel from exceeding reactor vessel pressure limits in Mode 4. Also, the requirements of ITS 3.4.8, " Residual Heat Removal - Cold Shutdown," will ensure adequate decay heat removal capabilities in Mode 4. Therefore, the reactor vessel venting requirement is unnecessary in the definition of Mode 4 and may be deleted.

ITS Section 2.0 L.1 This proposed change will delete the power transient safety limit (SL) requirements in CTS 1.1.C.' This SL is not in the STS. The intent of this requirement is to ensure that SLs are not exceeded and this requirement states that SLs are assumed to be exceeded when a scram is accomplished by means other than the expected scram signal. The scram setpoints are established in order to ensure margin to the SLs.

Exceeding the scram setpoint, in and of itself, does not necessarily indicate that a SL has been exceeded. Sections 2.1.A and 2.2.A of the CTS contain eleven trip settings that initiate a reactor scram. These scram settings are included in ITS Table 3.3.1.1-1.

The SR imposed on these scram setpoints in ITS Table 3.3.1.1-1 help to ensure that the margin to a safety limit is preserved. The redundancy built into the reactor protection system (RPS) is maintained by the action of ITS 3.3.1.1. Therefore, the safety intent of power transient F:. require nents in CTS 1.1.C is adequately maintained by ITS 3.3.1.1 for the RPS an'.i the proposed change is acceptable.

L.2 ~ The CTS St., in CTS 1.1.D, for the reactor vessel water level requires that the level be not less thon 18 inches above the top of the normal active fuel zone in the cold shutdown condition (ITS Modes 4 and 5). ITS SL 2.1.1.3 requires that the level be greater ttian the top of the active irradiated fuel in all modes. The reference to the top of active irradiated fuel from the STS is a more restrictive change for mode applicabi!ity; however, it is less restrictive for cold shutdown because the ITS SL is 18 inches less than the former limit. The change ensures adequate margin for effective action in the event of a level drop.

Once the reactor has been operated at core ly 21.0, CTS 1.1.D indicated the " top of the normal active fuel zone" and the " top of active fuel (TAF)" are one and the same.

Therefore, ITS 2.1.1.3 speaks of TAF as the " top of active irradiated fuel."

As the ITS 2.1.1.3 Bases states (and station design and operating license bases l conservatively confirm), below 2/3-core height is where elevated cladding temperatures and clad perforation would occur from decay heat without adequate cooling capability.

l The ITS lowest actuation levels of the emergency coolant systems are 95.19 inches I above 2/3-core height in all Modes.

The reactor vessel shroud level - level O conservatively corresponds to the tops of the jet pump risers. Below the tops of the jet pump risers, the only mechanism for loss of reactor water level, short of an extremely unlikely vessel failure or draining, is boiling Cooper Nuclear Station Draft Safety Evaluation

due to the irradiated fuel's intemal decay heat. This level is actually 15 inches above the true 2/3-core height. Two-thirds core height is 50 inches below TAF. Therefore, if a loss of vessel water level occurs at TAF, there is yet 50 inches more before getting to the 2/3-core-height level.

1 This provide sufficient time, in all modes, to take effective action for maintaining or restoring the ITS SL water level greater than the " top of active irradiated fuel" by using other water injection methods and sources. This is also true in ITS Mode 5 with the vessel head removed and with personnel working on the refueling bridge, after their first seeing the exposed tops of the handles of the seated fuel assemblies. Even in Mode 5, there are other monitoring methods and alarms of a loss of water level available to allow for effective action before getting to the ITS SL limit of " greater than the top of active irradiated fuel."

Therefore, the proposed change is acceptable.

ITS Section 3.0 L.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 l TS preclude the retum 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 retumed to service is very small, therefore, the probability of an  !

accident during that time period is also very small and insignificant. Therefore, l' proposed STS LCO 3.0.5 is acceptable.

ITS Specification 3.1.1 The requirement in CTS 3/4.3.A.1 that the SDM must always be 2 0.38% Ak/k is I L.1 proposed to be reduced in ITS LCO 3.1.1.b and SR 3.1.1.1.b in that the SDM could be a 0.28% Ak/k when the highest worth control rod is determined by testing. The reduced requirement is from the STS. The requirement for the SDM to be a 0.38% Ak/k would remain in the ITS for when the highest worth control rod is analytically determined. The 4

Cooper Nuclear Station Draft Safety Evaluation

SDM is based on the highest worth control rod being fully withdrawn. The CTS did not make a distinction between determining the highest worth control rod by analytical or testing means; however, the STS makes this distinction and the ITS would allow a lower l

SDM limit if the highest worth control rod is determined by testing. This lower SDM limit provides adequate shutdown margin and is acceptable because the lower SDM limit 1 would be based an measured value obtained for the highest worth control rod versus an analytical one which may contain uncertainties that have to be accounted for in the analysis. The measured value does not need to take into account analytical uncertainties which are greater than the measurement uncertainties. Therefore, the proposed ITS LCO 3.1.1.b and SR 3.1.1.1.b are acceptable.

ITS Specification 3.1.3 L.8 The CTS 3/4.3.B.2 requirement for the control rod drive (CRD) housing support to be in place is included in the CTS operability requirements for control rods, but is proposed to be omitted from the ITS. The STS does not have this requirement. Station configuration management provides adequate controls to assure the CRD housing support is in place. The CTS requires inspections of the CRD housing support following reassembly. The CTS requirement verifies that the CRD housing support is in place for reactor operation in Modes 1,2 and 3. Post-maintenance inspections conducted -

through station configuration management control have the same function as the CTS

]

requirement. Since work is not normally performed on the CRD housing support at power, and checks on its installation are not made at power, there is no current requirement to verify CRD housing support installation in power operating conditions.

' Therefore, the deletion of this CTS requirement is acceptable based on use of station  !

configuration management control to ensure proper CRD housing support installation. j ITS Specification 3.1.4 l l

L.1 The CTS 4.3.C.2 requirement for an evaluation to be made, whenever scram time surveillance are performed is proposed to be omitted from the ITS. The STS does not have this requirement. This evaluation is to provide reasonable assurance that proper control rod drive performance is being maintained; however, it is essentially a performance tracking requirement to help ensure control rod scram times are maintained within limits. The ITS 3.1.4 and associated SRs are adequate to ensure that scram time testing is performed and scram times are maintained within limits. In  ;

addition, the requirements of 10 CFR 50.65, requirements for monitoring the i effectiveness of maintenance at nuclear power plants, and the CNS implementation of these requirements, ensure equipment important to safety is adequately maintained (in this case, that control rod drive performance is being maintained).10 CFR 50.65 requires monitoring of the performance or conditions of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide assurance that such structures, systems, and components are capable of fulfilling their intended function. Compliance with 10 CFR 50.65 is required by the CNS operating license. Therefore, explicit control rod drive performance trending SRs are not required l

l Cooper Nuclear Station Draft Safety Evaluation l l l

)

to ensure control rod scram times are maintained within limits and are not included in the ITS. Therefore, not including this performance tracking requirement in the ITS is acceptable.

ITS Specification 3.1.5 L.2 The CTS 4.3.A.2.c requirement of a check of the status of the pressure and level alarms for each control rod scram accumulator once per week is proposed to be omitted from the ITS. The STS does not have this requirement. ITS SR 3.1.5.1 includes the acceptance criteria for accumulator pressure (a 940 psig); however, the STS do not specify indication-only or test equipment to be operable to support operability of a system or component. The control rod scram accumulator level alarms and pressure alarms do not necessarily relate directly to accumulator operability. Control of the availability of, and necessary compensatory activities, for alarms, are addressed by station procedures and policies. The requirement to verify control rod scram accumulator pressure (which does relate directly to accumulator operability) is within limits is in SR 3.1.5.1. Therefore, the requirements associated with the control rod accumulator pressure and level alarms can be omitted from the ITS.

ITS Specification 3.1.8

. L.3 The CTS 4.3.G.3 requirement of verification of scram discharge volume (SDV) vent and drain valve operability following any maintenance or modification to any portion of the SDV affecting the operation of the vent and drain valves is proposed to be omitted from the ITS. The STS does not have this requirement. Any time the operability of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate operability of the system or component. After restoration of a component that could cause a required SR to be failed, ITS SR 3.0.1 requires the appropriate SRs (in this case, ITS SRs 3.1.8.2 and 3.1.8.3 as applicable) to be performed to demonstrate operability of the affected components. Therefore, the explicit post maintenance SRs in the CTS are not needed and can be omitted from the. ITS.

ITS Specification 3.2.2 L.1 The CTS 4.11.C requirement that the minimum critical power ratio (MCPR) be verified following any significant change in power level or distribution that could cause ope, ration to be on the MCPR operating limit is proposed to be omitted from the ITS. The STS does not have this requirement. Operation of the core on the MCPR operating limit is extremely unlikely and the surveillance is seldom required. Deletion of the CTS 4.11.C SRs does not remove the requirement to maintain the MCPR within limits and ITS SR 3.2.2.1 require verification of the MCPR once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after exceeding 25% rated  !

thermal power (RTP) and every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. Therefore, deletion of this surveillance frequency in CTS 4.11.C is acceptable.

Cooper Nuclear Station Draft Safety Evaluation :

. . ITS Specification 3.3.1.1 L.8 The information contained in CTS Table 4.1.1, Note 5 conceming testing the RPS channel test switch after maintenance is proposed to be deleted. The STS does not have this requirement. Any time the operability of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate operability of the system or component. After restoration of a component that could cause a required SR to be failed, ITS SR 3.0.1 requires the appropriate SRs (in this case, SR 3.3.1.1.4) to be performed to {

demonstrate operability of the affected components. Therefore, an explicit post j maintenance SR is not needed and can be deleted from the ITS.' Entry into the J applicable modes is addressed in the Bases for ITS SR 3.0.1.

L.12 The proposed change would add a note from the STS to the 18-month channel calibration surveillance requirement in CTS Table 4.1.2 for the APRM neutron flux-high (flow biased) function (ITS SR 3.3.1.1.12) and the 18-month response time SR in -

' CTS Table 4.1.2, Note 5 (ITS SR 3.3.1.1.15). The note is less restrictive than the CTS table note 5 by excluding the neutron detectors from these ITS surveillance. The channel calibration is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy. The response time test (RTT) measures the instrument response time. The neutron detectors are excluded from the channel calibrations and RTT because they are passive devices with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performance of the 7-day calorimetric calibration (ITS SR 3.3.1.1.2) and the 1000 MWD /T local power range monitor (LPRM) calibration against the traversing incore probes (TIP) (ITS SR 3.3.1.1.8). Therefore, the addition of this note to ITS SRs 3.3.1.1.12 and 3.3.1.1.15 is acceptable.

ITS Specification 3.3.1.2 l

L2 The requirements in CTS 3.10.B for two source range monitors (SRM) during core alterations and in CTS 3.10.B.3 on the SRM count rate during spiral unloading are proposed to be omitted from the ITS. If a spiral offload or reload pattem is used, the ITS .

will allow a reduction in the number of SRM channels required to be operable during refueling. ITS 3.3.1.2 (Table 3.3.1.2-1 footnote (b) from the STS) will reduce the required number of SRM channels to be operable from 2 to 1 "during spiral offload or i reload when the fueled region includes only that SRM detector." A reduction in the

~

number of required operable SRM channels is acceptable when using a spiral pattem for loading or offloading fuel because the use of a spiral pattem provides assurance that  ;

the operable SRM is in the optimum position for monitoring changes in neutron flux '

levels resulting from the core alteration. In addition, ITS SR 3.3.1.2.2.a has been added to ensure an SRM is in the fueled region, and Note 2 to SR 3.3.1.2.2. has been added for clarity to allow the one SRM to satisfy more than one SRM location requirements.  ;

Due to these additions, CTS 3.10.B.3, which allows the SRM count rate to not be met l Cooper Nuclear Station Draft Safety Evaluation

- _ _ _ - _ - - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _._.__..__n_ _ - _ _ _ . _ - - - _ _ _ _ _ _ _ _ . - _ _ _ _ . _ . _

, during spiral unloading, is not needed and is proposed to also be deleted. This change is incorporated since during spiral unloading, only one SRM may have fuel around it, thus the other SRM currently required would not meet the count rate requirement. Since only one SRM is now required in this instance, the allowance has been deleted.-

Therefore, the deletion of these requirement in CTS 3.10.B and 3.10.B.3 is acceptable.

ITS Specification 3.3.2.1 L.1' The requirement in CTS 4.2.C (Table 4.2.C) of an instrument check (or channel check) j for the rod block monitor (RBM) Upscale, downscale and inoperative functions once per day is proposed to be deleted. As taken from the STS, ITS LCO 3.3.2.1 does not require a channel check of these functions. The RBM automatically re-nulls itself whenever a control rod is selected and retains the latest setting until another control rod is selected, making a channel check both impractical and meaningless. Proper functioning of the RBM is indirectly verified each time a control rod is selected however, by the re-nulling of the setpoint. Therefore, the elimination of a channel check for the RBM is acceptable.

L.5 The requirement in CTS 4.3.B.3.b.5 to perform an instrument functional test on the only operable RBM channel"during operation when a Limiting Control Rod Pattem for RWE

[ rod withdrawal error) exists"is proposed to be deleted because performing a functional test due to one channel being inoperable does not increase the reliability of the other channel. This change is consistent with the STS and acknowledges that an inoperable subsystem is not automatically indicative of a similar condition in the redundant subsystem unless a generic failure is suspected; and, that the periodic frequencies  !

specified to demonstrate operability have been shown to be adequate to ensure  ;

equipment operability. This change is justified for the following: it allows credit to be l taken for normal periodic Surveillance as a sufficient demonstration of operability; it recognizes that functional capability is maintained by the remaining components and the reduction in redundancy will be of limited duration; it reduces unnecessary challenges and wear to redundant components; it recognizes that testing of redundant channels does not make those channels more reliable; and, it incorporates operating experience ,

and analysis that have demonstrated the normal surveillance test interval is sufficient to y provide a very high degree of reliability of an instrument channel. ITS Table 3.3.2.1-1 lists the periodic SRs to demonstrate the operability of the RBM channels. Therefore, the deletion of the instrument functional test requirement in CTS 4.3.B.3.b.5 is acceptable.

l L.6 The proposed change would add a note from the STS to the 6-montn Channel

l. calibration surveillance requirement in CTS Table 4.2.C for the RBM upscale and downscale functions (ITS SR 3.3.2.1.5) and the RBM power range function (ITS SR 3.3.2.1.4). The note is less restrictive than the CTS table note 5 by excluding the neutron detectors from these surveillance. The channel calibration is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy. The neutron detectors Cooper Nuclear Station Draft Safety Evaluation

. are excluded from the channel calibrations because they are passive devices with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performance of the 7-day calorimetric calibration (ITS SR 3.3.1.1.2) and the 1000 MWD /T LPRM calibration against the TIPS (ITS SR 3.3.1.1.8). Therefore, the addition of this note to SRs 3.3.2.1.4 and 3.3.2.1.5 is acceptable.

ITS Specification 3.3.3.1 L.9 The requirement in CTS 1.0.J for the definition of LCO that precludes mode changes if a post accident monitor (PAM) instrument is inoperable is proposed to be deleted. A note from the STS will be added to the actions for ITS LCO 3.3.3.1 stating that ITS LCO 3.0.4 is not applicable. This note is less restrictive because it will exclude this ITS LCO from the mode change restrictions of ITS LCO 3.0.4. This ITS LCO 3.0.4 exception allows entry into an applicablo mode while relying on the Actions even though the Actions may eventually require a station shutdown. This exception is acceptable due to the passive function of the PAM instruments, the operators ability to diagnose an accident using attemative instruments, and the low probability of an event requiring this instrumentation.

ITS Specification 3.3.3.2 L.3 The requirement in CTS 1.0.J for the definition of LCO precludes mode changes if an attemate shutdown system (ASDS) instrument or control is inoperable is proposed to be deleted. A note from the STS will be added to the actions of ITS LCO 3.3.3.2 stating that LCO 3.0.4 is not applicable. This note is less restrictive because it will exclude this ITS LCO from the mode change restriction of ITS LCO 3.0.4. This ITS LCO 3.0.4 exception allows entry into an applicable mode while relying on the actions even though the actions may eventually require a station shutdown. This exception is acceptable due to the passive function of the ASDS instruments, the operators ability to diagnose an accident using alternative instruments, and the low probability of an event requiring this system.

ITS Specification 3.3.5.1 L.6 The requirements for high pressure coolant injection (HPCI) turbine stop valve monitor and suppression chamber HPCI suction valve 23-58 instrumentation in CTS Table 3.2.B, including actions, and CTS Table 4.2.B. are proposed to be deleted. These requirements do not necessarily relate directly to system operability. The STS do not specify indication-only equipment to be operable to support operability of a system or component. Control of the availability and necessary compensatory activities for indication instruments, monitoring instruments, and alarms are addressed by station operational procedures and policies. The functions these instruments monitor are all verified during SRs required by ITS LCO 3.5.1 on the emergency core cooling and reactor core isolation cooling (RCIC) systems. Therefore, if the HPCI turbine stop valve Cooper Nuclear Station Draft Safety Evaluation

, or suppression chamber suction valve was in the improper position, or if the HPCI control oil pressure was low, this would be discovered during the required surveillance.

In addition, valve manipulation is controlled by administrative procedures; therefore, it would be unlikely that the HPCI valves would be mispositioned. Monitoring of the HPCI system lineup and condition, conducted by the operations staff on a routine Dasis, would q

also likely find a valve out of position or control oil pressure low. Therefore, this instrumentation, along with the supporting Surveillance and actions are deleted from the TS.

In addition, the HPCI control oil pressure low function in CTS Table 3.2.B, including actions, and CTS Table 4.2.B also does not necessarily relate directly to HPCI system operability. The HPCI control oil pressure low pressure switch functions to actuate the auxiliary oil pump. The auxiliary oil pump serves as the main HPCI lube oil pump for the HPCI turbine stop and control valves on receipt of an HPCI actuation signal to open those valves to allow steam to the turbine. Once the turbine is up to speed, the auxiliary oil pump acts as back up support for the HPCI turbine-driven oil pump. The HPCI Control Oil Pressure Low pressure switch senses oil pressure at the discharge of the turbine-driven oil pump. The switch closes when the discharge pressure is 220 psig, starting the auxiliary oil pump, and it opens when the discharge pressure is 285 psig, stopping the auxiliary oil pump. This function and those of the auxiliary and turbine-driven oil pumps are tested every 92 days when ITS SR 3.5.1.7 is run to discover if the HPCI control oil pressure was low. In the same way, while monitoring the lineup and condition of the HPCI System on a routine basis, Operations staff would likely find control oil pressure law. If this pressure switch were to fail high such that insufficient oil pressure is sensed to exist (failure of 85-psig trip), then the auxiliary oil pump would start. The result of this would be that the auxiliary oil pump would not be able to overcome the pressure created by the turbine-driven pump (approx.100 psig), and the oil from the auxiliary purnp would recirculate back to the HPCI oil reservoir through the 80-psig relief valve, with no impact on HPCl System Operability. If the pressure switch were to fait low so sufficient oil pressure is sensed to exist (failure of 20-psig pump start), then it would be detected during ITS SR 3.5.1.7 on the flowrate from the HCPI pump. This switch failure would cause a failure of HPCI to start due to the lack of control oil pressure causing a failure of the stop and control valves to open, or, on turbine coast down, the need for the auxiliary lube oil pump to start would not be met. In addition, the definition of operable-operability for the HPCi system includes the requirement for support systems including this switch and the auxiliary oil pump to be Operable. Therefore, the deletion of the HPCI Control oil pressure low function from the ITS is acceptable.

ITS Specification 3.3.6.1 L.1 The allowable and actions, for the reactor vessel dome pressure, when operating the RHR system in the shutdown cooling mode, in CTS 1/2.2.2 on SL are proposed to be deleted from the TS section on SL and the allowable value for the pressure incorporated into ITS LCO 3.3.6.1 (item 6.a in Table 3.3.6.1-1, " Primary Containment isolation Cooper Nuclear Station Draft Safety Evaluation

. Instrumentation"). This is consistent with the STS. The RHR shutdown cooling system is designed with an interlock in the logic for the system isolation valves, that are closed during operation in Modes 1,2, and 3 when reactor pressure is greater than the interlock setpoint, to prevent opening of the RHR valves above a preset pressure setpoint (allowable value) of s 72 psig. This setpoint is selected to assure that pressure integrity of the RHR system is maintained. The CTS 1.2.2 requirement that the pressure be less than the limit when operating the residual heat removal pump is covered by the applicability of the instrumentation in ITS LCO 3.3.6.1. This applicability is Modes 1,2, and 3 when primary containment is required to be operable. In Modes 4 and 5 with the  !

pump operating, the reactor is depressurized and the potential for inadvertent pressurization of the containment is very low. The explicit action to decrease reactor vessel dome pressure to below the setpoint is not necessary to state; returning to within l operating limits is always required and need not be explicitly stated. Additionally, the l context of CTS 2.2.2 is covered by ITS LCO 3.3.6.1 Action F which requires that the  !

affected penetration flow path (s) be isolated. The high pressure interlock is only provided for equipment protection to prevent an intersystem loss of coolant accident (LOCA) and, as such, this function should not be considered a SL on station operation.

Therefore, transferring the RHR shutdown cooling system isolation valves setpoint from the TS section on SL (CTS 1/2.2.2) to the primary coolant isolation instrumentation specification (ITS LCO 3.3.6.1), and deleting the redundant explicit actions to be within limits, is acceptable.

ITS Specification 3.3.6.2 l

L.1 The information in CTS Table 4.1.1 that identifies those portions of the drywell high pressure instrument channel which require functional testing is deleted. In addition, the '

CTS Table 4.1.2 description of the type of test equipment used to perform the Drywell High Pressure calibration is also deleted. These changes are consistent with the STS.

These changes are acceptable because (1) the ITS definitions of channel functional teet and channel calitiration provide the necessary guidance and (2) the level of detail as ;o the type of test equipment to be used is not necessary in the ITS and is adequately addressed by station instrumentation test and maintenance procedures.

l ITS Specification 3.3.8.1 i L.4 The instrument check (i.e., channel check) requirement in CTS Table 4.2.B for the '

emergency buses undervoltage relay functions are proposed to be deleted. This is consistent with the STS. Relays that provide no extemal voltage indication are used to perform these functions. These relays are either in the " tripped" or "not tripped" condition, depending on the sensed voltage relative to the trip setpoint. There is no i read-out indication provided that can be used to compare these instruments to the indications of other similar instruments measuring the same parameter. The channel l check requirement is currently satisfied by verifying each of the relays is "not tripped".

This channel check methodology provides a comparison of the " tripped" and "not

. tripped" status of the relays, but does not provide indication of the overall condition of Cooper Nuclear Station Draft Safety Evaluation l

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the relay over and above that provided by annunciators of a diesel generator (DG) started or a bus tripped. Therefore, the verification of this status on a 12-hour  !

periodicity does not provide information that is not constantly available to the station operations staff through the absence of an annunciator. Therefore, deletion of the l

channel check for the emergency buses undervoltage relay functions is acceptable. l ITS Specification 3.4.1 L.1 The CTS 3.3.F.2.a requirement to isolate an idle recirculation loop electrically by disconnecting the breaker to the recirculation pump motor generator set drive motor is l proposed to be deleted. This is not required by the STS. This requirement is not needed to comply with ITS LCO 3.4.1 on operating recirculation loops, which stipulates I the necessary requirements for single loop operation. This CTS requirement is an operational aid designed to preclude accidentally starting of the recirculation pump in the idle loop, and is not directly related to the ability of the system to perform its safety function. Normal station controls provide adequate assurance that a recirculation pump will not be inadvertently restarted. In addition, ITS SR 3.4.9.4 requires verification that the difference between the reactor coolant temperature in the idle recirculation loop to be started and the reactor pressure vessel (RPV) coolant temperature is within the limits specified in the CNS pressure / temperature limit report (PTLR). This provides added assurance that restarting an idle recirculation pump will be performed in a controlled manner. Therefore, the proposed deletion of CTS 3.3.F.2.a is acceptable.

ITS Specification 3.4.4 L.1 The limit on identified RCS leakage is proposed to be increased from 25 gpm to 30 gpm.

CTS 3.6.C.1 specifies an unidentified RCS leakage limit of 5 gpm and an identified leakage limit of 25 gpm for an implied 30 gpm total RCS leakage limit and the ITS LCO 3.4.4 specifies a 5 gpm unidentified RCS leakage limit (ITS LCO 3.4.4.b) and a total RCS leakage limit of 30 gpm averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (ITS LCO 3.4.4.c). ITS LCO 3.4.4.c is less restrictive, in that identified RCS leakage under CTS 3.6.C.1 is limited to 25 gpm, whereas, under ITS LCO 3.4.4, identified Leakage can be as high as 30 gpm (if the unidentified leakage is zero). No applicable safety analysis assumes either the identified or the total Leakage limit. The limit considers RCS inventory makeup and drywell floor drain capacity. The limit of 30 gpm averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is well within the capacity of the control rod drive system (CRDS) pumps, and is well below the capacity of one drywell equipment drain or floor drain pump, which is used to pump the water out of the collecting sump. Additionally, the existing limits associated with unidentified leakage apply. Therefore, ITS LCO 3.4.4.c is acceptable.

ITS Specification 3.4.5 L.1 The requirements in CTS 3.6.C.2 and 3.6.C.3 are proposed to be reduced by adding a note from the STS stating that LCO 3.0.4 is not applicable to the actions to be taken for Cooper Nuclear Station Draft Safety Evaluation

. the condition of the drywell floor drain sump flow monitoring system being inoperable or i the required drywell atmospheric monitoring system being inoperable. The note would apply to ITS LCO 3.4.5 Action A and Action B. Adding a note that states ITS LCO 3.0.4  ;

is not applicable to these actions means that entry into an applicable mode can be made l even if the LCO is not met and, therefore, entry can be made into Modes 1,2, or 3 with l one of the two required RCS leakage detection instrumentation inoperable of ITS LCO 3.4.5. When this ITS LCO 3.0.4 exception is used, either the drywell floor drain sump flow monitoring system or the required drywell atmospheric monitoring system remains available, and the compensatory actions for the inoperable system (or the requirement that unidentified leakage be quantified in accordance with ITS LCO 3.4.5) will provide adequate indication of RCS leakage. The proposed exception is acceptable because:

(1) a 30 day allowed out of service time for one leakage detection system is acceptable based on industry operating experience; (2) a leakage detection system is still operable; and (3) compensatory measures will still ensure leakage is being quantified. j L.3 The instrument check requirement for the floor drain sump flow instrument in CTS Table 4.2.E is proposed to be deleted. This is not required in the STS. An instrument check will not consistently demonstrate operability because thera are no other instruments to check this instrument against and the reading is not consistently the same. Therefore, simply observing this instrument does not provide the information necessary to determine operability. The channel functional t,st is the best indicator of operability while operating, and this requirement is in ITS SR 3.4.5.2. Therefore, the deletion of CTS Table 4.2.E instrument check requirement for the floor drain sump flow instrument is acceptable.

' L.4 The requirements for the drywel'1 equipment drain sump flow monitoring system functions in CTS 3.6.C.2, Tables 3.2.E and 4.2.E, are proposed to be deleted. This instrument quantities identified RCS leakage and, because the purpose of ITS LCO 3.4.5, *RCS Leakage Detection Instrumentation, is to provide instrumentation requirements for early identification of RCS unidentified leakage, the drywell equipment drain sump monitoring system requirements of CTS 3.6.C.2, Table 3.2.E, and Table 4.2.E are not needed. This is consistent with the ISTS. The drywell equipment drain sump flow monitoring system does not necessarily relate directly to the RCS leakage ,

requirements because there are other means to quantify identified leakage are available (e.g., such as equipment drain sump pump-out times). Control of the availability of, and necessary compensatory activities if not available, for indications and monitoring instruments are addressed by station operational procedures and policies. The requirement to demonstrate RCS leakage is within limits is still maintained in ITS l SR 3.4.4.1. As a result, the requirement for a means to quantify identified leakage is adequately addressed by the requirements of ITS 3.4.5 and associated ITS SR 3.4.4.1.

Therefore, explicit requirements for the drywell equipment drain sump flow monitoring system instrumentation are not needed. Therefore, deletion of the drywell equipment drain sump flow monitoring system requirements in CTS 3.6.C.2, Table 3.2.E and Table 4.2.E is acceptable.

i Cooper Nuclear Station Draft Safety Evaluation I

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. L.5 The drywell air sampling system requirements in CTS 3.6.C.3 Table 3.2.E, and Table 4.2.E are reduced so that only one channel of either drywell atmospheric particulate radioactivity monitoring or drywell atmospheric gaseous radioactivity monitoring is required in ITS LCO 3.4.5 because either monitoring channel can provide the same type of !ndication. ITS LCO 3.4.5 requires a method which can quantify the unidentified RCS leakage and a diverse method providing indication of increased leakage: drywell floor drain sump flow monitoring system and one channel of the drywell atmospheric l particulate or atmospheric gaseous monitoring system. A diverse method to quantify l increased leakage is still provided by the drywell floor drain sump flow monitoring system, and this is the primary method for quantifying leakage. The CTS 3.6.C.3 action requirement to allow the required drywell atmosphere radioactivity monitoring

instruments (i.e., all drywell particulate and gaseous radioactivity monitoring capability) l ' to be inoperable for 30 days is maintained in ITS LCO 3.4.5 Action B.1. Therefore, this change is acceptable.

ITS Specification 3.5.2 l L.1 The requirement in CTS 3.5.F.5 for condensate storage tank (CST) inventory in Mode 5 during an operation with the potential of draining the RPV (OPDRV) is proposed to be l deleted. During removal of a control rod drive (which is an OPDRV) with the station in l Mode 5 with the fuel pool gate open, and water level 2 21 feet above the RPV flange l (the " flooded-up" condition), CTS 3.5.F.5 requires a" Tw pressure ECCS to be operable (except for the suppression chamber water supply to the pumps) and a supporting water j supply from the CST of a 230,000 gallons. In this condition, the CTS allows the suppression chamber to be drained and more than one low pressure ECCS pump to be j aligned to take suction from the CST. The basis for the 230,000 gallon CST volume in CTS 3.5.F.5.c was a value the NRC determined adequate in CNS License Amendment  ;

No.11, dated December 30,1975. This amendment added this requirement for the  !

situation in which the suppression pool is empty for required inspection and a control rod

drive is removed for maintenance.-

Under these same conditions, proposed ITS LCO 3.5.2 is not applicable. Therefore, no ECCS injection or spray subsystems or their required ' suppression chamber or CST water supplies are required to be operable. Therefore, when flooded-up, the ITS will  !

permit OPDRVs regardless of the availability of a suction source to the ECCS pumps, whether from the suppression chamber or the CST. Deletion of these CTS ECCS and l associated water supply system requirements are acceptable because the large volume of water available over the reactor in the flooded-up condition provides sufficient

. inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in the case of an inadvertent reactor vessel draindown. (Deletion of CTS 3/4.5.F.5 is still under review.)

l L.4 The details in CTS 3.5.F.5.d,3.5.F.5.e,3.5.F.5.g,3.5.F.5.h,3.5.F.5.1, and 3.5.F.5.]

relating to how to minimize and control RCS leakage during control rod drive removal activities with the suppression chamber drained are proposed to be deleted. These Cooper Nuclear Station Draft Safety Evaluation 1-l L-

, details are not necessary to ensure that the low pressure ECCS are capable of providing makeup water to the reactor vesselin the event of an inadvertent vessel draindown event. The requirements of ITS LCO 3.5.2, from the STS LCO 3.5.2, are sufficient to ensure this capability is maintained. GENE-187-18-0892, " Loss of Coolant Accident Analysis with 125 Volt DC Power Source Failure for Cooper Nuclear Station,"

demonstrates that only one low pressure ECCS injection / spray subsystem is required, post accident, to maintain adequate reactor vessel level. In the event that required low pressure ECCS are not available, ITS 3.5.2 Actions B.1 and C.1 require suspension of operations with the potential for draining the reactor vessel (which includes the control rod drive removal activities). Because these details are not necessary to ensure the low pressure ECCS are capable of providing makeup water to the reactor vesselin the event of an inadvertent vessel draindown event, and the requirements of ITS LCO 3.5.2 are sufficient to ensure that this capability is maintained, the deletion of these details is acceptable.

L.6 The implicit requirement in CTS 3.10.F to suspend refueling operations if the required low pressure ECCS are inoperable or not aligned to an operable condensate storage tank is proposed to be deleted. The refueling specifications in ITS Section 3.9 provide requirements to ensure safe operation during refueling operations and activities in the rea:: tor vessel other than OPDRV. There are requirements in ITS LCO 3.9.7 and 3.9.8 conceming low pressure ECCS because the low pressure ECCS function provides protection for loss of RPV vessel inventory events. ITS LCO 3.5.2 actions require either restoring the required number of low pressure ECCS subsystems within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (Action C.2) or establishing secondary containment integrity (Actions D.1, D.2, and D.3) if all low pressure ECCS subsystems become unavailable. These actions minimize the potential fission product release in the event of an inadvertent vessel draindown. In addition, if all low pressure ECCS subsystems become unavailable, action must be immediately initiated to suspend OPDRV (ITS 3.5.2 Required Action C.1). This action minimizes the potential for the occurrence of an inadvertent draindown event. Other refue!ing operations (e.g., Core alterations), however, do not initiate inadvertent vessel draindown events nor do they hamper the response of the low pressure ECCS and it is not necessary to suspend refueling operations under these conditions. This is consistent with the STS. Therefore, the deletion of the implicit requirement in CTS 3.10.F to suspend refueling operations if the required low pressure ECCS are inoperable or not aligned to an operable CST is acceptable.

ITS Specification 3.5.3 L.3 The requirement to perform a monthly operability test on the reactor core isolation cooling (RCIC) pump and motor operated valves in CTS 4.5.D.1.b and 4.5.D.1.c is proposed to be deleted. The requirements of these tests to verify operability are encompassed in quarterly pump and valve inservice testing. Performing these tests on a quarterly basis is consistent with the CNS Inservice Testing Program to meet American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requirements. Industry operating experience has shown testing the RCIC Cooper Nuclear Station Draft Safety Evaluation

, components on a quarterly basis is adequate for maintaining operability of the system.

, Deleting the monthly tests also reduces wear and tear on the pumps and valves caused by more frequent testing by approximately two-thirds. This is consistent with the STS.

Because these test requirements to verify operability of the RCIC pump and valves are encompassed in quarterly pump and valve testing of the inservice testing program, the proposed deletion is acceptable .

ITS Specification 3.6.1.3 L.1 The time limit of 90 days per year in CTS 3.7.A.2.b which limits the opening of the primary containment purge valves during inerting, de-inerting and pressure control, that are operational requirements, are proposed to be replaced with specific use criteria for opening these valves. The time limit was not based on any analytical requirement. ITS SR 3.6.1.3.1 Note 1, from the STS, provides limits on when the valves are permitted to be opened to assure appropriate controls on these valves so that the valves are used only for operational requirements and safety benefits (i.e., there is a need to have the valves open). These limits in the ITS will allow the valves to be open for as low as is reasonably achievable (ALARA) or air quality considerations for personnel entry (i.e.,10 CFR Part 20 occupational radiation protection requirement) and for surveillance that require the valves to be open in addition to the CTS limits of inerting, de-inerting, and pressure control. Therefore, use of the purge valves will continue to be minimized and limited to operational and safety related reasons. The CNS operating history indicates that these valves are only opened for the specified reasons and for cumulative periods that are generally less than the current allowed cumulative times. Therefore, the note in ITS SR 3.6.1.3.1 to restrict the use of the purge valves to certain valid purposes, instead of a time limit, and the additional uses of the purge valves allowed by the note from the CTS are acceptable.

ITS Specification 3.6.1.7 L.2 The requirement in CTS 4.7.A.3.a for the instrumentation and the instrument setpoints associated with the suppression chamber-to-reactor building vacuum breakers be checked for proper operation every three months is proposed to be deleted. Consistent with the STS, the requirements of ITS SR 3.6.1.7.3 to ensure that the vacuum breakers are full open at s 0.5 psid is sufficient because the vacuum breakers actuation instrumentation is required to be operable to satisfy the setpoint verification in ITS SR 3.6.1.7.3 for the vacuum breakers and to actuate the air actuated vacuum breakers during the functional test ITS in ITS SR 3.6.1.7.2. If the vacuum breaker actuation instrumentation is inoperable, then the SR cannot be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the actions of ITS LCO 3.6.1.7. As a result, the requirements for the vacuum breaker actuation instrumentation are adequately addressed by the requirements of ITS LCO 3.6.1.7, SR 3.6.1.7.2, and SR 3.6.1.7.3. Therefore, the deletion of the requirement of CTS 4.7.A.3.a discussed above is acceptable.

Cooper Nuclear Station Draft Safety Evaluation

j . ITS Specification 3.6.1.8 l L.5 The requirements in CTS 3.7.A.4.a and 4.7.A.4.c that the drywell-to-suppression I

chamber vacuum breaker position indication system be operable and verified operable once each operating cycle, respectively, is proposed to be deleted. Vacuum breaker position is required to be known to be able to satisfy the operability of the breakers in ITS LCO 3.6.1.8, and ITS SRs 3.6.1.8.1, 3.6.1.8.2, and 3.6.1.8.3, from the STS. If.

position indication is not available and vacuum breaker position can not be determined, then the LCO and SRs cannot be satisfied and the appropriate actions of ITS LCO 3.6.1.8 must be taken for inoperable vacuum breakers. As a result, the requirements for the vacuum breaker position indication are adequately addressed by the requirements of ITS LCO 3.6.1.8 and the associated SRs. Therefore, the deletion of the position f'

indication requirements of CTS 3.7.A.4.a and 4.7.A.4.c is acceptable..

. ITS Specification 3.6.2.1

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L.3 The requirement in CTS 4.7.A.1.c that an extemal visual inspection of the suppression chamber whenever there is indication of relief valve operation, with the local suppression pool temperature reaching 160*F or greater and primary coolant system

l. pressure is greater than 200 psig, is proposed to be deleted. As shown in NEDO-30832, " Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers," dated December 1984, that is applicable to the CNS, it has been demonstrated that there are no undue loads on the suppression pool or its components at elevated temperatures and pressures when relief valves discharge through an approved quencher (or sparger). The USAR states that each relief valve discharge line terminates in a T-quencher. This is an approved quencher. Therefore, the deletion of '

the requirement for an extemal visual inspection of the suppression chamber is acceptable.

ITS Specification 3.6.4.1 i L.1 The requirement in CTS 3.7.C.1.b inat the RCS be vented in order for secondary

! containment to not be required in Mode 4 or 5 is proposed to be deleted. As given in the STS, the applicability of ITS LCO 3.6.4.1 on secondary containment includes Modes 4 and 5, but it does not include this requirement on venting the RCS. Secondary

[ containment operability is required to ensure that fission products entrapped within the l' secondary containment structure will be treated by the standby gas treatment system l (SGTS) prior to discharge to the environment. When the reactor is in Mode 4 or 5, the l probability and consequences of the design basis accident (DBA) requiring secondary containment operability to be maintained are reduced due to the RCS pressure and temperature limitations in these conditions. Therefore, maintaining secondary containment operability is not required in Mode 4 or 5, except for situations for which  ;

significant releases of radioactive material can be postulated, such as during OPDRV, core alterations, or movement of irradiated fuel assemblies in the secondary ,

containment. The reactor in Mode 4 or 5 with the RCS not vented does not constitute a Cooper Nuclear Station Draft Safety Evaluatiun i

. situation for which significant releases of radioactive material can be postulated. The RCS will normally be vented when the reactor is in Mode 4 or 5. With the RCS not vented when the reactor is in Mode 4 (for example, during an inservice leak and hydrostatic test) or Mode 5, no mechanism exists to impart additional fission products into the reactor coolant. Under these conditions, activities for which the RCS would not be vented would be strictly controlled and monitored. As a result, leaks or pipe breaks would typically be detected before significant inventory loss occurred. These activities would typically be performed after refueling when few noncondensible gases remain in the reactor coolant. The temperature limitation of 212*F will ensure that water, not steam, would be emitted from the postulated leak or pipe break. In addition, under these conditions, stored energy is sufficiently low that even with a loss of inventory following a recirculation line break, core coverage would be maintained by the low pressure ECCS required per ITS LCO 3.5.2 and the fuel would not exceed its peak clad temperature limit. As a result, the potential for failed fuel and a subsequent increase in reactor coolant activity is minimized, and significant releases of radioactive material would not be expected to occur. Therefore, the proposed change is acceptable.

L.2 A requirement in CTS 4.7.C.1.c is to perform the secondary containment capability test with the SGTS at each refueling outage prior to refueling. The requirement that the test must be prior to each refueling is proposed to be deleted. As shown in the STS, it is not necessary to state that the test must be performed at a specific time, such as prior to each refueling; it is only necessary to have the test successfully completed in the required surveillance interval. ITS SR 3.6.4.1.4 will require the secondary containment capability test to be performed once per 18 months as is required by CTS 4.7.C.1.c (i.e.,

each refueling outage); but it does not require the surveillance at any specific time.

Therefore, the requirement of CTS 4.7.C.1.c to perform the surveillance prior to refueling is not necessary to ensure the secondary containment is maintained operable, and its deletion is acceptable.

ITS Specification 3.6.4.2 L.3 The requirement in CTS 3.7.C.1.b that the RCS to be vented in order for secondary containment isolation valves to not be required in Mode 4 or 5 is proposed to be deleted.

As given in the STS, the applicability of ITS LCO 3.6.4.2 on secondary containment isolation valves operability includes Modes 4 and 5, but does not include this requirement on venting the RCS. Secondary containment isolation valve operability is required to ensure that fission products entrapped within the secondary containment structure will be treated by the SGTS prior to discharge to the environment. When the reactor is in Mode 4 or 5, the probability and consequences of the DBA requiring secondary containment operability (which includes secondary containment isolation valve operability) to be maintained are reduced due to the RCS pressure and temperature limitations in these conditions. Therefore, maintaining secondary containment isolation valve operability is not required in Mode 4 or 5, except for situations for which significant releases of radioactive material can be postulated, such as during OPDRV, core alterations, or movement of irradiated fuel assemblies in the Cooper Nuclear Station Draft Safety Evaluation

. l l secondary containment. The reactor in Mode 4 or 5 with the RCS not vented does not i constitute a situation for which significant releases of radioactive material can be postulated. The RCS will normally be vented when the reactor is in Mode 4 or 5. With the RCS not vented when the reactor is in Mode 4 (for example, during an inservice leak or hydrostatic test) or Mode 5, no mechanism exists to impart additional fission products into the reactor coolant. Under these conditions, activities for which the RCS would not be vented would be strictly controlled and monitored. As a result, leaks or pipe breaks would typically be detected before significant inventory loss occurred. These activities would typically be performed after refueling when few noncondensible gases remain in the reactor coolant. The temperature limitation of 212F will ensure that water not steam would be emitted from the postulated leak or pipe break. In addition, under these conditions, stored energy is sufficiently low that even with a loss of inventory following a recirculation line break, core coverage would be maintained by the low pressure ECCS required per ITS LCO 3.5.2 and the fuel would not exceed its peak clad temperature limit. As a result, the potential for failed fuel and a subsequent increase in reactor coolant activity is minimized, and significant releases of radioactive material would not be expected to occur. Therefore, the proposed change is acceptable.

ITS Specification 3.6.4.3 i

L.4 The requirement in CTS 4.7.B.4.6 that the manual operability of the bypass valves for_

SGTS filter cooling be demonstrated once per operating cycle is proposed to be deleted.

This is not a requirement in the STS. The intent of this requirement is to verify that the ventilation mode of SGTS operation is available for removal of decay heat from the SGTS filters. However, at CNS, the bypass valves do not perform this function and are maintained in the closed position, except to perform CTS 4.7.B.4.b. These bypass valves serve no purpose and are not taken credit for in any accident of transient, and the ability of these valves to open, as demonstrated by CTS 4.6.B.4.b, is not required for SGTS operability. The SGTS filter cooling function is performed by the SGTS cross tie damper, room air supply check valves, and dilution air shutoff valves. The requirement to verify, once per 18 months (an operating cycle), the correct position of the cross tie damper and the ability of each valve to be opened is in ITS SR 3.6.4.3.4. Therefore, the requirement to demonstrate the manual operability of the bypass valves for SGTS filter cooling is unnecessary, and the deletion of this requirement is acceptable.

ITS Specification 3.8.4 .

L.2 The surveillance frequency requirement in CTS 4.9.A.3.b, that within 7 days after (1) a I battery discharge causing terminal voltage to drop below a specified value or (2) an overcharge causing terminal voltage to rise above a specified value, is proposed to be deleted in verifying that there is no abnormal corrosion at either battery terminals or 4 connectors, and the connection resistance is less than the required limits. This frequency is not required in the STS, and is omitted from ITS SR 3.8.4.2, on these verifications. ITS SR 3.8.4.2 retains the 3-month frequency also in CTS 4.9.A.3.b.

l_ Experience has demonstrated that corrosion rates and connection resistance are not l Cooper Nuclear Station Draft Safety Evaluation i

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immediately arid significantly affected by a severe discharge or overcharge condition; therefore, the 3-month frequency is adequate for ensuring that the battery terminal connections or intercell connections resistance has not degraded significantly.

Therefore, the proposed change is acceptable.

The licensee addressed the basis for deleting the 7-day interval for the verification of the average battery cell electrolyte temperatures in the less restrictive change L.1 for ITS 3.8.6 immediately below.

ITS Specification 3.8.6 L.1 The surveillance frequency requirement in CTS 4.9.A.3.b, that within 7 days after (1) a battery discharge causing terminal voltage to drop below a specified value or (2) an i overchar0e causing terminal voltage to rise above a specified value, is proposed to be deleted in verifying the average battery cell electrolyte temperature is > 70 degrees F.

This frequency is not required in the STS and is omitted from ITS SR 3.8.6.3 on verifying this temperature. ITS SR 3.8.6.3 on the average temperature of representative cells retains the 3-month frequency also in CTS 4.9.A.3.b. The proposed change is acceptable because severe discharging and overcharging generally increases the battery electrolyte temperature.

i L.2 The STS Table 3.8.6-1 footnote (a) has been proposed to be added to the electrolyte level Category A and B limits of CTS Table 3.9.1. This footnote (a) in ITS Table 3.8.6-1 allows for a temporary electrolyte level increase during and following an equalize charge of the batteries. This footnote is based on guidance from Appendix A to IEEE-450, 1987. The level excursion is due to gas generation during the equalize charge and would be expected to return to normal within 3 days following completion of the equalize charge, although it may take several days longer. Therefore. the addition of this note is acceptable.

L.3 The allowance of CTS Table 3.9.1, footnote (6), for utilizing charging current in lieu of battery cell specific gravity is proposed to be applied to all the category limits in CTS Table 3.9.1 for specific gravity. Currently, the footnote is not applied to the specific gravity Category B (quarterly) limits for each connected cell and allowable values below the average of all connected cells. The application of footnote (6)is consistent with STS Table 3.8.6-1 and is included in ITS Table 3.8.6-1 for all the category limits and allowable values for battery cell specific gravity. Application of this footnote to the additional battery cell specific gravity limits and allowable values not in CTS Table 3.9.1 is acceptable because all connected cells are affected by a battery charge, and charging current is a more accurate indication of battery state following a charge than is specific gravity.

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. ITS Specification 3.9.1 L.1 The requirement in CTS 4.10.A.1 for post maintenance testing of refueling equipment interlocks is proposed to be deleted. This surveillance requirement is not stated in the STS. Any time the operability of a system or component has been affected by repair, t' maintenance, or replacement of a component, post maintenance testing is required to demonstrate operability of the system or component. After restoration of a component that caused a SR to be failed, ITS SR 3.0.1 will require performing the appropriate SR (in this case, SR 3.9.1.1 for refueling equipment interlocks) to demonstrate operability of the affected components. Therefore, the explicit post maintenance surveillance in CTS 4.10.A.1 is not required, and the proposed change is acceptable.

ITS Specification 3.9.2 L.3 The requirement in CTS 4.10.A.1 for post maintenance testing of the refuel position one-rod-out interhek is proposed to be deleted. This surveillance requirement is not stated in the STS. Any time the operability of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate operability of the system or component. After restoration of a component that caused a SR to be failed, ITS SR 3.0.1 will require performing the appropriate SR (in this case, SR 3.9.2.2 for the refuel position one-rod-out interlock) to demonstrate operability of the affected components. Therefore, the explicit post maintenance surveillance in CTS 4.10.A.1 is not required, and the proposed change is acceptable.

ITS Specification 3.10.4 L.1 The requirement in the CTS not to withdraw control rods from the core while the station is in Mode 4 is proposed to be reduced to allow the withdrawal of one control rod. ITS 3.10.4 on single control rod withdrawal is a specification from the STS that is proposed to be added to the ITS to allow the reactor mode switch to be placed in the refuel position and a single control rod to be withdrawn, and its associated control rod drive removed, while in Mode 4. This would be allowed provided certain Mode 5 requirements are met. ITS LCO 3.10.4 contains requirements to ensure that the one-rod-out interlock is operable or a control rod withdrawal block is inserted so that (1) only one rod is withdrawn, (2) all other control rods are fully inserted, and (3) either the Mode 5 operability requirements are met for the RPS, the RPS electric power monitoring system, and the control rod to be withdrawn, or all other rods in a 5 by 5 array, centered on the control rod to be withdrawn, are disarmed, allowing a modification to the way in which the SDM requirement is met (i.e., the withdrawn control rod may be assumed to have the highest worth). These requirements effectively compensate for the reactor mode switch not being in the shutdown position with a rod withdrawn with the station in Mode 4. ITS 3.10.4 imposes the same types of requirements on the station as if the station were in Mode 5. These requirements, coupled with SDM requirements that assume the most reactive rod is fully withdrawn, are adequate to prevent an inadvertent Cooper Nuclear Station Draft Safety Evaluation

criticality when a single control rod is withdrawn for maintenance or testing. For these reasons, the proposed ITS 3.10.4 is acceptable.

ITS Specification 3.10.5 L.2 The requirement in CTS 3.10.A.S.a that allows the simultaneous withdrawal of two nonadjacent control rods for maintenance while shutdown with the reactor mode switch in the refuel position, is proposed to be replaced with the less restrictive requirements of ITS LCO 3.10.5.b, LCO 3.10.5.c (first part), and LCO 3.10.5.d. These proposed requirements are from the STS. Although the proposed requirements will allow the withdrawal of only one control rod and the subsequent removal of one control rod drive, this change is less restrictive because ITS LCO 3.10.5 allows a single control rod to be withdrawn and the control rod drive (CRD) to be removed under a wider range of station conditions; it also allows greater flexibility for establishing those conditions that will prevent inadvertent criticality than are currently available in CTS 3.10.A.S. With the reactor mode switch in the refuel position, the analyses for control rod withdrawal during refueling are applicable and will bound the consequences of an accident. The safety analyses demonstrate that the functioning of the refueling interlocks together with adequate SDM will preclude unacceptable reactivity excursions. While the refueling interlocks are allowed to be bypassed to remove the CRD, the requirements of ITS LCO 3.10.5.a and LCO 3.10.5.b will preclude any other control rod from being witharawn.

LCO 3.10.5.a requires maintaining all other control rods inserted, and LCO 3.10.5.b requires disarming those control rods in a five by five array around the withdrawn control rod. These restrictions satisfy the function of the refueling interlocks . Also, LCO 3.10.5.d prohibits making other core alterations during removal of a single CRD, and LCO 3.10.5.c requires inserting a control rod block. These restrictions will further ensure that no additional control rods will be withdrawn. In addition, appropriate action requirements, ITS 3.10.5 Actions A.1, A.2, and A.3 are provided to require restoration of requirements that were suspended in order to withdraw a control rod and subsequently remove the associated control rod drive in the event any of the requirements of ITS LCO 3.10.5 are not met. Finally, SRs are provided to ensure the requirements of the LCO are satWied. These requirements, coupled with SDM requirements for the most reactW and fully withdrawn, are adequate to prevent inadvertent criticality when a single CRD 4 Xnoved for maintenance or testing. Therefore, the relaxations contained in proposeu ITS 3.10.5 are acceptable.

ITS Specification 3.10.7 L.1 It has been proposed to suspend the requirement of CTS 4.3.B.3.b and the new specification ITS LCO 3.1.6 that the withdrawal operable control rods shall be in compliance with the banked position withdrawal sequence (BPWS). The proposed STS LCO 3.10.7 is a new specification to be added to the ITS to allow suspension of the requirements of ITS LCO 3.1.6 during performance of SDM testing, control rod scram time testing, and control rod friction festing, provided either (1) the BPW3 requirements of ITS SR 3.3.2.1.8 are changed to require the control rod sequence to conform to the Cooper Nuclear Station Draft Safety Evaluation

specified test sequence; or (2) the rod worth minimizer (RWM) is bypassed, the

, requirements of ITS 3.3.2.1, Function 2 are suspended, and conformance to the l approved control rod sequence for the specified test is verified by a second licensed operator or other qualified member of the technical staff. These two requirements limit the potential amount and rate of reactivity increase that could occur during a control rod i drop accident (CRDA). STS LCO 3.10.7 is necessary because with the station in Mode l 1 or 2, control rod testing is sometimes required which may result in control rod pattems J not in compliance with the prescribed sequences of ITS LCO 3.1.6. The adoption of ITS LCO 3.10.7, which is based on STS 3.10.7, is a less restrictive change because it provides flexibility not allowed by the CTS to perform certain operations by appropriately modifying requirements of other LCOs. The allowance to perform testing under this new specification is based on a reanalysis of the CRDA to demonstrate that the special sequence to be used will not result in unacceptable consequences, should a CRDA occur during the testing. These analyses must be performed in accordance with an NRC-approved methodology, and are dependent on the specific test being performed j Further, the BPWS requirements must be changed to be consistent with the revised CRDA analyses unless the RWM is bypassed, ITS LCO 3.3.2.1, Function 2 is suspended, and conformance to the new rod control pattem is verified by an authorized individual. Therefore, the proposed ITS LCO 3.10.7 is acceptable.

ITS Specification 3.10.8 L.2 The surveillance frequency requirement in CTG 4.22.A.1.b that the RWM be verified ,

operable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of a SDM demonstration is l proposed to be deleted. This surveillance is essentially a " paper check" of the  ;

applicable RWM SRs to ensure that they are current. It does not require the actual l performance of these surveillance. ITS SR 3.10.8.2 does not include this paper check, b t it maintains the requirement to perform the surveillance specified for the rod RWM in ITS LCO 3.3.2.1, Function 2 of Table 3.3.2.1-1, at the frequency specified therein.

In particular, ITS SRs 3.3.2.1.2 and 3.3.2.1.3 require performing a channel functional test (CFT) of the RWM with a frequency of 92 days. Both of these SRs contain notes that modify this frequency. The note in SR 3.3.2.1.2 allows delaying performance of the CFT for at least one hour after any control rod is withdrawn at s 10% rated thermal power (RTP) in Mode 2. The note in SR 3.3.2.1.3 allows delaying performance of the l CFT for at least one hour at s 10% RTP in Mode 1. Performance of these SRs at the specified frequency ensures the operability of the RWM is adequately verified prior to and during SDM. The deletion of the paper check, which is an administrative exercise appropriately addressed by station procedures, is acceptable.

ITS Section 4.0 l

L.1 The requirement of CTS 5.2.C to provide a report to the NRC concoming lead test assemblies (LTAs) and analyses within 30 days prior to startup is proposed to be - 3 deleted. This requirement is not in the STS. This report does nothing to ensure station  !'

operation in a safe manner and there is no requirement for the NRC to approve the Cooper Nuclear Station Draft Safety Evaluation l

report as a condition for station startup. In addition, CTS 5.2.C states that LTAs other than described in CTS 5.2.A and 5.2.8 may be installed under the provisions of 10 CFR 50.59 in conjunction with vendor test programs. This statement is also deleted because the applicability of the evaluation requirements of 10 CFR 50.59 clearly apply to such changes in the reactor core and does not need to be repeated in the ITS. However, the requirement of CTS 5.2.C for LTAs to be analyzed using methods previously approved by the NRC is retained in ITS 4.2.1 with clarification; it reads, " Fuel assemblies shall be limited to those fuel designs that have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions." This clarified language in ITS 4.2.1, in addition to the language specifying the number of fuel assemblies, and the composition of the fuel rods, is adequate to ensure the reactor core design is bounded by the assumptions of station accident analyses. Therefore, the reporting requirement and the redundant 10 CFR 50.59 evaluation requirement in CTS 5.2.C are not needed to ensure public health and safety, and the proposed deletion of these requirements is acceptable.

L.2 The requirement in CTS 5.2.A that each fuel assembly consist of zircaloy-clad fuel rods is proposed to be expanded to also include ZlRLO clad fuel rods as acceptable for fuel assemblies in the core. ZlRLO clad fuelis allowed in the STS. The proposed ITS 4.2.1 will allow the use of either zircaloy or ZlRLO clad fuel rods. The allowance to use either zircatoy or ZlRLO clad fuel rods has been generically approved by the NRC.10 CFR 50.44,

• Standards for combustible gas control system in light-water-cooled power reactors," and 10 CFR 50.46, " Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," allow the use of either zircaloy or ZlRLO clad fuel rods as acceptable fuel designs. The Statements of Consideration for changes to 10 CFR 50.44 and 10 CFR 50.46, published in 57 FR 39353 dated August 3,1992, state that including ZlRLO as an acceptable zirconium based fuel cladding material will not reduce the protection of the public health and safety. In addition, prior to use of ZlRLO fuel clad, ITS 4.2.1 will require that the licensee for the CNS analyze the fuel design using NRC-approved codes and methods and also ensure that the fuel design complies with all safety design bases. Therefore, the proposed change has no impact on station safety, and is acceptable.

ITS Section 5.2 L.1 The requirement in CTS 6.1.3.E that an individual trained and qualified in health physics shall be on site at all times that fuel is on site is proposed to be reduced to the requirement that this individual to be on site at all times only when fuel is in the reactor.

The new requirement is the proposed ITS 5.2.2.d that is from the STS. During periods when there is no fuelin the reactor, the probability of an accident that could cause a radioactive release is low. Also, the CNS radiation protection procedures provide sufficient guidance and controls for radiation protection coverage for the occasions when fuelis not in the reactor and are sufficient to maintain station personnel exposure to radiation and radioactive materials to ALARA values in accordance with the Cooper Nuclear Station Draft Safety Evaluation

, requirements of 10 CFR Part 20. Therefore, this relaxation of the health physics staffing requirement is acceptable.

ITS Section 5.5 L.1 The proposed change will apply SR 3.0.2 (allowing an extension of 1.25 times the l surveillance interval) and SR 3.0.3 (allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform the surveillance if it is missed) to the IST test frequencies. This is consistent with the requirements on the station IST program in the STS. The regulation 10 CFR 50.55a(g) requires inservice testing (IST) of the ASME Code Class 1,2, and 3 pumps and valves at the CNS. NRC Generic Letter (GL) 89-04 states that, if these pumps are within the action range of an LCO or the valves exceed the limiting full stroke 1,'me value, the associated component must be declared inoperable and the applicable LCO actions entered. The interval extensions of SRs 3.0.2 and 3.0.3 are not applied to the IST program in the CTS; however, the proposed IST program requirements in ITS 5.5.6 are consistent with this guidance. The change also adds a requirement that the ASME Code requirements will not supersede the requirements of any TS. The use of the surveillance interval extension allows for flexibility in scheduling and consideration of station conditions. The use of the 24-hour delay period should allow adequate time to complete the surveillance that has been missed. The basis for this time delay includes consideration of unit conditions, the time required to perform the surveillance, the safety significance of the delay in completing the required surveillance, and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the TS requirements. Therefore, the proposed application of the provisions of ITS SRs 3.0.2 and 3.0.3 to the IST program test frequencies is acceptable.

ITS Section 5.6 L.1 The proposed change to CTS 6.5.1.E.1 is to delay submitting the annual radiological environment report (AERP) by two weeks. CTS 6.5.1.E.1 requires the ARER to be submitted by May 1 of each year. The proposed ITS 5.6.2, as stated in STS 5.6.2, would allow the report to be submitted by May 15. This schedule change is not significant because the report is not on current station operation, but it covers the surveillance activities on radiologicalimpacts to the environment related to station operation in the previous calender year. The submittal of the report is clearly not necessary to assure station operation in a safe manner for the brief period from May 1 to May 15. Therefore, this schedule change has no impact on the safe operation of the station and is acceptable.

Table L of the less restrictive changes lists all CTS requirements that have been relaxed and which pertain to Category I though Vil and to the specific listing of changes discussed above.

Table L also list those less restrictive changes that are discussed individually in Section Ill.G below.

Cooper Nuclear Station Draft Safety Evaluation

. 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 Less Restrictive CTS Requirements (Not Entire Specifications) i When requirements in the TS have been shown to give little or no safety oenefit, their removal  !

from the TS may be appropriate. This section discusses the relocation of details within the CTS l l to licensee-controlled documents, instead of the relocation of entire specifications from the CTS to licensee-controlled documents which is discussed below in Section Ill.E. In most cases, l 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 operating experience, or (3) resolution of the Owners Groups comments on the STS (the TSTF process). The NRC staff reviewed generic relaxations contained in the STS l and found them acceptable because they are consistent with current licensing practices and the ,

l Commission's regulations. The CNS design was also reviewed to determine if the specific j l design basis and licensing basis of the CNS were consistent with the technical basis for the l l model requirements in the STS, and thus provide a basis for the proposed ITS. A significant l l number of changes to the CTS involved the removal of specific requirements and detailed l information from individual specifications evaluated to be Types 1 through 4 that follow:

[ 1 Type 1 Details of System Design l I Type 2 Descriptions of System Operation l Type 3 Procedural Details for Meeting TS Requirements Type 4 Administrative Requirements Redundant to Regulations

. The following discussions address why each of the four types of inforniation or specific requirements are not required to be included in ITS.

Details of Svstem 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 NRC-approved QA plan (referenced in the USAR). In 10 CFR 50.5g 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 50.5g l 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 f

Cooper Nuclear Station Draft Safety Evaluation

)

USAR, controlled design documents and drawings, or the TS Bases, as appropriate.

l 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.

]

J Descriptions of Svstem Ooeration (Type 2)

The plans for the normal and emergency operation of the facility are required to be l described in the USAR by 10 CFR 50.34. Controls specified in 10 CFR 50.59 apply to t

changes in procedures as described in the USAR. Controls specified in 10 CFR 50.59 ,

apply to changes in procedures as described in the USAR. Controls specified in 10 1 CFR 50.54(a) apply to changes to the QA Program. The ITS Bases also contain descriptions of system operation and controls specified in ITS 5.5.10 apply to changes to the Bases, which includes changes to Bases descriptions of system operation. It is l acceptable to remove details of system operation from the TS because this type of

- information will be adequately controlled in the USAR, QA program, station operating procedures described in the USAR, and the ITS Bases, as appropriate.

Procedural Details for Meetina TS Requirements (Type 3)

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 assessment manual (ODAM), 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.

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 j in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," allowing this procedural control is consistent with the vast majority of other SRs that do not dictate station conditions for surveillance. Prescriptive procedural information in an acuon requirement is unlikely to contain all procedural considerations .

necessary for the station operators to complete the actions required, and referral to l 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 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 l

Cooper Nuclear Station Draft Safety Evaluation l

. Requirements," 10 CFR 50.36 and 10 CFR 50.73 adequately cover the reports deemed to be necessary.

Some procedural details contained in the CTS are of such minor significance that including them in licensee-controlled documents noted above is not appropriate. In such cases, this information is regarded as deleted rather than relocated from TS. It will, however, continue to be contained in appropriate station procedures required by ITS 5.4.1.

Administrative Requirements Redundant to Regulations (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 and 10 CFR 50.36 allow licensees to voluntarily use the criteria in 10 CFR 50.36 to relocate to licensee-controlled documents CTS requirements that do not meet any of the criteria. Changes to the facility or to procedures as described in the USAR are made in accordance with 10 CFR 50.59. Changes made in accordance with the provisions of other licensee-controlled documents are subject to the specific requirements of those documents. For example,10 CFR 50.54(a) govems changes to the QA plan, and ITS 5.5.1 govems changes to the ODAM. Therefore, relocation of the administrative details identified above, is acceptable.

Table RL lists the requirements and detailed information in the CTS that is relocated to licensee-controlled documents and not retained in the ITS. Table RL is organized ir) ITS order by each LA-type DOC to the CTS. It includes: the ITS section or specification designation, as appropriate, followed by the DOC identifier; e.g.,3.1.1 followed by LA.1 means ITS Specification 3.1.1, DOC LA.1); CTS reference; a summary description of the relocated details; the name of the document to contain the relocated details or requirements (new location); ine method for controlling future changes to relocated requirements (control process); a characterization of the change; and a reference to the specific change type, as discussed above, for not including the information or specific requirements in ITS.

The NRC staff has concluded that these types of detailed information and specific requirements are not necessary 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 governed by a regulatory or TS requirement

. TS Bases controlled in accordance with 10 CFR 50.59, as stated in ITS 5.5.14, l " Technical Specifications Bases Control Program.'

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

. ODAM controlled by ITS 5.5.1

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

Cooper Nuclear Station Draft Safety Evaluation

For each of these changes, Table RL of details relocated from the CTS also lists the licensee-controlled documents and the TS or regulatory requirements goveming changes to those documents.

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 in the Final Policy Statement (discussed in Section ll 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 may be relocated from existing TS (an NRC-controlled document) to appropriate licensee-controlled documents. This section discusses the relocation of entire specifications in the CTS to licensee-controlled documents, instead of relocating details from specifications in the CTS to licensee-controlled documents I which is discussed in the previous Section Ill.D. These specifications include the LCOs, Action Statements (i.e., Actions), and associated SRs. In its application and its supplements, the licensee proposed relocating such specifications from the CTS to the USAR (which includes the TRM by reference), and the ODAM, as appropriate. The staff has reviewed the licensee's submittals, and finds that relocation of these requirements to the USAR, TRM, and ODAM is acceptable, in that changes to the USAR and TRM will be adequately controlled by 10 CFR 50.59 and changes to the ODAM will be controlled by ITS 5.5.1. These provisions will continue to be 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 certcin specifications from the CTS and place them in licensee-controlled documents noted in Table R of relocated current technical specifications. Table R lists all specifications that are relocated from the CTS based on the Final Policy Statement and 10 CFR 50.36, 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 ITS, to the extent possible, followed by a reference to the associated relocated CTS requirements; a summary description of the relocated CTS requirements; the name of the document that will contain the relocated requirements (new location); and the method for controlling future changes to the relocated requirements (control process). The NRC staffs evaluation of each relocated specification presented in Table R is provided below, mostly in CTS order with the corresponding DOC

. identifier is given in parenthesis after the title of each relocated specification. In a few cases, the staff determined that relocation of requirements designated as LA-type changes were more Cooper Nuclear Station Draft Safety Evaluation

, appropriately designated R-type changes. Therefore, Table R and the following evaluations contain some relocations that the licensee had designated as LA-type changes.

The evaluations given below on relocating whole specifications from the CTS to licensee-controlled documents are not in the same order as the list of relocations in Table R.

1.

3/4 2.A.4. Main Steam Line I ask Detection (Turbine Bniana) (3.3.6.1 - R.3)

Table 3.2.A Instrument Channel 4; Table 4.2.A Instrument Channel 5 The requirements in CTS Table 4.2.A on the turbine building area temperature switches instruments which are included in the main steam line (MSL) leak detection system (i.e.,

item identification (l.D.) no. MS-TS-143/150 of Table 4.2.A) are proposed to be relocated to the_TRM. The steam tunnel temperature switches wili be retained in the CTS and included in the ITS because they are taken credit for in the main steam line break (MSLB) DBA. The MSLB is a double-ended guillotine break and, for this size break, flow is limited by the main steam line flow restrictors, and the main steam line high flow instruments are credited to terminate the event by closing the MSIVs.

However, the turbine building area temperature switches are used to detect a main steam line leak in the turbine building heater bay of a magnitude of from 1-10% rated flow and, for small breaks of this size, no credit is taken for automatic isolation of the MSIVs by these area temperature switches. As discussed in the USAR, operator action is credited for terminating leaks of this magnitude, because the offsite dose consequences of this type of event have been calculated to be orders of magnitude -

below those of the MSLB because virtually no water carryover (and resulting iodine) is expected to occur. Thus, the turbine building area temperature switches are not assumed to be operable to mitigate the consequences of a DBA or transient, and are not an input assumption for any DBA analysis. Therefore, CTS 3/4.2.A.4 for these are temperature switches does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the MSL turbine building area temperature switches after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the MSL turbine building area temperature switches to the TRM is acceptable.

2. 3/4.2.B.7. Core Sorav System (CSS) Pumo Discharae Line Low Pressure (3.3.5.1 - R.1)

Table 3.2.B Instrument 7; Table 4.2.B, CS Instrument 7 l l This instrumentation provides indication that the CSS keep fill system may be inoperable  ;

and not maintaining the discharge line pressurized. This could allow the discharge line I l- to drain down, resulting in the CSS being unable to perform its design function.

However, this indication-only instrumentation is not assumed to function in any DBA.

The CS pump discharge line low pressure instrumentation does not necessarily relate directly to the CSS operability. The CTS require that the CSS discharge line be maintained and filled. Control of the availability of, and necessary compensatory l I

. Cooper Nuclear Station Draft Safety Evaluation j

. activities, if not available, for indications and monitoring instruments, are addressed by station operational procedures and policies. The requirement to maintain the CSS discharge line filled is retained as ITS SR 3.5.1.1 and is sufficient to sasure the operability of the CSS flow path. Therefore, CTS 3/4.2.B.7 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the CS pump discharge line low pressure instrumentation after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the CSS pump discharge line low pressure instrumentation to the TRM is acceptable.

3. 3/4.2.B.13. Drvwell Pressure Containment Sorav (3.3.5.1 - R.2)

Table 3.2.B Instrument 13; Table 4.2.B RHR Instrument 6 The containment spray mode is manually initiated, if needed, after the low pressure coolant injection (LPCl) cooling requirements are satisfied. The drywell pressure containment spray interlock prevents the operator from inadvertently initiating containment spray when not required to reduce drywell pressure during a LOCA. It also functions to preclude inadvertent containment spray initiation at other times (i.e., non-LOCA conditions). However, other features are installed that prevent containment l failure due to negative pressure and safety analyses assume these features function during an inadvertent containment spray initiation event. This instrumentation ;s not used to mitigate a DBA or transient. Therefore, CTS 3/4.2.B.13 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the drywell pressure containment spray interlock after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.5g, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the drywell pressure containment spray interlock to the TRM is acceptable.

4. 3/4.2.B.17. RHR Outboard Isolation Valve Interlock (3.3.5.1 - R.3)

Table 3.2.B Instrument 17; Table 4.2.B RHR Instrument 11 1 The purpose of RHR outboard isolation valve inteiuk instruments is to interlock open the RHR outboard isolation valves upon receipt of a LOCA signal, to ensure maximum ,

LPCI flow to the reactor vessel. This ensures that a loss of LPCI flow will not occur due '

to an operator inadvertently closing these valves. However, while this feature may  !

provide added assurance of LPCI flow under certain circumstances, it is not assumed in l any design basis analysis. In addition, under certain conditions the operator must secure LPCI flow, and thus must do so by other means that are not interlocked (e.g.,

secure the RHR pumps). Therefore, CTS 3/4.2.B.17 does not meet any of the criteria in  ;

10 CFR 50.36 and may be relocated out of the CTS. Any changes to these  :

requirements regarding the RHR outboard isolation valve interiock after they are  !

relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued Cooper Nuclear Station Draft Safety Evaluation

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, protection of the public health and safety, and the relocation of the CTS requirements for the RHR outboard isolation valve interlock to the TRM is acceptable.

5. 3/4 2.B.19. RHR Heat Exchanoer Bvn=== Time Delav (3.3.5.1 - R.4)

Table 3.2.B instrument 19; Table 4.2.8 RHR instrument 13 The RHR haat exchanger bypass time delay instrumentation in CTS Table 3.2.B is  ;

designed to preclude the heat exchanger bypass valve from being manually closed during a LOCA until a specified amount of time has passed. This ensures that the maximum amount of RHR flow reaches the reactor pressure vessel during a LOCA.

However, while this instrumentation provides added assurance of LPCI flow under certain conditions, it is not assumed to mitigate a DBA or transient. In addition, there are many other instances where the operator must reduce or secure LPCI flow, and must do so by other means that are not interlocked (e.g., secure the RHR pump).

Therefore, CTS 3/4.2.B.19 does not meet any of the criteria in 10 CFR 50.36 and may I be relocated out of the CTS. Any changes to these requirements regarding the RHR heat exchanger bypass time delay after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the RHR heat exchanger bypass time i delay to the TRM is acceptable.

6. 3/4.2.B.20. RHR Crosstie Valve Position (3.3.5.1 - R.5) l Table 3.2.B instrument 20; Table 4.2.B RHR Instrument 14 This instrument initiates annunciation in the control room when the LPCI cross-tie valve is not closed. During normal operation, the LPCI cross-tie valve is required to be closed. Therefore, this instrument is not the primary method for ensuring the valve remains closed, nor does any accident analysis take credit for this instrument. This instrument is not assumed to be operable to mitigate the consequences of a DBA or transient, and is not an input assumption for any DBA analysis. Therefore, CTS 3/4.2.B.20 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the RHR crosstie valve position instrumentation after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient

. regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the RHR crosstie valve position l instrumentation to the TRM is acceptable.

7. 3/4.2.B.23. RHR Pumo Discharoe Line (3.3.5.1 - R.8)

Table 3.2.B Instrument 23; Table 4.2.B RHR Instrument 17 This instrumentation provides iridication that the RHR keep fill system may be inoperable and may not be maintaining the discharge line pressurized. This could allow the discharge line to drain down, resulting in the RHR system (e.g., LPCl) being unable l

Cooper Nuclear Station Draft Safety Evaluation l

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i to perform its design function. However, this indication-only instrumentation is not assumed to function in any design basis event. The RHR pump discharge line low instrumentation does not necessarily relate directly to the RHR system operability. The CTS requires the RHR system discharge line to be maintained filled. Control of the availability of, and necessary compensatory activities if not available, for indications and monitoring instruments are addressed by station operational procedures and policies.

The requirement to ensure the RHR discharge line is maintained filled is sufficient. This instrumentation is not assumed to be operable to mitigate the consequences of a DBA or transient, and is not an input assumption for any DBA analysis. Thus, CTS 3/4.2.8.23 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the RHR pump discharge line instrumentation after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the RHR pump discharge line instrumentation to the TRM is acceptable.

8. 3/4.2.B.30. HFCI Turbine Hiah Exhaust Pressure (3.3.5.1 - R.7)

Table 3.2.B instrument 30; Table 4.2.8 HPCI instrument 4 3/4.2.B 31. HPCI Pumo Low Suction Pressure (3.3.5.1 - R.7)

Table 3.2.B instrument 31; Table 4.2.B HPCI instrument 5 3/4.2.B.38. HPCI Gland Seal Condenser Hotwell Level (3.3.5.1 - R.7)

Table 3.2.B Instrument 38; Tab's 4.2.B HPCI instrument 12 The function of the first two instruments, turbine high exhaust pressure and pump low suction pressure, is to provide a close signal to the HPCI turbine stop valve. In tum, the injection valve and the minimum flow valve will close. All of which will prevent the system from operating. Signals from any of these two instruments will result in the valves listed above receiving a signal to close (directly or indirectly). These instruments actuate to provide turbine / pump protection only to preclude turbine / pump damage and possible breach of the system. The valves are not credited with providing primary containment isolation on these signals, nor are they credited with closing to isolate a primary coolant leak on these signals. No design basis analysis takes credit for these instruments. The third instrument controls the HPCI gland seai condenser hotwell level.

With high hotwelllevel, HPCI cannot operate properly. Failure of this instrument could also result in failure of the HPCI System. Thus, it essentially provides HPCI pump protection, similar to the first two instruments. No design basis analysis takes credit for this instrument. Thus, CTS 3/4.2.B.30,3/4.2.B.31, and 3/4.2.B.38 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable.

Cooper Nuclear Station Draft Safety Evaluation

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3/4.2.B.42. HPCI Turbine Conditional Suoervisorv Alarm Actuation TimeJ (3.3.5.1 - R.8)

Table 3.2.B Instrument 42; Table 4.2.B HPCI instrument 14 3/4.2.B.43. HPCI Pumo Discharae Line Low Pressure (3.3.5.1 - R.8)

Table 3.2.B instrument 43; Table 4.2.B HPCI Instrument 15 The HPCI turbine conditional supervisory alarm actuation timer instrument provides an enabling signal to some of the HPCI alarms approximately 15 seconds after the HPCI steam supply valve comes off its closed seat. The HPCI pump discharge line low pressure instrument provides indication that the HPCI discharge line is not pressurized (i.e., drained down), resulting in the HPCI system being unable to perform its design function. However, these indication-only and monitoring instruments are not assumed to function in any DBA. The instruments do not necessarily relate directly to the HPCI system operability. The CTS require the HPCI system to be properly aligned, with power available, and with the discharge line maintained filled. Control of the availability of, and necessary compensatory activities if not available, for indications and monitoring instruments are addressed by station operational procedures and policies. The ITS requirement to ensure the HPCI system is operable is sufficient. These instruments are not assumed to be operable to mitigate the consequences of a DBA or transient, and are not any input assumption for any DBA analysis. Thus, CTS 3/4.2.B.42 and 3/4.2.B.43 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements to the TRM is acceptable.

10. 3/4.2 B.45. RCIC Hiah Turbine Exhaust Pressure (3.3.5.2 - R.1)

Table 3.2.B instrument 45; Table 4.2.B RCIC instrument 3 3/4.2.B.46. RCIC Low Pumo Suction Presste (3.3.5.2 - R.1)

Table 3.2.B instrument 46; Table 4.2.B RCIC Instrument 4 3/4.2.B.50. RCIC Low Pumo Discharce Fjow (3.3.5.2 - R.1)

Table 3.2.B instrument 50; Table 4.2.B RCIC Instrument 8 The function of the RCIC high turbine exhaust pressure instrument and the RCIC Low pump suction pressure instrument is to provide a close signal to the RCIC turbine trip throttle valve and the minimum flow valve, all of which will prevent the system from operating. Signals from any of these two instruments will result in the valves listed above receiving a signal to close (directly or indirectly). These instruments actuate to provide turbine and pump protection only to preclude turbine and pump damage and possible breach of the system. The valves are not credited with providing primary containment isolation on these signals, nor are they credited with closing to isolate a primary coolant leak on these signals. No design basis analysis takes credit for these instruments. The RCIC low pump discharge flow instrument controls the minimum flow valve and failure of this valve could result in failure of the RCIC system. The instrument actuates to provide pump protection only to preclude pump damage. No design basis Cooper Nuclear Station Draft Safety Evaluation

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.. analysis takes credit for these instruments. Thus, CTS 3/4.2.B.45 3/4.2.B46, and 1

3/4.2.B.50 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable.

11. 3/4.2.B 51. RCIC Pumo Discharae Line Low Pressure (3.3.5.2 - R.2)

Table 3.2.B Instrument 51; Table 4.2.B RCIC instrument 9 3/4.2.B 52. RCIC Turbine Conditional Suoervisorv Alarm Timer (3.3.5.2 - R.2) l Table 3.2.B instrument 52; Table 4.2.B RCIC Instrument 10 l

The RCIC pump discharge line low pressure instrument provides indication that the RCIC discharge line is not pressurized (i.e., drained down), resulting in the RCIC system i- being unable to perform its design function. The RCIC turbine conditional supervisory l alarm timer instrument provides an enabling signal to some of the RCIC alarms approximately 15 seconds after the RCIC steam supply valve comes off its closed seat.

However, these indication-only and monitoring instruments are not assumed to function in any DBA. The instruments do not necessarily relate directly to the RCIC system operability . The CTS require the RCIC system to be properly aligned, with power available, and with the discharge line maintained filled. Control of the availability of, and necessary compensatory activities if not available, for indications and monitoring instruments are addressed by station operational procedures and policies. The ITS requirement to ensure the RCIC system is operable is sufficient. Thus, CTS 3/4.2.B.51 and 3/4.2.B.52 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after

[ they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. .

L Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued

! protection of the public health and Safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable. -

12. 4.2.B.1. ADS Inhibit Switch (3.3.5.1 - R.10)

Table 4.2.B ADS Instrument 1  !

(Note: Table 3.2.B instrument 47, ADS Timer, is retained in ITS Table 3.3.5.1-1 )

l Functions 4.b and 5.b)

The automatic depressurization system (ADS) inhibit switch allows the operator to defeat ADS actuation as directed by the emergency operating procedures (EOPs) under j conditions for which ADS would not be desirable. For example, during an anticipated  ;

transient without scram (ATWS) event low pressure ECCS system activation would  !

dilute sodium pentaborate injected by the standby liquid control system (SLCS) thereby 4 reducing the effectiveness of the SLCS shutdown. The ADS inhibit switch function in CTS Table 4.2.B is an operational function only and is not considered in any DBA or transient, it does provide mitigation of the consequences of a non-design basis ATWS Cooper Nuclear Station Draft Safety Evaluation l

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. event; however the evaluation summarized in NEDO-31466, November 1987, determined that loss of ADS manual inhibit switch function to be a non-significant risk contributor to core damage frequency and offsite release. Thus, CTS 4.2.B.1 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the ADS inhibit switch after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and Safety, and the relocation of the CTS requirements for the ADS inhibit switch to the TRM is acceptable.

13. 4.2.B. Loaic Bus Power Monitors for Core Sorav. LPCI. HPCI. and ADS Control Power (3.3.5.1 - R.9)

Table 4.2.B CS, RHR, HPCI, and ADS Logic Instrument 1 Loaic Bus Power Monitors for RCIC (3.3.5.2 - R.3) l Table 4.2.B RCIC Logic Instrument 1  !

The bus power monitors for the RHR (LPCI), CSS, HPCI, and RCIC trip systems and j the control power monitor for ADS trip systems alarm if a fault is detected in the power system to the logic of the appropriate system. No DBA or transient analysis takes credit  ;

for the bus power monitors or control power monitors. This instrumentation only provides a monitoring / alarm function. These instruments are not assumed to be operable to mitigate the consequences of a DBA or transient, and are not any input assumption for any DBA analysis. Thus, the requirements of CTS 4.2.B for these l instruments do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after they relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and Safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable.

14. 3/4.2.C.1. APRM Uoscale (Flow Bias) (3.3.2.1 - R.1) i 3/4.2.C.2. APRM Unscale (Startuo) (3.3.2.1 - R.1) 3/4 2.C.3. APRM Downscale (3.3.2.1 - R.1) 3/4.2.C.4. APRM inoperative (3.3.2.1 - R.1)

Tables 3.2.C and 4.2.C, Functions 1 through 4

i. The average power range monitor (APRM) control rod blocks function to prevent a l control rod withdrawal error during power range operations utilizing LPRM signals to l- create the APRM rod block signal. APRMs provide information about the average core i power; however, there are no safety analyses that depend upon these rod blocks to l prevent, mitigate, or establish initial conditions for DBA or transients. The evaluation L summarized in NEDO-31466 determined that the loss of the APRM, IRM, SRM, recirculation flow, and scram discharge volume rod blocks would be a non-significant risk contributor to core damage frequency and offsite releases. The results of this evaluation have been determined to be applicable to CNS. Thus, CTS 3/4.2.C.1, Cooper Nuclear Station Draft Safety Evaluation l

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. 3/4.2.C.2,3/4.2.C.3, and 3/4.2.C.4 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after they are relocated to the TRM will require a safety evaluation pursuant i

to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and Safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable.

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15, 3/4.2.C.9. IRM Unscale (3.3.2.1 - R.1)

( 3L4.10.,10JEM Donnacala (3.3.2.1 - R.1) l 3/4.2.C.11. IRM Detector Not Full in (3.3.2.1 - R.1) 3/4.2.C.12. IRM inoperative (3.3.2.1 - R.1)

Tables 3.2.C and 4.2.C, Functions 9 through 12 l The intermediate Range Monitor (IRM) control rod blocks function to prevent a control l rod withdrawal error during reactor startup utilizing IRM signals to create the rod block signal. IRMs are provided to monitor the neutron flux levels during refueling, shutdown, i- and startup conditions. No DBA or transient analysis takes credit for rod block signals initiated by IRMs. The evaluation summarized in NEDO-31466 determined that the loss of the APRM, IRM, SRM, recirculation flow, and scram discharge volume rod blocks l would be a non-significant risk contributor to core damage frequency and offsite

! releases. The results of this evaluation have been determined to be applicable to CNS.

Thus, CTS 3/4.2.C.9,3/4.2.C.10,3/4.2.C.11 and 3/4.2.C.12 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these l requirements regarding these instruments after they are relocated to the TRM will l_

require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, l ' sufficient regulatory controls exist to ensure continued protection of the public health and Safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable.

16. 3/4.2 C.13. SRM Unscale (3.3.2.1 - R.1) j 3/4.2.C.14. SRM Detector Not Full In (3.3.2.1 - R.1) ,

i 3/4.2.C.15. SRM inoperative (3.3.2.1 - R.1) l l 3/4.2.C.18. SRM Downscale (3.3.2.1 - R.1) l L Tables 3.2.C and 4.2.C, Functions 13 through 15, and Function 18 )

l 1 l The SRM control rod blocks function to prevent a control rod withdrawal error during reactor startup utilizing SRM signals to create the rod block signal. BRM signals are L used to monitor neutron flux during refueling, shutdown and startop conditions. No DBA  !

[ or transient analysis takes credit for rod block signals initiated by the SRMs. The evaluation summarized in NEDO-31466 determined that the losn of the APRM, IRM, SRM, recirculation flow, and scram discharge volume rod blocks would be a non-significant risk contributor to core damage frequency and offsite releases. The results of i this evaluation have been determined to be applicable to CNS. Thus, CTS 3/4.2.C.13, 3/4.2.C.14,3/4.2.C 15 and 3/4.2.C.18 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding i

Cooper Nuclear Station Draft Safety Evaluation  !

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. these instruments after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and Safety, and the relocation of the CTS requirements for these instruments to the TRM is acceptable.

17. 3/4.2.C.16. Recircu:stion Flow = Flow Bias Comoarator (3.3.2.1 - R.1) l 3/4.2.C.17. Recirculation Flow = Flow Bias Uoscale/Inoo (3.3.2.1 - R.1)-

Tables 3.2.C and 4.2.C, Functions 16 and 17 The recirculation flow rod blocks are intended to prevent control rod withdrawal when station conditions make such withdrawalimprudent. An increase in reactor recirculation flow causes an increase in neutron flux which results in an increase in reactor power. 4 i

However, this increase in neutron flux is monitored by the neutron monitoring system l which has the capability of providing a reactor scram when required. No DBA or transient analysis takes credit for rod block signals initiated by the reactor coola-recirculation system. There are no safety analyses that depend upon these rod blocks to prevent, mitigate or establish initial conditions for design basis accidents or transients.

The evaluation summarized in NEDO-31466 determined that the loss o' the APRM, IRM, SRM, recirculation flow, and scram discharge volume rod blocks would be a non- I significant risk contributor to core damage frequency and offsite releases. The results of I l this evaluation have been determined to be applicable to CNS. Thus, CTS 3/4.2.C.16 l and 3/4.2.C.17 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding these instruments after l they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. 1 Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued -  !

protection of the public health and Safety, and the relocation of the CTS requirements l l for these instruments to the TRM is acceptable. I l

18. 3/4.2.C 19. SDV Water Level Hiah (3.3.2.1 - R.1)-

Tables 3.2.C and 4.2.C, Function 19 l

l The SDV control rod block functions to prevent control rod withdrawals during power range operations, utilizing SDV signals to create the rod block signal if water is accumulating in the SDV. The purpose of measuring the SDV water level is to ensure l that there is sufficient volume to contain the water discharged by the CRDs control rod I drives during a scram, thus ensuring that the control roos will be able to insert fully. This rod block signal provides an indication to the operator that water is accumulating in the SDV and prevents further rod withdrawals. With continued water accumulation, a RPS-initiated scram signal will occur. Thus, the SDV water level rod block signal provides an opportunity for the operator to take action to avoid a subsequent scram. No DBA or transient takes credit for rod block signals initiated by the SDV instrumentation. Thus, CTS 3/4.2.C.19 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding SDV control rod block instrumentation after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory Cooper Nuclear Station Draft Safety Evaluation l

. controls exist to ensure continued protection of the public health and Safety, and the relocation of the CTS requirements for SDV control rod block instrumentation to the TRM is acceptable.

19. 3/4.2 D.1. Steam Jet Air Eiector Off-Gas Svstem (3.3.6.1 - R.1)

Tables 3.2.D and 4.2.D instrument Channel 1 Table 4.2.D Logic System 1 The radioactive gas processing system is not a safety system and is it not connected to the primary coolant piping (i.e., the RCS). The offgas post-treatment and pre-treatment monitors are used to show conformance with the discharge limits of 10 CFR Part 20.

ITS 3.7.5 of the air ejector offgas is adequate to ensure the guidelines in 10 CFR 100 for accidents are not exceeded. Information provided by these instruments on the radiation

- levels would have limited or no use in identifying and assessing core damage and they are not installed to detect excessive reactor coolant leakage. These monitors are not l I

assumed to be operable to mitigate the consequences of a DBA or transient, and are not input assumptions for any DBA analysis. Thus, CTS 3/4.2.D.1 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding offgas post-treatment and pre-treatment monitors after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for offgas post-

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l treatment and pre-treatment monitors to the ODAM is acceptable. j i

20. 3/4.2.D.3. Liouid Radwaste Discharoe Isolation (3.3.6.1 - R.2)

Tables 3.2.D and 4.2.D Instrument Channel 3 Table 4.2.D Logic System 4 The radioactive liquid radwaste system is neither a safety system nor is it connected to the primary coolant piping. This instrumentation is used for the purpose of showing conformance to the discharge limits of 10 CFR Part 20. It is not installed to detect excessive reactor coolant leakage. The radioactive liquid effluent monitors are used routinely to provide continuous check on the release of radioactive liquid effluent from the normal station liquid effluent flow paths. The CTS require the licensee to maintain

operability of various liquid effluent monitors and establish setpoints in accordance with j the ODAM. The alarm / trip setpoints are established to ensure that the alarm / trip will l- occur prior to exceeding the limits of 10 CFR Part 20. Station DBA analyses do not l assume any action, either automatic or manual, resulting from radioactive effluent l monitors. Thus, CTS 3/4.2.D.3 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding liquid radwaste discharge isolation monitoring instrumentation after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health

l. and safety, and the relocation of the CTS requirements for liquid radwaste discharge isolation monitoring instrumentation to the ODAM is acceptable.

Cooper Nuclear Station Draft Safety Evaluation

. 21, 3/4 2 F. Primarv Containment Surveillance Instrumentation (3.3.3.1 - R.1)

Tables 3.2.F and 4.2.F instruments:

2. Reactor Pressure

4. Drywell Temperature

5. Suppression Chamber / Torus Air Temperature

7. Suppression Chamber / Torus Water Level

8. Suppression Chamber / Torus Pressure

9. Control Rod Position (indicating lights)

10. Neutron Monitoring (APRM) l 3/4.2.H. Post-Accident Monitoring (3.3.3.1 - R.1)

Tables 3.2.H and 4.2.H instruments:

1. Elevated Release Point Monitor (High Range Noble Gas)

2. Turbine Building Ventilation Exhaust Monitor (High Range Noble Gas)

3. Radwaste/ Augmented Radweste Exhaust Monitor (High Range Noble Gas) 3.6.D.4 and 4.6.D.6. SRV Position Indication (3.3.3.1 - R.1)

Each individual accident monitoring parameter has a specific purpose, however, the general purpose for all accident monitoring instrumentation is to provide sufficient information to confirm an accident is proceeding per prediction (i.e., automatic safety systems are performing properly, and deviations from expected accident course are minimal). These instruments in CTS Tables 3.2.F and 3.2.H, and CTS 3.6.D.4 are not credited as Category 1 or Type A variables. This evaluation was summarized in the NPC SE, dated October 27,1986. Further, the loss of these instruments is a non-significant risk contributor to core damage frequency and offsite release. Thus, the requirements specified for these functions in CTS Tables 3.2.F and 3.2.H, including the applicable actions, CTS Tables 4.2.F ar.d 4.2.H, CTS 3.6.D.4 and CTS 4.6.D.6.a and b do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. -

Any changes to these requirements regarding these instruments after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for these j instruments to the TRM is acceptable.

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22. 4.5.A 3.f. Drvwell and Sunoression Pool Sorav Air Test (4.5.A.3.f - R.1)

The drywell and suppression pool spray subsystem provides drywell and suppression pool spray capability as an attemate method for reducing containment pressure following a LOCA. The drywell spray headers condense any steam that may exist in the drywell, thereby lowering containment pressure, The suppression pool spray headers serve to cool any non-condensible gases collected in the free volume above the j suppression pool. Neither drywell or suppression pool spray is credited in any DBA (i.e., j they are not needed to function to mitigate the consequences of any DBA). They are considered secondary actions in the EOPs. Thus, CTS 4.5.A.3.f does not meet any of )

the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to l Cooper Nuclear Station Draft Safety Evaluation {

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e

_. I these requirements regarding drywell and suppression pool spray subsystems and associated air test after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the drywell and suppression pool spray subsystems and the associated air test to the TRM is acceptable.

23. 3/4.6.B.2. 3/4 6.B.3. 3/4 6.B.4. and 3/4.6.B.5. Chemistry (3/4.6.B - R.1)

Poor reactor coolant water chemistry may contribute to the long term degradation of system materials and, thus, is not of immediate importance to the station operator.

Reactor coolant water chemistry is monitored for a variety of reasons. One reason is to reduce the possibility of failures in the reactor coolant system pressure boundary

, caused by corrosion. Severe chemistry transients have resulted in failure of thin-walled LPRM instrument dry tubes in a relatively short period of time. However, these LPRM dry tube failures result in loss of the LPRM function and are, therefore, readily -

detectable. In summary, the chemistry monitoring activity serves a long term preventive rather than a short term mitigative purpose. Thus, CTS 3/4.6.B.2,3,4, and 5 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding reactor coolant water chemistry after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for reactor coolant water chemistry to the TRM is acceptable.

24, 3/4.6.G. Inservice insoection (ISI) (3/4.6.G - R.1)

The ISI programs at the CNS for the ASME Code Class 1,2, and 3 components ensure that the structural integrity of these components will be maintained throughout the life of

i. the component and the requirements in CTS 3/4.6.G on ISI is directed toward prevention of component degradation and continued long term maintenance of acceptable structural conditions. However, there are also other requirements in the CTS for the safety systems to be operable (for example, CTS 3/4.5.A for ECCS) and in a ready state for mitigative action and it is not necessary to retain CTS 3/4.6.G to L

ensure immediate operability of safety systems. Further, CTS 3/4.6.G prescribes inspection requirements which are performed during station shutdown and, therefore, l not directly important for responding to the DBAs at power. Thus, CTS 3/4.6.G does not L meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the inspections stipulated by CTS 3/4.6.G l after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the ISI inspections in CTS 3/4.6.G to the TRM is acceptable.

Cooper Nuclear Station Draft Safety Evaluation i

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l L , 25. 3/4.6.H. Shock Sunoressors (Snubhars) (3/4.6.H - LA.1) i

! The inspection and testing requirements in CTS 3/4.6.H do not demonstrate the l snubbers are operable, and are redundant to the ISI program on snubbers requirea by 10 CFR 50.55a. Compliance with 10 CFR 50.55a, and the ISI program, is required by the CNS operating license. These controls are adequate to ensure the required inspections and testing to demonstrate operability are performed. With the removal of explicit operability requirements for snubbers from the CTS, snubber Operability -

requirements will be determined in accordance with the Operability requirements in the j ITS for those systems designed with snubbers. Thus, CTS 3/4.6.H does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the inspections and testing stipulated by CTS 3/4.6.H after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure ,

continued protection of the public health and safety, and the relocation of the I inspections and testing of snubbers in CTS 3/4.6.H to the TRM is acceptable,

26. 3/4.10.D. Time Limitation (Decav Time) (3/4.10.D - LA.1)

The minimum time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following reactor shutdown which must elapse prior to L irradiated fuel movement in or above the reactor, will always be met for a refueling i outage because of the operations required prior to moving irradiated fuel in the reactor

vessel (e.g., containment entry, removal of drywell head, removal of vessel head,'

j removal of vessel intemals). Thus the time limit in CTS 3/4/10.D is not necessary to be ,

in the ITS to ensure decay the heat load has decreased sufficiently so that movement of

! irradiated fuel can be conducted safely. Because CTS 3/4.10.D does not in practice l Impose any limitation on how soon following shutdown fuel movement may begin, it

! does not meet any of the criteria in 10 CFR 50.36 and, thus may be relocated out of the ,

l CTS. Any changes to this time limit after it is relocated to the TRM will require a safety i l evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, l

and the relocation of the decay time limitation in CTS 3/4.10.D to the TRM is acceptable.  !

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! 27, 3/4.10.H. Soent Fuel Cask Handlina (3/4.10.H - R.1)

L Limitations on lifting spent fuel casks ensures that only the proper equipment will be used to handle fuel or casks within the storage pool, hoists have sufficient load capacity for handling fuel assemblies or other loads, and the possibility of dropping a load on the l core intemals and pressure vessel are minimized during lifting operations. Administra- '

L tively limiting the manner in which a spent fuel cask is handled on the refueling floor serves as a backup to minimize the damage that could result from dropping the cask.

Although CTS 3/4.10.H supports the maximum refueling accident assumption in the fuel handling DBA, the spent fuel cask handling limits are not monitored and controlled during operation; the limits are checked on a periodic basis to ensure operability of the handling equipment. Administrative monitoring of loads moving over the fuel storage Cooper Nuclear Station Draft Safety Evaluation

, l racks serves as a backup to the crane interlocks. Thus, CTS 3/4.10.H does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes i to these requirements regarding the spent fuel cask handling limits after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the spent fuel cask handling limits to the TRM is acceptable.

28. 3/4.11.B. Linear Heat Generation Rate (3/4.11.B - LA.1)

CTS 3/4.11.B which applies to the Linear heat generation rate (LHGR), is encompassed by the Average planar linear heat generation rate (APLHGR). ITS 3.2.1 adequately addresses the APLHGR requirements, thereby eliminating the need for a LHGR specification in the TS. This is consistent with a letter from A.C. Thadani (NRC) to J.S. Charnley (GE), " Acceptance for Referencing of Amendment 19 to General Electric Licensing Topical Report NEDE-24011-P-A (GESTAR-II), General Electric Standard Application for Reactor Fuel," dated April 7,1987. GESTAR-Il requires that the LHGR be monitored by the station computer, in keeping with the GESTAR-il requirements, the CTS 3/4.11.B LHGR requirements are proposed to be relocated to the TRM. Because the LHGR is not an initial condition assumed in any DBA or transient, it does not meet any of the criteria in 10 CFR 50.36. Thus the requirements on LHGR are not required to be in the ITS to provide adequate protetion of the public health and safety. Any changes to these requirements regarding LHGR in CTS 3/4.11.B after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the LHGR requirements in CTS 3/4.11.8 to the TRM is acceptable.

29. 3/4.12.D. Batterv Room Ventilation (3/4.12.D - R.1)

The requirements in CTS 3/4.12.D conceming operability and testing of the battery room exhaust fans are proposed to be relocated to the TRM. This system is not assumed to function during an accident nor does it act to mitigate the consequences of an accident.

The control building essential ventilation system provides ventilation flow to essential areas of the control building during emergency conditions. The battery room ventilation system was designed only to ensure the removal of hydrogen generated by the station batteries, a function no longer necessary due to the use of lead-calcium cells which do not generate significant amounts of hydrogen. The operability and testing requirements contained in CTS 3/4.12.D are not required to be included in the ITS to provide adequate protection of the public health and safety. Thus, CTS 3/4.12.D does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the battery room exhaust fans after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the Cooper Nuclear Station Draft Safety Evaluation I

l-. public health and safety, and the relocation of the CTS requirements for the battery room exhaust fans to the TRM is acceptable.

l

30. 3/4.13. River Level (3/4.13 - R.1)

A high river water level is a preliminary indication of flood conditions at the station site.

A flood is not a DBA or transient, thus river water levelis not credited in any safety analysis. The river water level TS requirements were put in place to ensure that actions to place the station in a safe condition with rising water level are performed in a timely manner. The requirement to implement the site flood procedure, and to shutdown the station if the river reaches 902 feet, ensures that safety related equipment will not be adversely impacted by a flooding event prior to the unit being shutdown. However, this is only an external event and is not assumed in any DBA or transient. Thus, CTS 3/4.13 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS.

Any changes to these requirements regarding the river level requirement after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the river level requirements in CTS 3/4.13 to the TRM is acceptable.

31. 3/4.14. Fire Detection Systems (3/4.14 - R.1)

The fire detection system instrumentation is used to detect and suppress fires in the station; however, the instruments are not used to detect a degradation of the RCS pressure boundary or mitigate a DBA or transient event. The fire detection instrumentation is required by 10 CFR 50.48 and 10 CFR 50, Appendix R. The acceptability of the relocation of the fire detection instrumentation from the TS has already been endorsed by the NRC as indicated in NRC GLs 86-10 and 88-12. The requirements in CTS 3/4.14 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the fire detection system instrumentation after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the fire detection system instrumentation to the TRM is acceptable.

32. 3/4.15. Fire Sucoression Water System (3/4.15 - R.1)

The purpose of the station fire suppression water system is to ensure that a postulated fire can either be confined or extinguished in any area of the station where safety-related equipment is located. The fire suppression capability, however, is beyond the DBA for the CNS. Consequently, the fire suppression water system is not considered in the station safety analyses and do not serve to detect, control, or mitigate DBA or transient events; however, the fire suppression water system is required by 10 CFR 50.48 and 10 CFR 50, Appendix R. The acceptability of the relocation of the fire Cooper Nuclear Station Draft Safety Evaluation 1

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. suppression water system from the station TS has already been endorsed by the NRC as indicated in NRC GLs 86-10 and 88-12. The requirements in CTS 3/4.15 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the fire suppression water system after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the fire suppression water system to the TRM is acceptable.

33. 3/4.16. Sorav and/or Sorinkler System (3/4.16 - R.1)

The purpose of the station spray and sprinkler system is to ensure that a postulated fire can either be confined or extinguished in any area of the station where safety-related equipment is located. The fire suppression capability, however, is beyond the DBAs for the CNS, and the spray and sprinkler system is not considered in the station safety analyses and do not serve to detect, control, or mitigate DBA or transient events. The spray and sprinkler system is required by 10 CFR 50.48 and 10 CFR 50, Appendix R.

The acceptability of the relocation of the spray and sprinkler system from the station TS has already been endorsed by the NRC as indicated in NRC GLs 86-10 and 88-12. The requirements in CTS 3/4.16 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the spray and sprinkler system after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the spray and sprinkler system to the TRM is acceptable.

34. 3/4.17. Carbon Dioxide and Halon Systems (3/4.17 - R.1)

The purpose of the station carbon dioxide and halon systems is to ensure that a postulated ' ire can either be confined or extinguished in any area of the station where safety-related equipment is located. The fire suppression capability, however, is beyond the DBA for the station, and the carbon dioxide and halon systems are not considered in the station safety analyses and do not serve to detect, control, or mitigate DBA or transient events. The carbon dioxide and halon systems are required by 10 CFR 50.48 and 10 CFR 50, Appendix R. The acceptability of the relocation of the carbon dioxide and halon systems from the station TS has already been endorsed by the NRC as indicated in NRC GLs 86-10 and 88-12. The requirements in CTS 3/4.17 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the carbon dioxide and halon systems after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the carbon dioxide and halon systems to the TRM is acceptable.

Cooper Nuclear Station Draft Safety Evaluation l

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. 35, 3/4.18. Fire Hose Stations (3/4.18 - R.1)

The purpose of the station fire hose stations is to ensure that a postulated fire can either be confined or extinguished in any area of,the station where safety related equipment is located. The fire suppression capability, however, is beyond the DBA for the station, and the fire hose stations are not considered in the station safety analyses and do not serve to detect, control, or mitigate DBA or transient events. The fire hose stations are required by 10 CFR 50.48 and 10 CFR 50, Appendix R. The acceptability of the relocation of the fire hose stations from the station TS has already been endorsed by the NRC as indicated in NRC GLs 86-10 and 88-12. Thus, CTS 3/4.18 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the fire hose stations after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the fire hose stations to the TRM is acceptable.

36. 3/4.19. Fire Barrier Penetration Fire Seals (3/4.19 - R.1)

Fire barriers and barrier penetration seals are provioad in the station to ensure that fire damage to safety-related fire areas, and/or to areas of redundant systems important to safe shutdown, is limited. They are designed to minimize the possibility of a single fire involving more than one fire area before the detection and extinguishment of the fire.

This capability, however, does not detect, control, or mitigate DBA or transient events.

Fire barrier penetration fire seals are required by 10 CFR 50.48 and 10 CFR 50, Appendix R. The acceptability of the relocation of the fire barrier penetration fire seals from the station TS has already been endorsed by the NRC as indicated in NRC GLs 86-10 and 88-12. The requirements in CTS 3/4.19 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the fire barrier penetration fire seals after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the fire barrier penetration fire seals to the TRM is acceptable.

37. 3/4.21.A 1. Liauid Effluent Monitorina (3/4.21.A.1 - R.1)

The radioactive liquid radwaste system is neither a safety system nor is it connected to l the RCS. The monitoring instrumentation for this system is used for the purpose of showing conformance to the discharge limits of 10 CFR part 20. It is not installed to detect excessive RCS leakage. The radioactive Iiquid effluent monitor is used routinely to provide continuous check on the release of radioactive liquid effluent from the normal station liquid effluent flow paths. CTS 3/4.21.A.1 requires the licensee to maintain operability of various liquid effluent monitors and establish setpoints in accordance with the ODAM that contains the methodology and parameters to calculate offsite does from Cooper Nuclear Station Draft Safety Evaluation 1

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^

l l

l . l radioactive liquid effluents. The alarm / trip setpoints are established to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. Station DBA l analyses do not assume any action, either automatic or manual, resulting from the l radioactive liquid effluent monitors. Thus, CTS 3/4.21.A.1 does not meet any of the l criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the liquid effluent monitoring instrumentation after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the liquid effluent monitoring instrumentation to the ODAM is acceptable.

38. 3/4.21.A.2. Gaseous Effluent Monitorina (3/4.21.A.2 - R.1) l The radioactive gaseous effluent monitoring instrumentation is neither a safety system nor is it connected to the RCS system. The primary function of this instrumentation is to show conformance to the discharge limits of 10 CFR Part 20. This instrumentation is not installed to detect excessive RCS leakage. The radioactive gaseous effluent monitors are used routinely to provide continuous check on the releases of radioactive gaseous effluents from the normal station gaseous effluent flow paths. CTS 3/4.21.A.2 requires the licensee to maintain operability of various effluent monitors and establish setpoints in accordance with the ODAM that contains the methodology and parameters to calculate offsite does from radioactive gaseous effluents. The alarm / trip setpoints are established to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. Station DBA analyses do not assume any action, either automatic or manual, resulting from radioactive effluent monitors. In addition, the explosive gas monitor instrumentation is provided to ensure that the concentration of potentially explosive gas mixtures contained in the gaseous radwaste treatment system is adequately monitored, which will help ensure that the concentration is maintained below the flammability limit of hydrogen. However, the offgas system is designed to contain '

detonations and will not affect the function of any safety related equipment. The concentration of hydrogen in the offgas stream is not an initial assumption of any DBA l

or transient analysis. Thus, CTS 3/4.21.A.2 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the radioactive gaseous effluent monitoring instrumentation , after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the radioactive gaseous effluent monitoring instrumentation to the ODAM is acceptable.

39. 3/4.21.B.1. Liauid Effluents - Concentration (3/4.21.B.1 - R.1)

Appendix B, Table 2, to 10 CFR 20.1302(b)(20(i) provides the gaseous and liquid effluent concentration limits for people in an unrestricted area. The concentration limits in the CTS 3/4.21.B.1 are not an initial condition of a DBA or transient analysis, neither does the liquid effluent system comprise a part of a safety analyses or a part of the RCS 1

l Cooper Nuclear Station Draft Safety Evaluation l

. pressure boundary. Effluent controlis for protection against radiation hazards from licensed activities which is adequately controlled by the regulations, and is not protection against any DBA. Administrative controls are included in ITS 5.6.3 on the radioactive effluent release report, ITS 5.5.4 on the radiological effluents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements.

Proposed ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix I and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations." Thus, CTS 3/4.21.B.1 does not meet any of the criteria in 10 CFR 50.36 end may be relocated out of the CTS. Any changes to these requirements regarding the radioactive liquid effluent concentration limits after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the radioactive liquid effluent concentration limits to the ODAM is acceptable.

40. 3/4.21.B.2. Liould Effluents - Liould Dose (3/4.21.B.2 - R.1)

Limitations of the quarterly and annual projected doses to members of the public which results from cumulative station liquid effluent discharges during normal operation over l extended periods is intended to assure compliance with the dose objectives of 10 CFR

- Part 50, Appendix 1. These limits are not related to protection of the public from any DBA or transient. Administrative controls are included in ITS 5.6.3 on the radioactive effluent release repoit, ITS 5.5.4 on the radiological effluents control program, ITS 5.5.4 on the radiological effluents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements. ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I and do not adversely impact the accuracy l or reliability of effluent, dose, or setpoint calculations." Thus, CTS 3/4.21.B.2 does not j meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any j changes to these requirements regarding the radioactive liquid effluent dose limits after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, j under ITS 5.5,1, sufficient regulatory controls exist to ensure continued protection of the 1- public health and safety, and the relocation of the CTS requirements for the radioactive liquid effluent dose limits to the ODAM is acceptable.

41. 3/4 21.C.1. Gaseous Effluents - Concentration (3/4.21.C.1 - R.1)

Appendix B, Table 2, to 10 CFR 20.1302(b)(20(i) provides the gaseous and liquid effluent concentration limits for people in an unrestricted area. The concentration limits in the CTS 3/4.21.C.1 are not an initial condition of a DBA or transient analysis; neither does the gaseous effluent system comprise a part of a safety analyses or a part of the HCS pressure boundary. Effluent controlis for protection against radiation hazards from Cooper Nuclear Station Draft Safety Evaluation r

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. licensed activities which is adequately controlled by the regulations, and is not protection against any DBA. Administrative controls are included in ITS 5.6.3 on the radioactive effluent release report, ITS 5.5.4 on the radiological effluents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements.

Proposed ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix l and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations." Thus, CTS 3/4.21.C.1 does not toeet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the radioactive gaseous effluent concentration limits after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the radioactive gaseous effluent concentration limits to the ODAM is' acceptable.

42. 3/4.21.C 2. Gaseous Effluents - Noble Gases Dose (3/4.21.C.2 - R.1)

Limitation of the quarterly and annual air doses from noble gases in station gaseous effluents during normal operation over extended periods is intended to assure compliance with the dose objectives of 10 CFR Part 50, Appendix 1. These limits are not related to protection of the public from any DBA or transient. Administrative controls are included in ITS 5.6.3 on the radioactive effluent selease report, ITS 5.5.4 on the radiological effluents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements. ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will" maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix ! and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations " Thus CTS 3/4.21.C.2 does not meet any of the criteria in 10 CFR 50.3-6 and may be relocated out of the CTS. Any changes to these requirements regarding the gaseous effluent noble gases dose limits after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the gaseous effluent noble gases dose limits to the ODAM is acceptable.

43. 3/4.21 C.3. Gaseous Effluents -lodine and Particulate (3/4.21.C.3 - R.1)

Limitation of the quarterly and annual projected doses to members of the public from radionuclides other than noble gases during normal operation over extended periods is intended to assure compliance with the dose objectives of 10 CFR Part 50, Appendix 1.

These limits are not related to protection of the public from any DBA or transient.

Administrative controls are included in ITS 5.6.3 on the radioactive effluent release report, ITS 5.5.4 on the radiological effluents control program, and ITS 5.5.1 on the l

ODAM to ensure compliance with the applicable regulatory requirements. ITS 5.5.1 l Co.oper Nuclear Station Draft Safety Evaluation

l 1 . specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix 1 and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations." Thus, CTS 3/4.21.C.3 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS.

Any changes to these requirements regarding the gaseous effluents iodine and particulate limits after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure l continued protection of the public health and safety, and the relocation of the CTS requirements for the gaseous effluents iodine and particulate limits to the ODAM is acceptable.

l 44. 3/4 21.C 4. Gaseous Effluents - Gaseous R=la==as (3/4.21.C.4 - R.1)

. The requirements in CTS 3/4.21/C.4 apply to the train of charcoal adsorbers in the main condenser air ejector offgas system and to the exhaust ventilation treatment system (EVTS).

l The offgas system reduces the activity level of the non-condensible fission product i gases that are removed from the main condenser prior to their release to the l environment. The inlet offgas radioactivity is an initial condition of a DBA and is retained in ITS LCO 3.7.5; however, the operability of the offgas system is only required to meet the requirements of Appendix i to 10 CFR Part 50 (i.e., releases of.Moactive materials in gaseous effluents will be kept to Al. ARA levels) and the operability is not assumed in the analysis of any DBA or transient. Therefore, there is no need to retain the requirements on system operability in CTS 3/4.C.4.

The EVTS is intended to provide reasonable assurance that releases of radioactive materials during normal operation of the station are at ALARA levels and help assure compliance with the dose objectives of 10 CFR Part 50, Appendix 1. These objectives are not related to protection of the public from any DBA or transient. Administrative controls are included in ITS 5.6.3 on the radioactive effluent release report, ITS 5.G.4 on l the radiological effluents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements. ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations."

The requirements in CTS 3/4.21.C.4 do not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the gaseous radwaste treatment and EVTS after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the Cooper Nuclear Station Draft Safety Evaluation I

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. relocation of the CTS requirements for the offgas system and EVTS to the ODAM is acceptable.

45. 3/4.21.C.7. Gaseous Effluents - Containment (3/4.21.C.7 - R.1)

The drywell vent and purge system is used primarily to maintain the primary containment atmosphere, to reduce drywell a;rbome radioactivity levels before personnel entry, and also as a means of controlling drywell and torus pressure during abnormal conditions. The LCO in CTS 3.21.C.7 provides reasonable assurance that releases from normal drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for unrestricted areas. These limits are not related to protection of the public from any DBA or transient. Venting or purging of the drywell through the SGTS is not assumed to be operable to mitigate the consequences of a DBA or transient, and is not an input assumption for any DBA analysis. Administrative controls are included in ITS 5.6.3 on the radioactive emuent release report, ITS 5.5.4 on the radiological effluents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements. ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive emuent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix I and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations." Thus, CTS 3/4.21.C.7 does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the venting or purging of the drywell through the SGT system after they are relocated to the ODAM will have to be in accordance with ITS 5.5.1.

Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the ralocation of the CTS requirements for the venting or purging of the drywell through the SGTS to the ODAM is acceptable.

l 46. 3/4 21.D. Effluent Dose Liouid/ Gaseous (3/4.21.D - R.1) i This LCO limits in CTS 3.21.D on the annual doses to iridividual members of the public from all station sources are intended to assure that normal operation of the station is in compliance with the provisions of 40 CFR Part 190. These limits are not related to the l l protection of the public from any DBA or transient. The emuent dose liquid / gaseous limits are not assumed to be operable to mitigate the consequences of a DBA or transient, and are not input assumptions for any DBA analysis. Administrative controls are included in ITS 5.6.3 on the radioactive emuent release report, ITS 5.5.4 on the l radiological emuents control program, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements. ITS 5.5.1 specifies that future changes to i the ODAM will be reviewed to ensure that such changes will " maintain the levels of

l. radioactive emuent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, i

and 10 CFR 50, Appendix 1, and do not adversely impact the accuracy or reliability of emuent, dose, or setpoint calculations." Thus, CTS 3/4.21.D does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the emuent dose liquid / gaseous limits after they are relocated to 1

Cooper Nuclear Station Draft Safety Evaluation

l l the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, I

sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the effluent dose liquid / gaseous limits from all station sources to the ODAM is acceptable.

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! 47. 3/4.21.E Solid Radioactive W==+= (3/4.21.E - R.1) l The solid radweste system is designed to collect, process, package, and provide temporary storage of radioactive solid wastes prior to shipment offsite for disposal. The system must meet the requirements in (1) the GDC 60,61, and 63 of 10 CFR Part 50, Appendix A, on control of releases of radioactivity to the environment, storage and handling of radioactivity, and monitoring waste storage at tha station and (2) 10 CFR Part 71 on packaging of solid wastes for shipment offsite. M3 system serves to control the creation, storage, packaging, and shipment of solid waste, and not to any accidental release of radioactivity. Hence, the solid radwaste system is not assumed to be operable to mitigate the consequences of a DBA or transient, and is not an input assumptions for any DBA analysis. Administrative controls are included in ITS 5.6.3 on the radioactive effluent release report that includes solid waste shipped from the site, l ITS 5.6.1 on the occupational radiation report, and ITS 5.5.1 on the ODAM to ensure compliance with the applicable regulatory requirements. Thus, CTS 3/4.21.E does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the solid radwaste system after they are relocated to the TRM will require a safety evaluation pursuant to 10 CFR 50.59.

Therefore, under 10 CFR 50.59, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the solid radwaste system to the TRM is acceptable.

48, 3/4 21.F. Monitorina Proaram (3/4.21.F - R.1)

'The radiological environmental monitoring program required by CTS 3/4.21.F provides l measurements of radiation and of radioactive materials in those exposure pathways and i for those radionuclides which lead to the highest potential radiation exposures for I

members of the public resulting from station operation. This program monitors the long term impact of normal station operations and is not used for accident monitoring. These requirements are not related to protection of the public from any DBA or transient and i the monitoring program is not assumed to be operable to mitigate the consequences of l

a DBA or transient, and is not an input assumptions for any DBA analysis.

Administrative controls in ITS 5.5.1 on the conduct of the radiological environmental monitoring program ensure compliance with the applicable regulatory requirements.

Also, ITS 5.5.1 specifies that future changes to the ODAM will be reviewed to ensure that such changes will " maintain the levels of radioactive effluent control required by 10 CFR 20.1302,~40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix 1 and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations."

.. Thus, CTS 3/4.21.F does not meet any of the criteria in 10 CFR 50.36 and may be l

relocated out of the CTS. Any changes to these requirements regarding the radiological i

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. environmental monitoring program after they are relocated to the ODAM will have to be in accordance with ITS a '5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the radiological 1 environmental monitoring program to the ODAM is acceptable.

4g. 3/4.21.G. Interlaboratory Comnarison Proaram (3/4.21.G - R.1)

The interlaboratory comparison program required by CTS 3/4.21.G confirms the accuracy of the measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures for members of the public resultin0 from station operation. This

j. program ensures independent checks on the precision and accuracy of the l instrumentation used in the measurements of radioactive material for the radiological l environmental monitoring program are performed. This program is not utilized for any post accident monitoring. These requirements are not related to protection of the public

. from any DBA or transient. ITS 5.5.1 specifies that future changes to the ODAM will be l

reviewed to ensure that such changes wiiP' maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix I and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations." Thus, the interlaboratory comparison program is not assumed to be Operable to mitigate the consequences of a DBA or transient, and is not an input assumptions for any DBA analysis. Thus, CTS 3/4.21.G does not meet any of the criteria in 10 CFR 50.36 and may be relocated out of the CTS. Any changes to these requirements regarding the interlaboratory comparison program after they relocated to the ODAM will have to be in accordance with ITS 5.5.1. Therefore, under ITS 5.5.1, sufficient regulatory controls exist to ensure continued protection of the public health and safety, and the relocation of the CTS requirements for the interlaboratory comparison program to the ODAM is acceptable.

The relocated specifications from the CTS discussed above are not required to be in the TS because they do not meet any criteria 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 the effect of the provisions in these specifications. The NRC staff has concluded that appropriate controls have been established for all of the current specifications, information, and requirements that are being moved to the USAR, TRM, or ODAM. This relocation is the subject of a license condition discussed in Section V 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, the NRC will audit the removed provisions to ensure that an appropriate level of control has been i 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 Cooper Nuclear Station Draft Safety Evaluation l

I

SE, may be relocated from the CTS and placed in the identified licensee-controlled documents as specified in the licensee's letter dated March 27,1997. l 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 in Section Ill.D and Ill.E above. The facility and procedures described in the USAR and 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 ODAM can be changed in accordance with ITS 5.5.1, the emergency plan implementing procedures (EPIPs) can be changed in l accordance with 10 CFR 50.54(q); and the administrative instructions that implement the QA plan can be changed in accordance with 10 CFR 50.54(a) and 10 CFR Part 50, Appendix B. 1 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 QA plan for CNS and such applicable regulations as 10 CFR 50.59.

The license condition for the relocation of requirements from the CTS in Section V of this SE will address the implementation of the ITS conversion, and when the licensee stated that the

< relocation of the CTS requirements into licensee-controlled documents will be completed. The relocations to the USAR and TRM may be included in the next required update of these documents in accordance with 10 CFR 50.71(e).

l G. Evaluation of Other TS Changes included in the Application for Conversion to ITS This section addresses the beyond scope issues in which the licensee proposed changes to i both the CTS and STS. The staff has provided notices of consideration for these beyond scope l issues in the Federal Register; however, some of the notices issued for the proposed i amendments were provided for changes to the CTS that are now not considered beyond scope

! issues in that they are now not considered changes to both the CTS and STS.  ;

1 The changes discussed below are listed in the order of the applicable ITS specification or section, as appropriate (from ITS 3.1.8 to ITS 5.5.9). J

1. ITS 3.1.8 CTS 4 3.G. Scram Discharae Volume (SDV) Vent and Drain Valves. Revise STS l LCO 3.1.8 Action A to Reauire That. Within 7 Davs for Vent and Drain Lines with I One Inonerable Valve. the Lines Are Isolated instead of the Valves Are Restored to Operable Status (3.1.8 - L.4) i The CTS does not contain any LCO actions if one SDV vent and/or drain valve is inoperable and open. The licensee has proposed to isolate the associated line when one cuch valve is Cooper Nuclear Station Draft Safety Evaluation

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, inoperable, instead of requiring the valve to be restored to operable status as is given in the STS. The proposed change would be different from that required in the CTS and STS. j The purpose of the SDV is to serve as a collection volume for reactor coolant displaced by the CRD pistons during a reactor scram. The SDV is described in Sections 111-5.5.2.7 and ill- l 5.5.2.8 (Section 5.5.2.7 and 5.5.2.8 of Chapter lil) of the USAR. The SDV consists of header  ;

piping that connects to the scram outlet valves of each control rod hydraulic unit and drains into  !

an instrument volume. There are two headers and two instrument volumes, each receiving about half of the CRD piston discharges.

During normal operation, each SDV header collects water from the movement of any control l rods through the operation of the respective CRD pistons. The valves on the vent and drain lines are open, and the water collected is drained iiom the SDV. The primary function of the vent and drain valves (two valves in series in each line) is to isolate the SDV during a scram to prevent potentially highly radioactive liquid and gaseous being released from the reactor coolant system.

Each instrument volume has instrumentation to measure the amount of water accumulated in 4 the SDV. Two differential pressure transmitters and three level-measuring switches set at the same low, intermediate, and high levels on each instrument volume prevent operating the station without sufficient free volume present in the SDV to accommodate the water discharged during a reactor scram. The transmitters and level switches meet the requirements for redundancy and diversification in that each RPS division on each instrument volume has one transmitter and level switch for scram initiation.

With one of the two valves in a vent or drain line inoperable, the licensee proposed to repair the inoperable valve within 7 days or isolate the associated line with the inoperable valve. The STS stated to restore the inoperable valve within 7 days or enter LCO 3.0.3 because the Action can not be met (i.e., the inoperable valve can not be made operable). LCO 3.0.3 would have the station shut down. Therefore, the licensee has proposed to isolate the vent or drain line and continue operating. The proposed Action does not affect the Action for the case if both valves in the line are inoperable (i.e., Action B for ITS 3.1.8).

The act of isolating an SDV vent or drain line that has an inoperable valve will place the SDV in a condition where the vent and drain valves will perform their safety function (i.e., in the case of a reactor scram, the vent and drain lines will be isolated). Each affected vent and drain line could be separately opened under administrative control to permit draining and venting the SDV; however, with the transmitters and switches in the instrument volume, the RPS will ensure that the station is operating with sufficient free volume in the SDV to accommodate a

, scram. Therefore, what the licensee has proposed will not prevent the SDV from preforming its l safety function.

l Based on the above, tne staff concludes that the proposed change is acceptab'e.

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2. CNS Setooint Methodoloav Revise the Setooint and Allowable Values in ITS Section 3.3 From the Values in the CTS in the ITS conversion, the licensee has based the calibration frequencies and allowable values in ITS 3.3, " Instrumentation," upon the use of an NRC-approved GE topical report. The licensee has stated that it has verified the acceptability of TS setpoints and allowable values based on the GE instrument Setpoint Methodology in Topical Report NEDC-31336 dated October 1986, and the accompanying staff SE dated February 9,1993, approving the topical report. The setpoints and allowabie values based on this topical report may be different from the CTS and the STS.

l By letter dated September 29,1997, the licensee submitted an interim report describing a r

condition where safety-related setpoint calculations provided to the licensee by General Electric Nuclear Energy were found to contain non-conservative errors. However, these errors were primarily a result of calculation and math errors that had minimal impact on nuclear safety.

i Accordingly, no substantial safety hazard could have been created. Nevertheless, the licensee committed to reviewing and updating all setpoints and their associated calculations. While the ITS setpoint calculations have been reviewed by an independent third party to assure accuracy and correctness, the administrative processing and reformatting necessary for formal approval has yet to be completed. By letter dated April 16,1998, the licensee stated that it expects to j have a formal approval of the entire project completed by December 31,1998. At an April 7, l 1998 meeting with the staff the licensee agreed to include NEDC-31336 in the TS references and to correct inconsistencies with Engineering Procedure 3.26.3, Revision 3, " Instrument Setpoint and Channel Error Calculation Methodology" so that it accurately reflects the GE instrument setpoint methodology (NEDC-31336).

The licensee stated in its submittal of March 27,1997, that the setpoint calculations at the CNS are performed in accordance with engineering procedure 3.26.3 which is based upon NEDC-31336. The GE instrument setpoint methodology was approved by the staff in its SE dated February 9,1993 The methodology conforms to the guidelines of NRC Regulatory Guide, j l 1.105, Revision 2, " Instrument Setpoints for Safety-Related Systems" and ANSI /ISA-S67.04- l l 1982, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Plants."

The GE setpoint methodology has been previously reviewed and approved for use at other ]

l BWR facilities in support of an ITS conversion; however, the topical report is not to be used by l a licensee to validate its plant-specific setpoint values. The GE setpoint methodology is a generic methodology that requires plant-specific calculations with plant-specific data. The licensee stated that it has developed the ITS allowable values by performing plant-specific l analyses based on the setpoint methodology to incorporate CNS plant-specific information and operational considerations. The staff reviewed the licensee's calculations NEDC-90-384 and 92-0501 to determine that the GE setpoint methodology as outlined in NEC-31336 was indeed ,

being followed by the licensee and plant-specific data was being used to determine setpoint i values. The calculations include the plant-specific drift data to demonstrate that the drift for the sample instrumentation remains within the procedure drift allowance for the proposed 18-month  ;

extended surveillance interval. '

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l Based on the information submitted by the licensee, the staff concludes that the proposed ITS trip setpoints and allowable values are intended to maintain acceptable margins between operating conditions and trip setpoints, and do not significantly increase the likelihood of a false trip nor failure to trip upon dernand. In addition, the proposed 18-month surveillance interval  !

does not have a significant impact on safety and is, therefore, acceptable. Thus, the existing CNS licensing basis is not affected.

Based on the above, the staff concludes that the licensee's setpoint methodology using NEC-31336, and the resulting trip setpoints, allowable values, and surveillance interval incorporated in ITS 3.3 as part of the ITS conversion are consistent with the CNS licensing basis, and are, therefore, acceptable. '

3. ITS 3 3.2.1 CTS Table 3.2.C. Relocate The Uoscale Trio Level Settinos (or Allowable Values) for The Rod Block Monitor Unscale Trios to The COLR (3.3.2.1 - RL.2)

The CTS and the STS each contain a table of the control rod block instrumentation in which the upscale trip level settings (or allowable values) are listed for the low power, intermediate power,  !

and high power range of the RBM. The licensee proposed to specify these upscale allowable values in the COLR and not in Table 3.3.2.1-1 of ITS 3.3.2.1. Although item e in CTS 6.5.1.G on the COLR states that core operating limits for the RBM upscale setpoint shall be established for CTS Table 3.2.C in accordance with an NRC-approved methodology, the proposed change deletes settings from CTS Table 3.2.C and the corresponding table in the STS does not reference the COLR. Therefore, the proposed change would be different from that required in the CTS and STS. j As required in ITS 5.6.5 on the COLR, the COLR lists certain core operating limits that shall be  !

established prior to each reload cycle, or prior to any remaining portion of a reload cycle. The licensee has proposed to include the upscale allowable values for the three power-range RBM trips for ITS 3.3.2.1 as one of the core operating limits specified in the COLR. The COLR also lists the analytical methods that have been approved by NRC to determine the core operating l limits listed in the COLR and that, in accordance with ITS 5.6.5, will be the only methods allowed to determine these limits. For the RBM upscale allowable values, the licensee has l specified NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," as the i NRC-approved analytical method to determine these upscale allowable values for the RBM.

ITS 5.6.5 also requires that the core operating limits shall be determined such that all applicable limits of the CNS safety analysis are met.

The upscale allowable values for the tnree power ranges of the RBM are core operating limits which can be specified in the COLR. The staff concludes that the NRC-approved analytical  ;

method to determine these allowable values is in the above document that the licensee has proposed to also list in the COLR. Therefore, as required by ITS 5.6.5, the RBM upscale allowable values for CNS will be determined correctly such that all applicable limits of the CNS safety analysis are met.

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68-Based on the above, the staff concludes that the proposed change is acceptable.

4. ITS 3.3.3.2 CTS Tables 3.2.l-1 and 4.2.1. Relocate the List of Alternate Shutdown Instrumentation and the Minimum Number of Channels for Each Instrument to the IST Bases (3.3.3.2 - RL.1)

The CTS and the STS each contain a table of the attemate (or mmote) shutdown instrumentation and the minimum number of channels required. The licensee has proposed to relocate this table to the ITS Bases where the information could only be changed in accordance with the change criteria in 10 CFR 50.59, as required in ITS 5.5.9, " Technical Specifications (TS) Bases Control Program." For CTS Table 4.2.1 only the list of instrumentation will be relocated. The deletion of the readout location of the instrumentation in CTS Table 3.2.1-1 is addressed as change "3.3.3.2 L.4"in the Table of Less Restrictive Changes. The relocation of CTS Table 3.2.1-2,"Altemate Sh.utdown Controls,"is addressed as change *3.3.3.2 LA.1"in the Table of Relocated Details.

In the conversion of the CTS to the ITS, there are many requirements in the CTS that are being relocated to licensee-controlled documents. These are addressed in Sections Ill.D and Ill.E. of this SE for the ITS conversion. Licensee-controlled documents are documents, such as the USAR, quality assurance plan, physical security plan, and emergency preparedness plan.

Changes to these documents are controlled by 10 CFR 50.59 or 10 CFR 50.54, and changes by the licensee to commitments in these documents may be made without prior staff approval if the changes meet the criteria in the appropriate regulation. The ITS Bases is also considered a licensee-controlled document because ITS 5.5.10 provides criteria whereby licensees may make changes to the Bases and this criteria is the 10 CFR 50.59 criteria.

The proposed change is Type 4, performance requirements for indication-only instrumentation j and alarms, of relocated CTS requirement, as discussed in Section lil.D of this SE for the ITS 4 conversion. Indication-only instrumentation, test equipment, and alarms are usually not required to be operable to support TS operability of a system or component unless these items are included in the TS as Accident Monitoring instrumentation. Thus, with the exception of the Accident Monitoring instrumentation, the STS do no include operability requirements for indication-only instrumentation and alarms. The availability of such indication-only instruments, l rnonitoring instruments, and alarms, and the necessary compensatory activities if they are not available, are more appropriately specified in station operational, maintenance, and  !

annunciator- response procedures required by ITS 5.4.1. Removal of requirements for indication-only instrumentation and alarms from the CTS is acceptable because they will be adequately controlled in station procedures.

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l The details in Tables 3.2.1-1 and 4.2.1 are not necessary to ensure the attemate shutdown l system is operable. The requirements of ITS 3.3.3.2 which require that the system is operable and the definition of operability in the ITS are sufficient. Therefore, the details in these tables l on the list of instrumentation and the minimum number of channels required are not required to be in the ITS. Changes to these requirements will be adequately controlled through ITS 5.5.10 in that changes will be required to be in accordance with 10 CFR 50.59.

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j Based on the above, the staff concludes that the proposed change is acceptable.

5. ITS 3.5.1 CTS 3.5 A. 3.5.C. and 3 5.E. Core Sorav and LPCI Svstems. HPCI Svstem. and Automatic Deoressurization System. Add Actions B and H to Allow Combinations of Low-Pressure and Hiah-Pressure ECCS Subavstems to be Inoperable for 72 Hours (ITS 3.5.1 - L.18)

The CTS and the STS do not contain any actions for the following two combinations of inoperable low-pressure and high-pressure ECCS for Modes 1,2, and 3: (1) one LPCI subsystem and one low pressure CSS subsystem are inoperable, and (2) one automatic depressurization system (ADS) valve and the HCPI system are inoperable. The licensee has proposed a separate Action for each combination that would allow each one for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The proposed change is different from that required in the CTS and the STS in that the CTS and the STS would require immediate action through LCO 3.0.3 to begin station shutdown.

The licensee stated that the above two combinations of inoperable low-pressure and high-pressure ECCS are not outside the design basis for the CNS. The station safety analyses are summarized in NEDC-32687P, ' Cooper Nuclear Station SAFER /GESTR-LOCA Loss-of-Coolant Accident," Revision, dated March 1997. This document was approved by the staff in the letter dated September 23,1997.

The ECCS at CNS are one HPCI system, six ADS relief valves, two CS subsystems, and two LPCI subsystems. Each LDCI subsystem has two 100% pumps. The CS and LPCI are for large breaks when the RCS would be at low pressure, and the HCPI and ADS valves are for small breaks when the RCS would be at high pressure (until the ADS operated).

For the low pressure ECCS, the licensee stated that the approved NEDC document shows that adequate core cooling is provided by two LPCI subsystems, with up to one pump inoperable in each subsystem, and one CS subsystem. This is for a large break in the recirculation loop discharge line. The largest pipe break for the RCS is the suction of the recirculation loop l because of the effective break size and its low elevation relative to the core; however, one LCPI

. would discharge through the break if the break was on the discharge side of the recirculation loop. Table VI-5-3 of the USAR states the PCT is highest (2200 degrees F) for the recirculation loop discharge line. Therefore, the most limiting large break is the recirculation loop discharge line and the peak clad temperature (PCT), for this break size and the available ECCS, does not exceed the 2200 degrees F ECCS acceptance criteria in 10 CFR 50.46.

l For the high pressure systems, the licensee stated that the approved NEDC document shows that adequate core cooling is provided by 5 ADS valves, and the LPCI and CS subsystems.

! For this condition, the resulting PCT also does not exceed the ECCS allowable criteria. The  !

small break region is defined as that portion of the line break spectrum where failure of HPCI is the most limiting failure. If a small break occurs, the RCS vessel depressurizes slowly or not at all. For CNS, USAR Section VI-5.3.1.1 states that the limiting small breaks were a 0.15 square foot break in the recirculation loop discharge piping and a 0.07 square foot break in the recirculation loop suction piping.

Cooper Nuclear Station Draft Safety Evaluation i

The proposed Action statements are for combinations of inoperable (1) low-pressure and (2) high-pressure ECCS where, for the most limiting break, there would be sufficient ECCS flow with the remaining operable ECCS to meet the ECCS acceptance criteria. However, with the inoperable ECCS, the operable ECCS subsystems and pumps would have been reduced to where any additional single active failure would result in the PCT exceeding the ECCS acceptance criteria for the most limiting break. The proposed 72-hour allowable outage time (AOT), for one LPCI pump inoperable in one or both LPCI subsystems or one CS subsystem inoperable, is consistent with the AOT for one high-pressure ECCS system and one low-pressure ECCS system in STS 3.5.1. It is also consistent with the recommendations of a reliability study (i.e., NRC memorandum from R.L. Baer to V. Stello, Jr., " Recommended interim Revisions to LCOs for ECCS Components," dated December 1,1975). The 72-hour AOT will reduce the chance of unnecessary shutdown transients for CNS.

After reviewing the proposed change, the staff concludes that it needs more time to complete its review because of the generic implications of the change and the impact of a decrease in defense-in-depth because of the change on the overall risk of the plant. Therefore, the staff requests that the licensee withdraw this proposed change and resubmit it after the proposed

- ITS have been approved.

6. ITS 5.5.9 CTS 4 9.A.2.d and e. Diesel Fuel Oil Testina Proaram. Addition of a New Test That Could Reolace the Clear and Brioht Annenrance Test in CTS 4.9.A.2.e.1.d)

(ITS 5.5 - M.4)

The CTS and the STS do not contain a test of the diesel generator fuel oil whereby the water and sediment content is within limits. The licensee has proposed to add this test as a -

replacement test to the clear and bright appearance with proper color test if the oil is too dark.

The licensee uses the ASTM-D4176-1991 clear and bright test to provide a qualitative assessment of the acceptability of new diesel fuel oil with regard to water and sediment content.

The ASTM clear and bright test is a visual check for evidence of water and particulate contamination of the fuel oil. However, this test should only be used for fuel oil meeting the color requirements of the standard (i.e., ASTM color 5 or less) because darker fuel oil will obscure the presence of water or particulate. The intentional addition of dyes to fuel oil by suppliers (such as to identify sulphur coatent) can make the fuel oil darker than ASTM 5. To address the method of determining the presence of water and particulate in these darker fuel oils, the licensee proposed to allow the use of ASTM-D9751989a water and sediment by centrifuge test in place of the CTS test. The Bases of ITS 3.8.3 states this information and can only be changed by ITS 5.5.10 (i.e., by following the change criteria in 10 CFR 50.59).

The use of ASTM-D975-1989a for fuel oils darker than ASTM 5 will ensure that these darker fuel oils have water and particulate levels within acceptable values. For these darker fuel oils, the sensitivity for detecting water and particulate in the fuel oil is better with the proposed standard. Therefore, the proposed change is to add an additional requirement to account for a deficiency in an existing standard used in the CTS.

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The licensee has stated that the centrifugal test for the determination of water and particulate content of the fuel oil was specified in ASTM-Dg75-1989a. The staff believes the correct ASTM standard to be referenced in the ITS is ASTM-D1796-83. The former ASTM-D975-1989a only references ASTM-D1796-83, and the latter standard contains the actual centrifugal test j requirements. The licensee is requested to revise its submittals to specify ASTM-D1796-83 in j the ITS. With this revision, the staff concludes that the proposed change is acceptable.

7. ITS 5.7 CTS 6.3.4.a. Reduce the Hiah Radiation Area Access Controls to Be Cori=3= tant with the Current Requirements in 10 CFR 20.106(a) and (b) (ITS 5.7 L.1)

The CTS Subsection 6.3.4.A and STS Section 5.7 have the requirements on the access to high radiation areas. The licensee has proposed changes that reduce the requirements in CTS 6.3,4.A on this access. In accordance with 10 CFR 20.1601(c), the changes are proposed as alternatives to the requirements in 10 CFR 20.1601 and are based on recent significant revisions to 10 CFR 20.1601 on control of access to high radiation areas, guidance provided in RG 8.38, " Control of Access to High and Very high Radiation Areas in Nuclear Power Plants,"

and current industry technology in controlling such access. The revisions to CTS 6.3.4.A includes a capping dose rate to differentiate a high radiation area from a very high radiation area, additional requirements for groups entering high radiation areas, and clarification of the need for control of workers in high radiation areas. The overall change is less restrictive because of the increased flexibility afforded by the proposed ITS Section 5.7 in comparison to CTS 6.3.4.A. The proposed ITS Section 5.7 'is different from the STS Section 5.7. The staff has reviewed the proposed ITS 5.7 and concluded that it is acceptable except for the following substantive differences:  !

1. Proposed paragraph ITS 5.7.1.4.b does not end with the phrase, "and with the means to communicate with individuals in the area who are covered by such surveillance."

2. Proposed paragraph ITS 5.7.2.a does not include the requirement that the guard is

" continuously" present at an unlocked entryway to prevent unauthorized entry.

Therefore, the licensee has proposed requirements that are different from the CTS and STS.

In its letter of May 6,1998, the licensee addressed the aoove differences and stated that (1) the first difference from the TSTF is justified by the CNS plant-specific design and current administrative practice in accordance with RG 8.38, and (2) the second difference is justified because the word " continuously" is not needed in ITS 5.7.2.a to require that, if an entryway to a high radiation area does not have a locked door or gate, there must be a guard continuously present at the entryway to prevent authorized entry. The licensee concluded that the revised l-requirements, including the above differences, will not decrease the ability to provide control of exposures from extemal sources in restricted areas.

The staff does not agree with the licensee explanation that the above differences are justifed l by the CNS plant-specific design, the current administrative practice in accordance with RG 8.38, and that the word " continuously" is not needed. The licensee must provide the details, on Cooper Nuclear Station Draft Safety Evaluation L----_-___------------__----_--.-

I l e how the CNS plant-specific design and the administrative practices in RG 8.38 Justify the first difference, for the staff to review. Without these details, the staff does not agree that the first .

difference is not important; however, it is stated in RG 8.38 in Section 2.4, " Alternative Methods l for Access Control," for workers in a high radiation area under continuous closed circuit TV l surveillance by personnel qualified in radiation protection procedures, that the qualified personnel must be able to provide positive exposure control over the activities being performed in the area. This means that there would be the capability to communicate with the individuals ,

in the area.

For the second difference, the staff concludsc that if the word " continuously" is not present in ITS 5.7.2.a, then the unlocked entryway is not required to be continuously guarded and 10 CFR 20.1602(b) requires that entryways to high radiation areas that are unlocked (except when access is required) have continuous direct or electronic surveillance that is capable of preventing unauthorized entry. The word " continuously" is needed in ITS 5.7.2.a.

Therefore, The staff concludes that this proposed change is not acceptable. j l'

IV. COMMITMENTS RELIED UPON I

in reviewing the proposed ITS conversion for the CNS, the staff has relied upon the licensee commitment to relocate certain requirements from the CTS to licensee-controlled documents as described in Table RL of Details Relocated from Current Technical Specifications and the l Table R of Relocated Current Technical Specifications attached to this SE. These tables reflect l the relocations described in the licensee's submittals on the conversion. The licensee has been l requested to submit a license condition to make this commitment enforceable. Such a  !

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

V. LICENSE CONDITIONS in its application of March 27,1997, the licensee discussed the problems with the first performance of the SRs in the ITS that will be new or revised compared to the SRs in the CTS, The licensee has been requested to submit a license condition to define the schedule to begin performing the new and revised SRs during or after the implementation of the ITS. This schedule should be the following:

. For SRs that are new in this amendment, the first performance is due at the end of the l 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 l surveillance performed after implementation of this amendment.

Cooper Nuclear Station Draft Safety Evaluation

t 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 date of implementation of this amendment.

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

The staff has reviewed the above schedule for the licensee to begin performing the new and revised SRs and concludes that it is an acceptable schedule.

VI. STATE CONSULTATION in accordance with the Commission's regulations, the Nebraska State official was notified of the l proposed issuance of the ITS conversion amendment for the CNS. The State official had no comments.

Vll. 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 March 17,1998 (63 FR 13074) for the proposed conversion from the CTS to the ITS for the CNS. 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.

Included in these amendments are changes that were beyond the scope of the ITS conversion for the CNS. These changes are discussed in Section Ill.G of this safety evaluation and were either included in the Federal Register notice of March 17,1998 (63 FR 13074) or in separate notices in the Federal Register. These changes altered 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 these 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. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 19971,63 FR 25111, and 63 FR 25112). Accordingly, the amendment meets 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 issuance of the amendment.

Vill. CONCLUSION The CNS ITS provides clearer, more readily understa:idable requirements to ensure safer operation of the station. The NRC staff concludes that the ITS satisfy the guidance in the Cooper Nuclear Station Draft Safety Evaluation

I j Commission's Final Policy Statement with regard to the cc,ntent of TS, and conform to the STS l provided in NUREG-1433 with appropriate modifications for plant-specific considerations. The l NRC staff further concludes that the ITS satisfy Section 182a of the Atomic Energy Act,10 CFR l 50.36, and other applicable standards. On this basis, the NRC staff concludes that the proposed ITS 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 concludes that these changes are acceptable.

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

l (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 amendments will not be inimical to the common defense and security, or to the health and safety of the public.

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

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

3. Table L of Less Restrictive Change to Current Technical Specifications

4. Table RL of Details Relocated from Current Technical Specifications

5. Table R of Relocated Current Technical Specifications Principal Contributors C. Harbuck C. Shiraki R. Tjader R. Giardina T.Liu f

J. Foster J.Donohew Date:

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1 Cooper Nuclear Station Draft Safety Evaluation i

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APPENDIX C ADDITIONAL CONDITIONS l FACILITY OPERATING LICENSE NO. DPR-46 1

Nebraska Public Power District shall comply with the following conditions on the schedules noted below:

l Amendment implementation Number Additional Conditions Date This amendment authorizes the The amendment shall relocation of certain Technical be implemented by Specification requirements to [date).

l licensee-controlled documents.

Implementation of this amendment shallinclude the relocation of these technical specification requirements to the appropriate documents, as described in Table LR of Details Relocated from Current Technical Specifications and Table R of Relocated Current Technical Specifications that are attached to the NRC staffs Safety Evaluation enclosed with this amendment.

l ENCLOSURE 2 i

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