ML20216D005

Refers to License Amend Request Submitted by Bg&E on 961204 to Convert Calvert Cliffs Current TS to Improved Ts.Drfat SE W/Ogc Comments Encl.Nrc & Bg&E Will Meet on 980319 to Review Status of Draft SE Comments

ML20216D005
Person / Time
Site:	Calvert Cliffs
Issue date:	03/05/1998
From:	Dromerick A NRC (Affiliation Not Assigned)
To:	Cruse C BALTIMORE GAS & ELECTRIC CO.
References
TAC-M97363, TAC-M97364, NUDOCS 9803160258
Download: ML20216D005 (42)

Text

'b Mr. Ch rl:s H. Cruse March 5, 1998 l

Vice President - Nucle:r Energy B ltimore G:s and Electric CompIny.

Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657-4702 i

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING THE TECHNICAL SPECIFICATIONS CHANGE REQUEST TO CONVERT TO THE IMPROVED TECHNICAL SPECIFICATIONS FOR THE CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 (TAC NOS. M97363 AND M97364)

Dear Mr. Cruse:

1 On December 4,1996, Baltimore Gas and Electric Company (BGE), submitted a license amendment request to convert the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2 Current Technical Specifications to the Improved Technical Specifications. On January 29,1998, the Calvert Cliffs draft Safety Evaluation (SE) report was issued and simultaneously sent to BGE and NRC Office of General Counsel (OGC) for review. The OGC completed their review and their comments have been included in the draft SE as redline / strikeout comments. A copy is enclosed for your consideration.

On March 19,1998, BGE and NRC staff will meet at NRC to review the status of draft SE comments to bring the review comments to closure. Resolution of the comments will be reflected in the Final SE, which is anticipated to be issued on or about mid-April 1998. Should you have any questions, please do not hesitate to contact me at (301) 415-3473.

Sincerely, CRIGINAL SIGNED BY:

Alexander W. Dromerick, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318

Enclosure:

Request for Additional information cc w/ encl: See next page Distribution:

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.,g March 5, 1998 Mr. Charles H. Cruse Vice President-Nuclear Energy Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657-4702

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION (RAl) REGARDING THE TECHNICAL

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SPECIFICATIONS CHANGE REQUEST TO CONVERT TO THE IMPROVED TECHNICAL SPECIFICATIONS FOR THE CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND 2 (TAC NOS. M97363 AND M97364)

Dear Mr. Cruse:

On December 4,1996, Baltimore Gas and Electric Company (BGE), submitted a license amendment request to convert the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2 Cunent Technical Specifications to the improved Technical Specifications. On January 29,1998, the Calvert Cliffs draft Safety Evaluation (SE) report was issued and simultaneously sent to BGE and NRC Office of General Counsel (OGC) for review. The OGC completed their review and their comments have been included in the draft SE as redline / strikeout comments. A copy is enclosed for your consideration.

On March 19,190 SGE and NRC staff will meet at NRC to review the status of draft SE comments to bri ne review comments to closure. Resolution of the comments will be i

reflected in the Fs, il SE, which is anticipated to be issued on or about mid-April 1998. Should you have any questions, please do not hesitate to contact me at (301) 415-3473.

Sincerely, W

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Alexan er W. Dromerick, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/ll Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318

Enclosure:

Request for Additional Information cc w/ encl: See next page I

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Mr. Charles H. Cruse Calvert Cliffs Nuclear Power Plant Baltimore Gas and Electric Company Units Nos.1 and 2 cc:

1 President Mr. Joseph H. Walter, Chief Engineer Calvert County Board of Public Service Commission of Commissioners Maryland 175 Main Street Engineering Division Prince Frederick, MD 20678 6 St. Paul Centre Baltimore, MD 21202-6806 James P. Bennett, Esquire Counsel Kristen A. Burger, Esquire Baltimore Gas and Electric Company Maryland People's Counsel P.O. Box 1475 6 St. Paul Centre Baltimore, MD 21203 Suite 2102 Baltimore, MD 21202-1631 Jay E. Silberg, Esquire

. Shaw, Pittman, Potts, and Trowbridge Patricia T. Bimie, Esquire 2300 N Street NW Co-Director Washington, DC 20037 Maryland Safe Energy Coalition P.O. Box 33111 Mr. Thomas N. Pritchett, Director Baltimore, MD 21218 NRM Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatelt 1650 Calvert Cliffs Parkway NRC Technical Training Center Lusby, MD 20657-4702 5700 Brainerd Road

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Chattanooga, TN 37411-4017 Resident inspector U.S. Nuclear' Regulatory Commission P.O. Box 287 St. Leonard, MD 20685 Mr. Richard I. McLean, Manager Nuclear Programs Power Plant Research Program

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Maryland Dept. of Natural Resources Tawes State Office Building, B3 Annapolis, MD 21401 l

Regional Administrator, Region I U.S. Nuclear Regulatory Commission l.

475 Allendale Road King of Prussia, PA 19406 l'

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t DRAFT Revised: 3/9/98 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. XXX TO FACILITY OPERATING LICENSE DPR-53 AND AMENDMENT NO. XXX TO OPERATING LICENSE DPR-69 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 AND 2 BALTIMORE GAS AND ELECTRIC ENERGY COMPANY DOCKET NOS. 50-317 AND 50-318 i

1. INTRODUCTION Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2 (CCNPP) has been operating with Technical Specifications (TS) issued with the original operating licenses on July 31,1974 for j

Unit 1 and November 30,1976 for Unit 2, as amended from time to time. By letter dated December 4,1996, as supplemented by letters dated March 27, June 9, June 18, July 21, August 14, August 19, September 10, Septerr,ber 23, October 6, October 20, October 23, November 5,1997, January 12, and January 28,1998, Baltimore Gas and Electric Energy i

Company (the licensee) proposed to amend Appendix A of Operating License Nos. DPR-53 and DRP-69 to completely revise the CCNPP TS. The proposed amendment was based upon NUREG-1432, " Standard Technical Specifications for Combustion Engineering Plants,"

Revision 1, dated April 1995, and upon guidance in the "NRC Final Policy Statement on Technical SpecMication 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 completely the existing TS for CCNPP.

Hereinafter, the proposed TS are referred to as the improved TS (iTS), the existing CCNPP TS are referred to as the current TS (CTS), and the TS in NUREG-1432 are referred to as the standard TS (STS). The corresponding TS Bases are ITS Bases, CTS Bases, and STS Bases, respectively, in addition to basing ITS on STS, the Final Policy Statement, and 10 CFR 50.36, the licensee t

retained portions of the CTS as a basis for the iTS. Plant-specific issues, including design Calvert Cliffs Unit Nos.1 and 2

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i DRAFT Revised: 3/9/98 features, requirements, and operating practices, were discussed with the licensee during a series of conference calls and meetings that concluded on January 26,1998. Based on these discussions, the tiensee proposed matters of a generic nature that were not in STS. The NRC staff requested that the licensee submit such generic issues as a proposed change to STS through the Nuclear Energy Institute's Technical Specifications Task Force (TSTF). These generic issues were considered for specific applications in the CCNPP ITS. Consistent with the Final Policy Statement, the licensee proposed transferring some CTS requirements to licensee-controlled documents. In addition, human factors principles were emphasized to add clarity to the CTS requirements being retained in the ITS and to define more clearly the appropriate scope of the ITS. Further, significant changes were proposed to the CTS Bases to make each ITS requirement clearer and easier to understand.

The Commission's proposed action on the CCNPP application for an amendment dated December 4,1996, was published in the Feders/ Register on January 31,1997 (62 FR 4816).

The Staff's evaluation of the application, including supplements to the licensee's ITS proposal, submitted by letters dated March 27, June 9, June 18, July 21, August 14, August 19, September 10, Ocptcmbcr 20, October 6, October 20, October 23, November 5,1997, January 12, and January 28,1998, that resulted from NRC requests for information and discussions with the licensee during the NRC staff review, is presented in this Safety Evaluation (SE). 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 scope of the action described in the federa/ Register notice.

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 CCNPP 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. The NRC staff approves the licensee's changes to the CTS with modifications documented in the revised submittals.

For the reasons stated infra in this SE, the NRC staff finds that the TS 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:

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[S]uch technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization... of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and the mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS "those items that are directly 1

related to maintaining the integrity of the physical barriers designed to contain radicactivity."

I 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 I

include 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 (SR); (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 a1d the NRC staff developed improved standard technical specifications that would establish models of the Commission's policy for each primary reactor type. In addition, the NRC staff, licensees, and Owners Groups developed generic administrative and editorial guidelines in the form of a

" Writer's Guide" for preparing 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-1432, which was developed using the guidance and criteria contained in the Commission's interim policy statement. STS were established as a model for developing improved TS for Combustion Engineering plants in general. 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 concerning various drafts of STS, so that the application of the TS criteria and the Writer's Guide would consistently reflect detailed system configurations and operating characteristics for all NSSS designs. As such, the generic Bases presented in NUREG-1432 provide an abundance of information regarding the extent to which the STS present requirements that are necessary to protect public health and safety.

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y DRAFT Revised: 3/9/98 On July 22,1993, the Commission issued its Final Policy Statement, expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36 (58 FR 39132). The Final Policy Statement described the safety benefits of the improved STS, and encouraged licensees to use the improved STS as the basis for plant-specific TS amendments, and for complete conversions to improved STS. Further, the Final Policy Statement gave guidance for evaluating the required scope of the TS and defined the guidance criteria to be used in determining which of the LCOs and associated surveillances should remain in the TS. The Commission noted that, in allowing certain items to be

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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 Port /and GeneralE/ectric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). There, the Appeal Board observed:

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

operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

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

Criterion 1 i

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

Criterion 2 i

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 Calvert Cliffs Unit Nos.1 and 2 L

. DRAFT Revised: 3/9/98 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

Part lli of this SE explains the NRC staff conclusion that the conversion of the CCNPP CTS to those based on STS, as modified by plant-specific changet, is consistent with the CCNPP current licensing basis and the requirements and Wdance of the Final Policy Statement and 10 CFR 50.36.

Ill.

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

In addition to the initial submittal of December 4,1996, as supplemented, the NRC staff review identified the need for clarifications and additions to the submittal in order to establish an appropriate regulatory basis for translation of current TS requirements into ITS. Each change proposed in the amendment request is identified as either a discussion of change (DOC) to CTS or a justification for deviation from STS. The NRC staff comments were documented as requests for additional information (RAls) and forwarded to the licensee for response by letters dated March 27, May 7, May 16, May 29, June 6, June 11, October 22, and November 7,1997. The licensee provided written responses to the NRC staff requests in letters dated March 27, June 9, June 18, July 21, August 14, August 19, September 10, September-24 October 6, October 20, November 5,1997, January 12, and January 28, 1998. The docketed letters clarified and revised the licensee basis for translating C TS requirements into ITS. The NRC staff finds that the licensee's submittats provide sufficient detail to allow the staff to reach a conclusion regarding the adequacy of the licensee's proposed changes.

The license amendment application was organized such that changes were included in each of the following CTS change categories, as appropriate: administrative changes, technical changes - less restrictive (specific), technical changes -less restrictive (generic), technical changes - more restrictive, and relocated specifications.

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Administrative Changes, (A), i.e., non-technical changes in the presentation of existing requirements; l

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Technical Changes - More Restrictive, (M), i.e., new or additional CTS requirements; (3)

Technical Changes - Less Restrictive (specific), (L), i.e., changes, deletions and relaxations of existing TS requirements; (4)

Technical Changes Less Restrictive (generic), (LA), i.e., deletion of existing TS requirements by movement of information and requirements from existing i

specifications (that are otherwise being retained) to licensee-controlled documents, including TS Bases; and (5)

Relocated Specifications, (R1), i.e., relaxations in which whole specifications (the LCO and associated action and SR) are removed from the existing TS (an NRC-controlled document) and placed in licensee-controlled documents.

These general categories of changes to the licensee's current TS 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 i

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 STS as guidance to reformat and make other administrative changes. Among the changes proposed by the li:~nsee and found acceptable by the NRC staff are:

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

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

(3). changing the wording of specification titles in STS to conform to existing plant practices (4) splitting up requirements currently grouped under a single current specification to more appropriate locations in two or more specifications of ITS i

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(5) combining related requirements currently presented in separate specifications of the CTS into a single specification of ITS.

Table A lists the administrative changes proposed in iTS. Table A is organized by the corresponding ITS section discussion of change, and provides a summary description of the administrative change that was made, and CTS and ITS LCO references. The NRC staff l

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 substantive change in operating requirements and are consistent with the Commission's regulations.

B. Technical Changes - More Restrictive The licensee, in electing to implement the specifications of STS proposed a number of requirements more restrictive than those in the CTS. 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 STS.

I Examples of more restrictive requirements are placing an LCO on plant equipment which is not required by the CTS to be operable, more restrictive regarements to restore inoperable i

equipment, and more restrictive SRs. Table M lists all the more restrictive changes proposed in ITS. Table M is organized by the corresponding ITS section discussion of change and provides a summary description of the more restrictive change that was adopted, and CTS i

and ITS LCO references. These changes are additional restrictions on plant operation that enhance safety and are acceptable.

C. Technical Changes - Less Restrictive (Specific)

Less restrictive requirements include changes, deletions and relaxations to portions of current TS requirunents that are not being retained in ITS. When requirements have been shown to give little or no safety benefit, their removal from the TS may be appropriate. In most cases, 1

relaxations previously granted to individual plants on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the Owners Groups comments on STS. The NRC staff reviewed generic relaxations contained in the STS and found them acceptable because they are consistent with current licensing practices and the Commission's regulations. The CCNPP design was also reviewed to determine if the specific design basis and licensing basis are consistent with the technical basis for the model requirements in STS, and thus provide a basis for ITS.

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A significant number of changes to the CTS involved changes, deletions and relaxations to

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portions of current TS requirements evaluated as Categories I through Vill that follow:

Category 1 - Relaxation of Applicability Calvert Cliffs Unit Nos.1 and 2 m

w DRAFT Revised: 3/9/98 Category II - Relaxation of Surveillance Frequency Category lli - Relaxation of Allowed Outage Time Category IV - Relaxation of Required Actions

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Category V - Relaxation of Surveillance Requirement Acceptance Criteria Category VI-Deletion of Requirement for 30-day Special Report to NRC Category Vil-Relaxation of LCO Category Vill - Deletion of SR The following discussions address why various technical specifications within each of 1:.v eight categories of information or specific requirements are not required to be included in ITS.

Relaxation of Acolicability (Category //

Reactor operating conditions are used in CTS to define when the LCO features are required to be operable. CTS applicebilities can be specific defined terms of reactor conditions: hot shutdown, cold shutdown, reactor critical or power operating condition. Applicabilities can also be more general. Depending on the circumstances, CTS may require that the LCO be maintained within limits in "all modes" or "any operating mode." Generalized applicability conditions are not containec' in STS, therefore ITS eliminate CTS requirements such as "all modes" cr "any operating mode," replacing them with ITS defined modes or applicable

- conditions that are consistent with the application of the plant safety analysis assumptions for operability of the required features.

In another application of this type of change, CTS requirernents may be eliminated during l

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 indeterm!nant in daten-Maant or which are inconsistent with application of accident analyses assumptions is acceptable because when LCOs cannot be met, the TS are i

satisfied by exiting the applicability thus taking the plant out of the conditions that require I

the safety sysan to be operable. These changes are consistent with STS and changes specified as Category I are accertable.

Relaxation of Surveillance Frecuencv (Category ///

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 due to test which also increases h

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~ DRAFT Revised: 3/9/98 equipment availability. In general, the STS contain test frequencies that are consistent with industry practice or industry standards for achieving acceptable levels of equipment

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reliability. Adopting testing practices specified in the STS is acceptable based on similar design, like-component testing for the system application and the availability of other TS requirements which provide regular checks to ensure limits are met.

Reduced testing can result in a safety enhancement because the unavailability due to test is reduced; in turn, reliability oi the affected structure, system or component should remain constant or increase. Reduced testing is acceptablo where operating experience, industry practice or the industry standards such as manufacturers' recommendations have shown that these components usually pass the Surveillance when performed at the specified

nterval, thus the frequency is acceptable from a reliability standpoint. Surveillance I

frequency changes to incorporate alternate train testing have been shown to be acceptable where other qualitative or quantitative test requirements are required which are established predictors of system performance, e.g., the CTS Current Technl cal Spccifications require a battery service test and battery charger to be demonstrated capable of recharging the battury every 18 months. The ITS nog,-, munical Spccifications will require performance every 24 months. This change decreases the Frequency of a charger test and service test from 18 to 24 months. The Frequency is acceptable g;ven the unit Conditions required to perfmm the test, and the other administrative controls existing to ensure adequate charger performance during the 24-month intervals. This Frequency is consistent with the refueling interval and is consistent with recommendations of Regulatory Guides (RG)1.32 and 1.129 that the battery service test should be performed during refueling operations, or some other outage. This is consistent with IEEE 450.

Additionally, surveillance frequency extension 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. These changes are consistent with STS and changes specified as Category 11 are acceptable.

1 Rdaxation of Allowed Outaae Time (Category ////

Upon discovery of a failure to meet an LCO, STS specify times for completing reauired actions of the associated TS conditions. Required actions of the associated conditions are i

used to establish rernedial rneasures that must be taken within specified completion times (allowed outage times). These times define limits during which operation in a degraded condition is permitted.

Adopting completion times from the STS is acceptable because completion times take into account the operability status of the redundant systems of TS required features, the capacity and capabillty of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a design basis; accident (DBA) occurring Calvert Cliffs Unit Nos.1 and 2

10-DRAFT Revised: 3/9/98 during the repair period. These changes are consistent with STS and allowed outage time extensions specified as Category lli are acceptable.

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Relaxation of Reauired Actions (Category /V)

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i CTS provides lists of acceptable devices that may be used to satisfy LCO requirements.

1 The iTS reflect the STS approach to provide LCO requirements that specify the protective j

limit that is required to meet safety analysis assumptions for required features. The I

protective limits replace the lists of specific devices previously found to be acceptable to the NRC staff for meeting the LCO. The ITS changes provide the same degree of 1

protection required by the safety analysis and provide flexibility for meeting limits without adversely affecting operations since equivalent features are required to be operable. These changes are consistent with STS and chang 6s specified as Category IV are acceptable.

1 Relaxation of Surveillance Reauirement Accentance Criteria /Cateaorv td CTS require safety systems to be tested and verified operable prior to entering applicable conditions. ITS provide the additional requirement to verify operability by actual or test conditions. Adopting the STS allowance for " actual" conditions is acceptable because TS required features cannot distinguish between an " actual" signal or a " test" signal, i

Category V also includes changes to CTS requirements that are replaced in the ITS with separate and distinct testing requirements which when combined include operability l

verification of all TS required components for the features specified in the CTS. Adopting this format preference in the STS is acceptable because TS SRs that remain include testing of all prev ous features required to be verified operable. These changes are consistent with STS and changes specified as Category V are acceptable.

Deletion of Reouirement for 30 Dav Soecial Reoort to NRC (Category V//

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

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

1 heyxation of LCO (Category Vil) 2/19/98;t.icensee requested to expand " Relaxation of LCO" Calvert Cliffs Unit Nos.1 and 2

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Revised: 3/9/98 CTS provides LCO requirements. Tha ITS reflect the STS approach to provide LCO requirements that specify the protective limit that is required to meet safety analysis assumptions ~ for required features. The protective limits replace the lists of specific devices previously found to be acceptable to the NRC steff for meeting the LCO. The ITS changes provide the same degree of protection required by the safety ar:alysis and provide flexibility l

for meeting limits without adversely affecting operations since equivalent features are l

required to be operable. These changes are consistent with STS and changes specified as Category Vll are acceptable.

Deletbn of Surveillance Reauirements (SR) (Category V///)

CTS require safety systems to be tested and verified operable prior to entering applicable conditions. The ITS reflect STS required surveillance requirements, eliminating unnecessary CTS surveillance requirements that do not contribute to verification that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues tc, ce tested. These changes are consistent with STS and changes specified as Category Vlli are acceptable.

i Table L lists all the less restrictive changes proposed in the ITS. Table L is orgm.; zed by the corresponding ITS specificcion discussion of change and provides a summary description of the less restrictive change that was adopted, CTS and ITS reference, and category of change.

Additionally, in electing to implement the specifications of STS, the licensee also proposed a number of less restrictive changes to the CTS which do not apply to the above Categ0 ries of changes, deletions and relaxations of CTS requirements. These changes are discussed below.

The associated discussion of change identifier (e.g., L1) is provided for these unique less restrictive changes.

l Section 1.0 - Less Restrictive L1 The CTS defines Core Alteration as "the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel." The words "any component" are replaced by the words "any fuel, sources, or reactivity control components." The CTS and ITS requirements that use the defined term (Core Alteration; are specified to protect against or mitigate reactivity excursion events. The movement of components other than fuel, sources, or reactivity control components has, at most, a negligible effect on core reactivity. Thus, the ITS definition appropriately restricts a Core Alteration to movement of components that could affect core reactivity.

Removal of the CTS restriction on the movement of components other than fuel, sources, and reactivity control components does not reduce CTS requirements that control changes in core reactiuty, and thus does not adversely affect plant safety.

l Therefore, the ITS definition oi " Core Afteration" is acceptable.

I Calvert Cliffs Unit Nos.1 and 2

c~

l 1

l l

l l DRAFT Revised: 3/9/98 L2 The CTS definition of " Shutdown Margin (SDM) requires the calculation of SDM to l

account for the single control element assembly (CEA) of highest worth being fully withdrawn. The ITS definition of SDM is relaxed to allow not accounting for the single CEA of highest worth being fully withdrawn if 611 CEAs are verified fully inserted by two independent means. This change is acceptable because requiring the CEA of higNst reactivity worth to be assumed withdrawn is overly conservative when information on actual CEA position can be obtained by two independent means.

i 1

L5 The CTS defines definition of " Operable - Operability" such that a requirca that system, i

subsystem, train, component, or device is operable or has operability if among other things, all necessary normal and emergency electrical power sources bc avsllsble for ths are capable of performing their related support functions. The ITS definition of

. Operability" is changed to depend on the capability of either normal ar_ emergency l

power, rather than both normal and emergency power. Thc lTS definition of "Operablc -

1 Operablllty" requires cl:hcr norma l cr smcrgency povver bc ava!labls. This is acceptable because, in the event of an inoperable electrical power source, the ITS action requirements for electrical power systems will (a) limit the time the plant is operated in this degraded condition, and (b) ensure that adequate measures (cross-train checks to ensure operability of the components, systems, etc. redundant required features) are taken so that any loss of function condition created by the inoperability is identified and specified action is taken.

Section 2.0 - Less Restrictive 1

L1 CTS Actions 2.1.1.b, c, d and 2.1.2.b, c, d require submitting a report and notifying the NRC Operations Center, the Vice Presider.t-Nuclear Energy, and the Offsite Review Function when a Safety Limit is exceeded vislsted. The 10 CFR.50.36 requires, in.part, that if any safety limit is exceeded, the reactor must be shut downf the licensee;shall notify the Commission, review the matter, and record the results of the reviewsThe licensee shall notify the Commission as required.by 10 CFR 50.'72 # 4 submit a, Licensee Event. Report,(LER) to the, Commission as requi. red by 1.0,CFR 50.73; ITS deletes these l

requirements which are duplicative and are contained in 10 CFR 50.36. This change is consistent with the STS and with the approved TSTF-5 change incorporated. This change deletes notification, reporting and restart requirements if a Safety Limit is i

violated.

Section 3.0 - Less Restrictive l

L1 CTS 3.0.4 and CTS 4.0.4 prohibit disallovi entry into a Mode or other specified condition in rcqu; red by the Applicability unless the LCO or SR, respectively, is sat.Jied (with I

certain some exceptions). Corresponding ITS LCO 3.0.4 and ITS SR 3.0.4 specify that this requirement only applies for entry into a Mode or other specified condition in the Applicability in Modes 1,2,3, and 4. This requirement is acceptable becausu the Calvert Cliffs Unit Nos.1 and 2

i t.

l DRAFT Revised: 3/9/98 l

l l

l specifications that apply in Modes 5 ar'd 6 contain adequate remedial measures allowing mode changes between Modes 5 and 6.

L2 CTS 3.0.5 provides that eHows systems, subsyctems, trains, components, or devices may.be. considered is remain Operable when either the normal or emergency power source is inoperables provided, This is only a!!cvsed, however, en the cond;;icas that (1) l one power source remains operable and (2) all redundant systems, components, subsystemsstrainsf and devices eterr are operable. If these requirements are not met within two hours, CTS,3.0.5l requires, in part,7that _ action be_ initiated to place the unit in a Mode in:which the. applicable LCO_no longer applies; a shutdovin is rcquired. These requirements conservatively ensure that appropriate actions will be taken in the event of a totalloss of a specified safety function caused in part by a loss of electrical power sources. ITS LCO 3.0.6 replaces CTS 3.0.5 and omits these explicit conditions and the 2-hour shutdown action requirement. ITS LCO 3.0.6 applies in the event a support system LCO is not met, and generally only requires completing the support system specification's action requirements, unless the evaluation it requires in accordance with l

ITS 5.5.15, " Safety Function Determination Program" identifies a loss of a safety function. In such a situation, ITS LCO 3.0.6 requires taking the action requirements of the LCO in which the loss cf safety function exists. The overall effect of adding ITS LCO 3.0.6 is that TS no longer require following (commonly referred.to as cascading) to supported system specification action requirements when a support system is inoperable. This change is acceptable because, in the ITS, the actions required to ensure the unit is maintained in a safe condition are contained in the support system i

specification. In some cases, however, the action requirements of a support system specification will direct completion of the action requirements of a supported system specification. Such action requirements constitute exceptions to ITS LCO 3.0.6.

L3 CTS 4.0.3 providescin part, that failure to perform an;SR.within theispecified frequency

^

l

, shall constitute noncompliance with the Operabi_lity requirements of th_e. associated Action. requirements, and that the associated Action requirementa apply at the, time that l

frilure.to satisfy the SR is identified. Further, CTSf3/4.0.3fpermits.the' delay,of,the-

)

applicat_lon of the:LCO. Action requirements for up;to 24. hours _tc allogcompletion of the l

allowed surveillance,' depending on the frequencynif_the~ allowed' surveillance.LCO l

outage time is less then.24 ho_urs.9 requirea s:ar lng the Ocmpic:lon Times of appiicabic ac len requacments if a survclllance ls missed for spcclflca;lcas vihcss Pcquired Action l

Comp letica Times are.1. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but allcres 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to compic:c th; survc llance. lf the Comp;etlon Times are < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the CTS allswa a 24-hour delay is perform the survelllance 'ueforc starting the Comp letica Times of the required Actions:

Corresponding ITS SR 3.0.3 modifies this provision by basing the delay time on the specified surveillance Frequency instead of on the Required Action Completion Time. It specifies the delay time as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the limit of the specified Frequency, whichever is less. It also delays the start of the Completion Times until after the delay period has expired, or sooner if the surveillance is failed. For CTS Completion Times 2 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Calvert Cliffs Unit Nos.1 and 2

l 1

l.

i.

(

l DRAFT 3

Revised: 3/9/98 this change is less restrictive because the Completion Time would not start until after l

expiration of the delay period, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For CTS Completion Times under 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, this change is more restrictive when the associated surveillance Frequency is under 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Whether more or less restrictive, this change is acceptable because the j

time required to perform the surveillance is usually brief, the safety significance of the i

delay in completing the surveillance is low, and the most probable result of any particular surveillance being performed is the verification of conformance with the LCO requirements.

L4 CTS 3.0.4 does not allow entry into an operational changmg Modes or other specified condition unless the provisions of the LCO are met without reliance on provisions contained in the Acdons. The iTS adds an exception to allow entry into a mode or other applicable condition with an LCO not being met, when the an action statement for the LCO allows oper.stion 5 that Mode for an unlimited period of time. ITS LCO 3.0.4 is ecceptable Decease compliance with required actions that permit continued operation of the unit for an unlimited period of time in a Mode or other specified condition provides an acceptable level of safety for continued operation.

Specification 3.3.7 - Less Restrictive l

L1 CTS 3.3.2.1 Table 3.3-3 Functional Unit 6.a requires, in part, two containment purge j

valve isolation manual trip channels per penetration to be Operable. ITS 3.3.7 reduces this requirement to one manual actuation channel per penetration. This change is acceptable because the capability to manually close the purge valves using the purge valve control switches is maintained with one channel Operable. Also, the containment purge and exhaust isolation occurs automatically on a high radiation signal or a containment isolation signal. Since only one manual actuation channel is required by ITS LCO 3.3.7, the requirements of Action 8 of CTS Table 3.3-3 for the condition of one of two channels inoperable are no longer needed. In the event the one required channelin a penetration is inoperable, and the other additional channel is also inoperable, ITS 3.3.7 Required Action B.1, would immediately requirc closing and maintaining closed the containment purge and exhaust valves, consistent with CTS Action 8. This action is acceptable because closing the containment purge and exhaust valves accomplishes the safety function of isolating the penetration. With the penetration isolated, the manual trip function is no longer needed.

i Specification 3.8.1 - Leo Restrictive L13 CCNPP has'a common sontrol room with_two. trains each;of Control Rooni Ernergency l

Ven_tilation System (CREVS). Control Room EmergencyTemperature. Systems (CRETS),

and _ Hydrogen' Analyzer (H Analyzer). One train o_f.CREVS, CRETS, and.HiAnalyzer is 2

powered from Unit 3, and the second train is powered from Unit'2? Both trains are required to be. Operable to support either unit at power; i.e., to_ support CREVS, CRETS, i

Calvert Cliffs Unit Nos.1 and 2 l

l

l J

1 i

- DRAFT i

Revised: 3/9/98 I

and H,*AnalyzerLiri Modes 1 through,4funit;1jrequired: alternating currentJA.C) power from Unit l27and Unit:2 r64uirep AC. power from_ Unit:1gThe CTS, definition of Operability; requires the:se?ystems to. have both'the normal and_e.mergency. power sources. OperspleRif the nonna.1 or emergency:AC powetsource to a CREVS, CRETS, or H,' Analyzer is inoperable; CI.%3,0.5_ allows.the.aff.ected system;to;be; considered

)

Operable provided the system'.s alternate. power source is Operable and the..re#mdant system.is OperableRW. both conditions are not met, CTS 3.0.5 requires;thatla plant

{

l shutdown bel commenced within two hours.lUnder_ CTS,Lan inoperable'oppositefunit AC.

j power source.for a CREVS, CRETS, H ' Analyzer: coincident;with an inoperable given unit 2

AC power source for afedundant CREVS, CRETS, or H Analyrer requires a plant 2

shutdown per 3.0.5CThis CTS scenario is too restrictivefand is changed in the lTS as follows. The CTC def ni lcn of Operablll:y reqwres norms: and cmcrgcacy AC povacr fw a Sys:cm to b2 cons ldcred Cperablc. Clnce certa:n equlpment accded is mcc: cac unl:'s accident-enalyrdsas-powered-from the other-unit's-AG-scurec:: 'CRETS, CRETS, cnd H, j

Analyzerl, thcsc AE sourcca arc a sc icqv; red to be Operablc. Wr,cn the AC scurces-ere incperablc, nc Acticas arc provided ln the CTC, thus the assoc sted train of CREVC, GRETSrand44, Analyzer are declared inoperab!c and the Actlcns of the indlvidual SpetH; cations are takca. In add l:icn, vehen a given unit's AC sourcc that pcvvers s

GREVSrGRETSrand44, Analyzer is a sc lnoperabic, CTC 0.06 vaculd icqu;rc an j

tmmediate shutdoven.

in the ITS, new Actions are being provided when the other unit's AC sources are i

inoperable. Actions A;and B require that the CREVS,' C.RETS, or H Analyzer be. declared 2

inoperable.after.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the given unit's normal or emergency!ACisource_is j

inoperable and_ the redundant CREVS, CRETS, or H, Analyzer is. inoperable'.NCondition C l

requires the same action ~when the opposite ~ unit's diesel. generator (DG) is inoperable 1

and requires.the.CREVS,~ CRETS, or H ' Analyzer _ supported by the. inoperable _DG,to'be 2

declared inoperable after.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,'regardless of thelredundant systern;statusMAc:lon A velll allcis 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> vshca the c hcr unl:'s offe :e circui ls noperable end Actlcn O ;slll allcvs 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> vshen the c her unl:'s 00 s lacperable. This is acceptable,'since when only one source is inoperable, the CREVS, CRETS, and H Analyzer can still perform their 2

safety function; they are still energized. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is also consistent with Regulatory Guide 1.93, which allows this same time to restore the AC source. If When the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> has expired, the associated CREVS, CRETS, and H 2

Analyzer will be declared inoperable, consistent with current requirements. Actions C and D will also include the similar compensatory measures as is required when a given unit's AC source is inoperable (i.e., performance of SR 3.8.1.1, SR 3.8.1.2, and SR 3.8.1.3 and cross-train checks). These compensatory measures will ensure an additional l

failure of an AC source will not go undetected. In addition, when both offsite sources or both DGs that supply power to the CREVS, CRETS, and H Analyzer are inoperable, the 2

ITS will allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore one of the offsite sources and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore one of the DGs te Operable status (second part of Conditions E and G). This also is consistent with the recommendations of Regulatory Guide 1.93 and the current Actions Calvert Cliffs Unit Nos.1 and 2

,. DRAFT Revised: 3/9/98 for the same situation for both like AC sources w 6he given unit. During this short time, the CREVS, CRET3, and H Analyzer are also energized, thus they can perform their 2

safety functions.

j Specification 3.8.5 - Less Restrictive L3 CTS Currcnt Tcchnicci Opccificctica 3.8.2.4 Action c contains the requirements when one required DC source is inoperable in Modes 5 and 6. The Actions require all containmer.t penetrations providing direct access from the containment atmosphere to

]

the outside atmosphere to be either closed by an isolation valve, blind flange, or manual valve, or be capable of being closed by an Operable automatic purge valve. The Actions also require a minimum of one door in each air lock to be closed, and the equipment door closed and held in place by a minimum of four bolts. ITS...y.m-

. -.nicc!

Opccificction 3.8.5 will not include this requirement. The concerns in Modes 5 and 6 are having the appropriate equipment available to mitigate accidents described in the accident analyses, and to cool the core. The Required Actions in CTS 3.8.5 Action a to j

suspend Core Alterations, movement of irradiated fuel, and positive reactivity additions mitigate the concerns of the potential accidents described in the accident analyses, and to cool the core. The Required Actions to suspend Core Alterations, movement of irradiated fuel, and positive reactivity additions mitigate the concerns of the potential accidents described in the accident analyses (fuel handling accident and boron dilution event). Other Technical Specifications exist (RCS Loops in Mode 5 and Shutdown Cooling requirements in Mode 6) that require the appropriate actions if core cooling is lost. Therefore, by following the Actions specified in contclncd voth the Direct Current (DC) Sources - Shutdown Technical Specification ITS 3.8.5, and other Technical Specifications that exist to ensure adequate core cooling, CTS 3.8.2.4 Action c can be deleted. The elimination of a Required Action constitutes a less restrictive change. This change is consistent with NUREG-1432.

Specification 3.8.9 - Less Restrictive L1 CTS Gerrent TechWMeecificstion 3.8.2.1 requires the AC busses buses to be Operable and energizt.v vom sources of power other than the DGs (i.e., from the offsite sources). ITS lmprovcd Tcchnicc: Spaclficstion 3.8.9 requires specifiedlAC)usses to be Operableibut will not require the AC busses buses to be energized from the offsite sources!; Rather, ITS 3.8.9 does not add.ress_the source Loi power to-the'specifiedfAC busses,'but it will allow the busses buses to be energized from any source (DG or offsite source!. The AC buses can perform their safety function as hes long as they are energized. They are not dependent on the offsite sources to perform their functions functlcn. With the AC buses energized from the DGs, they are fully capable o1 performing their safety functions function. The offsite sources are already required tc. be Operable by CTS 3.8.1.1 (TS 3.8.1), and actions are provided in ITS 3.8.1 if an offsite source !s inoperable, in addition, CTS 3.0.5 allows an offsite circuit to be inoperable ano' Calvert Cliffs Unit Nos.1 and 2

. DRAFT Revised: 3/9/98 does not require the supported system to be declared inoperable, provided a source of power is available to power the supported system and its redundant system is also Operable. These actions are also maintained in the ITS. Therefore, since the AC busses t

buses can continue to perform their safety functions function when powered from the DGs, this change is acceptable and is consistent with NUREG-1432.

4 Section 5.0 - Less Restrictive i

L1 CTS 6.2.2.d requires an individual qualified in radiation protection procedures to be on site ensite when fuel is in the reactor. ITS 5.5.2.d allows the position to be vacant for not more than two hours, in order to pro' vide for unexpected absence, provided immediate action is taken to fill the required position. This is a less restrictive i

requirement. This change is reasonable because it allows time to restore a required staffing position for unexpected absences without violating the TS Administrative Controls section, while ensuring the position is filled in a timely manner. Therefore, this change is acceptable and is consistent with NUREG-1432.

Table L lists all CTS requirements that have been relaxed and which pertain to Category 1 though Vlli and to the specific listing of changes discussed above. Table L is organized by ITS section and includes: the section designation, followed by the discussion of change identifier, e.g.,1.14-0 L1 (ITS Section 1.1, DOC L1); a summary description of the change; CTS and ITS LCO references; and a reference to the applicable change categories as discussed above (if applicable) and a "Chareetcsstion" of the discussica of change. Note to Calvert Cliffs Draft SE Reviegers: The " Characterization" column was deleted from CCNPP L Tables beccuse this column was used so infrecuentiv for CCNPP. Deletina this column allowed additional soace to exoand the Discussion column which was needed. Anv

_ Characterization" information in the Calvert Cliffs L Tables is caotured in the L Table Cateaorv colutnn.

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

D. Relocated Less Restrictive Requirements When requirements have been shown to give little or no safety benefit, their removal from the TS may be appropriate. In most cases, relaxations previously granted tc individual plants on a plant-specific basis were the result of (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements ar'd operating experience, or (3) resolution of the Owners Groups comments on STS. The NRC staff reviewed generic relaxations contained in STS and found them ucceptable because they are consistent with current licensing Calvert Cliffs Unit Nos.1 and 2 I

L

1 DRAFT Revised: 3/9/98 practices and the Commission's regulations. The CCNPP design was also reviewed to l

determine if the specific design basis and licensing basis are consistent with the technica!

basis for the model requirements in STS, and thus provide a basis for ITS. A significant number of changes to the CTS involved the removal of specific requirements and detailed information from individual specifications evaluated to be Types 1 through 4 -3 that follow:

Type 1 Details of System Design and System Description including Design Limits i

Type 2 Descriptions of System Operation Type 3 Procedural Details for TS Requirements and Related Reporting Problems Type 4? Relocation.of TS Administrative Requirements Redundant to Regulations Note to the Calvert Cliffs SE Reviewers: The Robinson "Tvoe 4 Performance Reauirements for Indication On!v instrumentation and Alarms" was not included in the CCNPP Draft SE because this cateaorv was not anolicable to CCNPP Relocated Less Restrictive Reauirements. The CCNPP "Tvoe 4 Relocatinn of TS Administrative Reauirements Redundant to Reaulations was included because these relocated less restrictive reauirements Tvoe 4 reauirements are anolicable to four Calvert Cliffs relocated Section 5.0 chanaes.

l The following discussions address why each of the four three types of information or specific requ:rements are not required to be included in ITS.

Details of System Desian and Svstem Descriotion includina Desian Limits (Type 1)

The design of the facility is required to be described in the UFSAR by 10 CFR 50,34. In addition, the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50 require that plant design be documented in r:ontrolled procedures and drawings, and maintained in accordance with an NRC-approved QA plan (reference in the UFSAR). In 10 CFR 50.59 controls are specified for changing the facility as describcd in the UFSAR, and in 10 CFR 50.54(a) criteria are specified for changing the QA plan in ITS, the Bases also contain descriptions of system design. ITS 5.5.14 specifies controls for changing the Bases.

Removing details of system design from the CTS is acceptable because this information will be adequately controlled in the UFSAR, controlled design documents and drawings or the TS Bases, as appropriate. Cycle-specific design limits are moved from the CTS to the Core Operating Limits Report (COLR) in accordance with Generic Letter (GL) 88-16. ITS L

Administrative Controls are revised to include the programmatic requirements for the COLR.

Descriotions of Systems Ooeration (Type 2)-

Calvert Cliffs Unit Nos.1 and 2

l i

1 DRAFT Revised: 3/9/98 The plans for the normal and emergency operation of the facility are required to be described in the UFSAR by 10 CFR 50.34. ITS 5.4.1.a requires written procedures to be established, implemented, and maintained for plant operating procedures including procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Controls specified in 10 CFR 50.59 apply to changes in procedures as described in the UFSAR. In ITS, the Bases aNo contain descriptions of system operation. It is acceptable to remove details of system operation from the TS because this type of information will be adequately controlled in the UFSAR, plant operating procedures, and the TS Bases, as appropriate.

Procedural Details for Meetina TS Recuirements & Related Reoortino Problems (Type 3)

Details for performing action and surveillance requirements are more appropriately specified in the plant procedures required by ITS 5.4.1, the UFSAR, and ITS Bases. For example, con + ol of the plant conditions appropriate to perform a surveillance test is an issue for procedures and scheduling and has previously been determined to be unnecessary as a TS restriction. As indicated in Generic Letter 91-04, allowing this procedural controlis consistent with the vast majority of other SRs that do not dictate plant conditions for surveillances. Prescriptive proceduralinformation in an action requirement is unlikely to contain all procedural considerations necessary for the plant operators to complete the 4

actions required, and referral to plant procedures is therefore required in any event. Other changes to procedural details include those associated with limits retained in the ITS. For example, the ITS requirement may refer to programmatic requirements such as COLR, included in ITS Section 5.6.5, which specifies the scope of the limits cor.tained in the COLR and mandates NRC approval of the analytical methodology.

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

cover the reports deemed to be necessary.

Relocation of TS Administrative Reauirements Redundant to Reaulation!(Type ~4)

Certain CCNPP, CTS administrative requirements were redundant to regulations _and were relocated from;the:TS.to thefSAR. or other lice.nsee; controlled documentsMbe Final Pol _ icy Statement and 10_CFR.50.36 allow licensees to voluntarily:use the criteria.in these l

regulations to_ reif :ste_ existing Technical Specifications that do not meet any;of the criteria i

to licensee-controlled documents. Changes,to_the facility'or to proceduresjdescribed in the l

FSAR are made in accordance with 10 CFR 50.59.S Changes _.madejn_ accord.ance with the provisions oflother licensee-controiled documents,(e.g.iOA plan, ODCM[are subject;to the specific requirements o.f those documents. For examplet CTS.6.2.2.e, requirements lof the Calvert Cliffs Unit Nos.1 and 2 I

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F I

t 1

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' DRAFT Revised: 3/9/98

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l fire brigade have been relocated to the Fire Protection Program'in the UFSAR,1which is controlled by,lTS'5.4.1.d and 10 CFR 50.59, and substantive fire protection.reouirements in11.0.CFR _5048 Appendix.R and approved exemptions. Therefore, relocation of the administrative details identified cbove,'is acceptable.

Table RL lbts CTS specifications and detailed information removed from individual specifications that are relocated to licensee-controlled documents in ITS. Table RL is organized by ITS section and includes: the section designation followed by the discussion of change identifier idsntified, e.g., 3.1.1 LA1 (ITS Section 3.1.1, DOC LA 1); CTS reference; a summary description of the change; the name of the document that retains the CTS requirements; the method for controlling future changes to relocated requirements; 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 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: (1) TS Bases controlled by ITS 5.5.14, " Technical Specifications Bases Control Program;" (2) UFSAR (includes the Technical Requirements Manual (TRM) by referenci.

trolled by 10 CFR 50.59; (3) the Offsite Dose Calculation Manual (ODCM) controlled oy n S 5.5.1; and (4) the QA plans as approved by the NRC and referenced in the UFSAR and controlled by 10 CFR Part 50, Appendix B. For each of these changes, Table RL also lists the licensee-controlled documents and the TS or regulatory requirements governing 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 Pclicy Statement (discussed in Part ll of this safety evaluation).

.Accordingly, existing detailed information and speci' requirements, such as generally described above, may be deleted from the CTS.

E.

Relocated Specifications The Final Policy Statement states that LCOs and associated requirements that do not satisfy or fall within any of the four specified criteria may be relocated from existing TS (an NRC-controlled document) to appropriate licensec-controlled documents. These requirements include the LCOs, Action Statements (ACTIONS), and associated SRs. In its application, the i

licensee proposed relocating such specifications to the Updated Final Safety Analysis Report (UFSAR) (includes the Technical Requirements Manual (TRM) by reference), and the ODCM, Calvert Cliffs Unit Nos.1 and 2

. DRAFT Revised: 3/9/98 l

l as appropnate. The staff has reviewed the licensee's submittals, and finds that relocation of

. these requirements to the UFSAR (and TRM) and ODCM is acceptable, in that changes to the l

UFSAR will be adequately controlled by 10 CFR 50.59 and changes to the ODCM will be controlled by ITS 5.5.1. These provisions will continue to be implemented by appropriate plant procedures; i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and. work control procedures.

The licensee, in electing to implement the specifications of STS, also proposed, in accordance with the criteria in the Final Policy Statement, to entirely remove certain TS from the CTS and place them in licensee-controlled documents noted in Table R. Table R lists all specifications-and specific CTS details that are ralocated, based on the Final Policy Statement, to licensee-controlled documents in ITS. Table R provides: a CTS reference; a summary description of the requirement; the name of the document that retains the CTS requirements; the method for L

controlling future changes to relocated requirements; and a characterization of the discussion of change. The NRC staff evaluation of each relocated specification and specific CTS detail presented in Table R is provided below.

CTS 3.1.1.3 - Boron Dilution Requirements for boron dilution are relocated to the TRM. Boron Dilution requires a minimum flow rate of at least 3000 gpm to ensure This ensures that reactivity changes are gradual during boron concentration reductions in the Reactor _ Coolant Sys. tem (RCS)[ffhe operator; therefore,1will be able to recognize and control the -The reactivity change rates rate associated with buron concentration reductions. wlll, therefore, be wl:hin the espebll;ty of opereter l

rece;in;;;en en-l eentre. However, this requirement is not an input assumption for any design basis accident (DBA) analysis and is not required.to dose act mitigate any accident. CTS 3.1.1.3 does not meet any of the criteria in 10 CFR 50.36. Therefore, per the_NR.C Final Policy; Statement:10 CFR 50.00 this Specification is relocated out of the ITS. Any changes to these fc'rmer requirements regarding. boron.dilutionilas relocated tythe TRM will require a 10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation ensures that any changes to these procedures;and;descriptionsjequirenients will be evaluated for safety kpact. This change is consistent with the_STS NUCCO 1402.

CTS 3.1.2.1 - Boration Flow Paths - Shutdown; CTS 3.1.~ 2:2 ^-- - :*= Maw Les Qgg=*A m CTS 3.1.2.3 Raratia4 Chwules Pt.rs.ae; Shae%.." CTS 3.1.2.5TBone Acid Pt-..ewBV" J_.T CTS 3rt2.8m-;c%3d Piiinse v O---

.. acta'312.7C r=^--t W1?1--

r masaramas She>+ dawn? CTS 3:t2.9 7 Rarated Wsts SaanrnamT One ^'.o Note to the Draft SE Reviewers: In the CCNPP Draft SE of 2/17/98, the Boration and Boric Acid Specs (listed above) were addressed separately as less restrictive changes to CCNPP l

Draft SE. The NRC staff requested that these less restrictive changes be combined and

• l addressed together because of the descriptive redundancy.

l Calvert Cliffs Unit Nos.1 and 2 t

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. DRAFT Revised: 3/9/98 The Boration System Specifications,_ CTS. 3.1_.2.1, CTS 3.1.2.2,' CTS 3.1.2.3,; CTS 3.1.2.5, CTS 3112.6,' CTS 3.1.2.7, and CTS 3.1'.2.9 are relocated to the.TRM. ;The Boration System is;a subset of the Chemical Veiume Control Systern (CVCS) and is required to contr.ol.the

{

l chemical neutron absorber (boron) concentration in, the'R.CS and to _ help maintain the _SDM.

)

To accomplish this functional requirement, a source of borated water,:oneLor,more flow' paths l

tp inject this borated water into the RCS, and appropriate cha_rging_ pumps to provide the necessary charging head are required. '.Therefore, the Boration System helps _ ensure that negative reactivity controlis available during all M_ odes of facil.ity operation. The Operability of the Boration Subsystems (except for CTS 3.1.2.4,' Boration Charging Pumps; Operating, which is discussed below) are not input assumptions for any.DBA analysis and are not required to mitigate any' accident. The Boration Subsystems do not meet any'of the criteria in 10 CFR 50.36. Therefore, per the NRC Final Policy Statement, these specifications are relocated out of the ITS. Any changes to these former requirements regarding the Boron Subsystems, as relocated to the TRM will require a 10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation ensures that any changes to these procedures and descriptions are evaluated for safety impact. These changes are consistent with the STS.

CTS 3.1.2.4 - Boration Charaina Pomos - Ooeratina Requirements for Boration Charging Pumps - Operating are relocated to the TRM. The Boration System is a subset of the CVCS and is required to control the boron concentration in the RCS. The charging pumps are used to maintain RCS volume during normal operation. The charging pumps are assumed to be Operable to mitigate a small-break loss-of-coolant-accident (LOCA) above 80% Rated Thermal Power (RTP) in Mode 1. This requirement is given in CTS 3.1.2.8 and 3.5.2. In other operational Modes, the charging pumps are not assumed to be Operable to mitigate the consequences of a DBA or transient, and are not an input assumption for any DBA analyses. The charging pumps play a significant role in the Calvert Cliffs Probabilistic Risk Assessment to support once-through-core cooling, and for Anticipated Transient Without Scram mitigation. However, the requirements for charging pumps located in ITS 3.5.2 are sufficient to ensure that the charging pumps are available to perform these functions. CTS 3.1.2.4 does not meet any of the criteria in 10 CFR 50.36. Therefore, per the NRC Final Policy Statement, this Specification is relocated out of the ITS and incorporated into the Updated Final Safety Analysis Report (UFSAR). Any changes to these former requirements regurding,the Boron Subsystems, as relocated to the TRM, will require a safety evalu_ation. pursuant to10 CFR 50.59 evektstion. The 10 CFR 50.59 evaluation ensures that any changes to these procedures and descriptions are evaluated for safety impact. This change is consistent with the STS.

CTS 3.1.3.3 - Position Indicator Channels Specifications for the Position Indicator Channels are relocated to the TRM. The Position Indicato Channels are required to determine control element assembly (CEA) positions, thereby ensuring compliance with the CEA alignment and insertion limits. The Operability Calvert Cliffs Unit Nos.1 and 2

!4 -

i DRAFT Revised: 3/9/98 OPERA 96E+TV of the Position Indicator Channels are not assumed to mitigate the.

consequences of a DBA or transient, and are not an input assumption for any DBA analysis.

CTS 3.1.3.3 does not meet any of the cliteria in 10 CFR 50.36. Therefore, per the NRC Final

@hcy Statement (10 OF'150.00, this Specification is relocated out of the ITS Any changes

. to these former requirements regarding th@osition Indicator;ChannelsRas relocatedAthe

.TRM;will, require a.10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation ensures that any changes to the former requirements are evaluated for safety impact. This change is consistent with the STS.

j ;~

- CTS 3.1.3.4 - CEA Droo Time L

Specifications'for the CEA Drop Time are relocated to the TRM. The CEA drop time is required to ensure that the CEAs insert within the time assumed in the safety analyses. The CEA Drop Time TS was approved for relocation by the NRC because CTS 3.1.3.4 it does not L

. meet any of the NCO section criteria in 10 CFR 50.36. However, the NRC requires retaining the CEA Drop Time Surveillance Requirements (SR) as requ; red to demonstrate Operability OPECA0lUTY for retained Specification ITS 3.1.5, CEA Alignment, EGO thereby verifying CEA Operability 1the SR is moved to SR 3.1.5.6). Therefore, per the NRC Final Policy. Statement 2

1C CTC 5.0.00, this Specification is relocated out of the ITS. Any changes to these former I

requirements regarding the CEA Drop Time Specification, as relocated to the TRM,1will require i

a safety, evaluation pursuant to;10 CFR 50.59 cveluet;sn. The 10 CFR 50.59 evaluation ensures that any changes to the former requirements are evaluated for safety impact. This.

j change is consistent with the STS.

CTS 3/4.3.3.1 - Radiation Monitorina Instrumentation L

CTS 3/4.3.3.1 3/4.0.0.1.1.a, containment purge and exhaust isolation area monitors, 3./4.3.3:1 G/4.0.0.1.2.b, noble gas effluent process monitors, their; associated SR,jand associated requirements in CTS Tables 3.3-6 and 4.3-3 (Functional Units 1.a and 2.b) are

- relocated to the TRM. The radiation monitoring instrumentation monitors for radiation -

i throughout the plant. Some of the radiation monitoring instrumentation specified by CTS 3/4.3.3.1 provides inputs to safety systems in order.for these systems to mitigate design basis events (DBEs)..The radiation monitors specified in the requirements being relocated, however, are not required to mitigate any DBEs, nor do they provide input into any system i

required to mitigate DBEs. These radiation monitors do not meet any criteria in -

j.

10 CFR 50.36(c)(2)(ii) for inclusion in TS. Thereforetper the NRC Final Policy Statemopt,1this l.

Specifiestion is relocated. out the ;lTS2Any changes to these former requirements ggenfing-radiation monitoring;instrumentationias relocated,to;in the TRMfWill. require requlree a safety

- evaluation pursuant to 10 CFR 50.59. Thus sufficient regulatory controls exist under 10 CFR 50.59 to ensure continued protection of the public health and safety. Therefore, the L

CTS 3(4;3.3;1 O!4.0.0.1.1.; and 3/4.3.310/4.0.0.1.2.b requirements for radiation 1

- monitoring instrumentation as specified,in CTS Tables!3.3-6 and 4.343 (Functional, Units:1;a and 2.b);;may be relocated to the TRM.

Calvert Cliffs Unit Nos.1 and 2

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! DRAFT Revised: 3/9/98 CTS 3/4.3.3.4 - Meteoroloaical instrumentation CTS 3/4.3.3.4 meteorological monitoring instrumentation and associatea requirements in Tabloc 3.3-8 and 4.3-5 are relocated to the TRM. Meteorologicalinstrumentation measures environmental parameters such as wind speed and direction and the temperature height gradient that may affect distribution of fission products and gases following a DBE, but these parameters are not an input assumption for any DBE analysis. In addition, this instrumentation does not function to mitigate any DBE. Thus this instrumentation does not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former requirements raarding meteorological instrumentation, as relocated to M the TRM will require rcqv;res a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the.NRC Final Policy Statement, this Specification is relocated out of the ITS. Thus sufficient regulatory controls exist under 10 CFR 50.59 to ensure continued protection of the public health and rafety.

Therefore, the CTS 3/4.3.3.4 requirements for meteorological monitoring instrumentation may be relocated to the TRM.

CTS 3/4.3.3.7 - Fire Detection Instrumentation CTS 3/4.3.3.7 and associated requirements in Table 3.3-11 are relocated to the TRM. The fire detection instrumentation detects fires in the plant and initiates alarms to alert plant staff to initiate fire suppression efforts. The fire detection instrumentation, however, is not used to detect a degradation of the reactor coolant pressure boundary, nor is it assumed to mitigr4a a DBE or transient. Thus the fire detection instrumentation does not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former requirements regarding b fire detection instrumentation, as relocated to the TRM, will require requ;res a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the NRC Fina! Policy Statement, I

this Specification is relocated out of the ITS. In addition, the fire de tection instrumentation is I

required by 10 CFR 50.48 and 10 CFR Part 50, Appendix R. Any changes would also have to conform to the requirements of these regulations. Thus sufficient regulatory controls over j

these requirements exist to ensure continued protection of the public health and safety.

Relocating the fire detection instrumentation requirements from the CTS is also consistent with the recommendations of Generic Letter 86-10. Therefore, the CTS 3/4.3.3.7 requirements for fire detection instrumentation may be relocated to the TRM.

CTS 3/4.4.7 - Chemistrv

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Water chemistry limits are relocated to the TRM. Reactor coolant water chemistry is monitored for a variety of reasons. One reason is to reduce,the possibility of failures;in the RCS pressure boundary; caused by corrosion. Poor coolant water chemistry contributes to the long-term degradation of system material of construction but is not of immediate importance to the plant. Therefore, this requirement has more of a long-term preventative purpose rather than mitigative purpose. This Specification does did not meet any of the four screening criteria in'10 CFR 50.36(c)(2)(ii), a,nd, therefore, does not require for retention in the Calvert l

Calvert Cliffs Unit Nos.1 and 2

. DRAFT Revised: 3/9/98 i

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Cliffs Technical Specifications. In particular, this %is specification was not found to be a significant risk contributor to the core damage frequency and offsite releases. Any changes to these former requirements regarding water chemistry, in as relocated to the TRM, will j

require a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the NRC Final Policy i

Statement, this Specification is relocated to the TRM. ;Thus, sufficient regulatory controls exist under 10 CFR 50.59 to ensure continued protection of the public health and safety.

j Therefore, the CTS 3/4.4.7 requirements for water chemistry limits may be relocated to the TRM.

CTS 3/4.4.9.2 - Pressurizer Pressure /Temoerature Limits Pressurizer pressure / temperature limits are relocated to the TRM. The Pressurizer Pressure / Temperature Limits Specification places heatup and cooldown limits on the pressurizer to prevent non-ductile failure and assure compatibility o' operation with the fatigue i

analysis performed. The limits meet the requirements given in the ASME Code. These

{

limitations are consistent with structural analysis results but they are not initial condition assumptions of a DBA or transient. These limits represent operating restrictions and Criterion 2 inciodes operating restrictions. However, it should be noted that in 10 CFR 50.36, the Criterion 2 discussion specifies only those operating restrictions required to preclude unanalyzed accidents and transients be included in the Technical Specifications. This Specification did not meet any of the four screening criteria for retention in the Calvert Cliffs Technical Specifications. This Specification was not found to be a significant risk contributor to the core damage frequency and offsite releases. Any changes to these former requirements regarding pressurizer pressure /temperaturo limits, as_ relocated to in the TRM, will require rcqwss a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the NRC Final Policy; Statement; this specification is relocated out of the ITS. Thus, sufficient regulatory controls exist under 10 CFR 50.59 to ensure continued protection of the public health and safety. Therefore, the CTS 3/4.4.7 requirements for pressure / temperature limits may be relocated to the TRM.

CTS 3.4.10.1 - ASME Code Class 1. 2. and 3 Comoonents Specifications for ASME Code Class 1,2, and 3 components are relocated to the TRM. The inspection programs for ASME Code Class 1,2, and 3 components ensure that the structural integrity of these components will be maintained throughout the components' life. ASME Code Class 1, 2, and 3 components are monitored so that the possibility of component structural failure does not degrade the safety function of the system. The monitoring activity l

is of a preventative nature rather than a mitigative action. This Technical Spe^ification is more directed toward prevention of component degradation and continued long-terrn maintenance of acceptable structural conditions. Thus it is not necessary to retain this Specification to ensure immediate operability of safety systems. This Specification does did

> t meet any of the four screening criteria for retention in the Calvert Cliffs Technical Specifications. This Specification was not found to be a significant risk contributor to the Calvert Cliffs Unit Nos.1 and 2 t

. DRAFT Revised: 3/9/98 core damage frequency and offsite releases. Therefore, this requirement is not essential for responding to a DBA or transient. Any changes to these former requirements regarding ASME l.

Code Class 172,:and 3 components, in as. relocated.to in the TRM.will_ require a safety -

s l

evaluation pursuant to 10 CFR 50.59. There. fore, per,the NF.C. Final Policy Statement;this

.Specificatiottis relocated out of the lTS.3Thus sufficient regulatory controls exist under 10 l

CFR 50.59 to ensure continued protection of the public health and safety. Thus CTS

.3.4.10.1 requirements for ASME Code Class 1,2, and 3 components may be relocated to the

'TRM.

1 i

CTS 3.4.11 - Core Barrel Movement

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l Specifications for the Core Barrel Movement are relocated to the TRM. This Specification is i

provided to ensure early detectia of excessive core movement if it sho' ld occur. The u

movement is detected by using four excore detectors to obtain axial power distribution (APD) and safety, analyses (SA). This Technical Specification is for prevention of component degradation and continued long-term maintenance of acceptable structural conditions and does not mitigate design basis accidents or transients. Hence is not necessary to retain this Specification to ensure immediate operability of safety systems. This specification does not meet any of the criteria for retention in the Calvert Cliffs Technical Specifications; nor-was it was not found to be a significant risk contributor to the core damage frequency and offsite releases. Therefore, per the NRC Final Policy Statement, this Specification is relocated out of L

the ITS. Any changes to these former requirements regarding core _ barrel movementias relocated to,the TRM, will require a safety evaluation pursuant to 10 CFR 50.59 cvaketion.

The 10 CFR 50.59 evaluation ensures that any changes to these requirements are evaluated for safety impact. This change is consistent with NUREG-1432.

l' CTS 3/4.4.12 - Letdown Line Excess Flow CTS 3/4.4.12 requirements for the letdown line excess flow are relocated to the TRM. This specification is provided to ensure the letdown excess dow check valve is closed to ensure that ruptures downstream will not exceed the accident analysis. This valve is used to relieve bleedoff pressure after a letdown excess flow check valve isolation. Since crifices were placed on the valves the bleedoff function is no longer required. The letdown excess flow check valve is a manual valve that is located inside containment and kept locked closed. This Specification does did not meet any of the screening criteria in 10 CFR 50.36 for retention in the Calvert Cliffs Technical Specifications. Any changes to thessiformer;this requirements reprement regarding letdown.line excess. flow, as relocated tojin the TRM, will require j

requires a safety evaluation pursuant to 10 CFR 50.59. Thereforefper,the;NRC;Finai Policy Statement,1this. Specification is relocated.out otthe ITS. Thus sufficient regulatory controls exist under 10 CFR 50.59 to ensure continued protection of the public health and safety.

Therefore, the CTS 3/4.4.12 requirement for let down line excess flow may be relocated to

[

the TRM.

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I Calvert Cliffs Unit Nos.1 and 2 i

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'27-DRAFT Revised: 3/9/98 i

CTS 3/4.4.13 - Reactor Coolant System Vents CTS 3/4.4.13 requirements for the reactor coolant system vents are relocated to the TRM.

This specification is provided to ensure the reactor vessel head vents are available to exhaust noncondensible gases that wheeh could inhibit natural circulation from the:RCS. Their.

function, capabilities, and testing requirementa are consistent with the requirements of NUREG-0737. The operation of the reactor vessel head vents, however, is not part of the primary success path for mitigating a DBE or transient. The operation of these vents is an operator action after the event has occurred, and is only required when there is indication that natural circulation is not occurring. Thus the reactor vessel head vents do not meet any

. criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these formar requirements regarding reactor coolant, system vents,"as relocated to in the TRMRwift require l

a safety evaluation pursuant to 10 CFR 50.69. Therefore, per the.NRC. Final. Policy Statement, this. Specification.is relocated out of the lTS.0Thus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety.

Therefore, CTS 3/4.4.13 may be relocated to'the TRM.

CTS 3/4.7.2.1 - Steam Generator Pressure Temoerature Umitation

. Requirements to ensure that either side of the SG is not pressurized above 200 psig if the temperature of either the prhery side of the SG er.d. ease:is below 80Y902F are relocated to the TRM. SG pressure and temperature (P/T) limits ensure that pressure-induced stresses on the SGs do not exceed the maximum allowable fracture toughness limits. P/T limits are based on maintaining SG RTuor sufficient to prevent brittle fracture. Technical Specifications are provided for limits on variables consistent with structural analysis results. However, P/T limits do not represent initial condition assumptions of a UFSAR acc! dent analysis. While P/T limits represent operating restrictions and Criteria 2 includes operating restrictions, Criterion 2 does not apply since it applies only to.those operating restrictions required to preclude analyzed accidents and ent transients. Any changes to these former requirements regarding steam generator pressure temperature;limitationgas. relocated to in the TRM2will require a j

safety evaluation pursuant to 10 CFR 50.59. - Therefore,-;per the NRCfinal Policy Stetement, this Specification is relocated,out.of the,lTSMThus sufficient regulatory controls exist under 10 CFR 50.59 to ensure continued protection of the public health and safety. Therefore, CTS L 3/4.7.2.1 requirements for steam generator pressure temperature limitations may be relocated to the TRM.

I CTS 3/4.7.8.1 - Snubbers Requirements for snubbers which ensure that the structural integrity of the RCS and all other safety-related systems is maintained during and following events initiating dynamic loads are relocated to the TRM. This Specification only requires snubbers to'be Operable. Snubbers do I

are not function regred to mitigate any design basis _ event (DBE):or transient, nor do they provide input to any system required to mitigate DBEs. This equipment does not meet any Calvert Cliffs Unit Nos.1 and 2

1 DRAFT Revised: 3/9/98 criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former requirements regarding snubbers, as relocated to in the TRM, will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the NRC Final Policy Statement,;this_ Specification is relocated out.of.the ITS. -Thus sufficient regulatory controls exist to ensure continued protection of the public health and safety. Therefore, CTS 3/4.7.8.1 may be relocated to the TRM.

CTS 3/4.7.9.1 - Sealed Source Contamination CTS 3/4.7.9.1 requirements for sealed source contamination ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. The limitation on removable contamination ior sources requiring leak testing, including alpha emitte's, is based on 10 CFR 70.39(a)(3) limits for plutonium. Limiting leakage of radioactive material from sealed sources is not required to mitigate a DBE or transient, and limits on such leakage are not an initial assumption in any DBE analysis. Thus limits on sealed source contamination do not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS.

Any changes to these former requirements regarding sealed source contamination, as relocated to in the TRM will require a safety evaluation pursuant to 10 CFR 50.59. Any changes must also conform to the requirements of 10 CFR 70.39(a)(3). Therefore, per the NRC Final Policy Statement, this Specification is relocated out of the ITS. Thus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.7.10 may be relocated to the TRM.

CTS 3/4.7.10 - Watertiaht Doors CTS 3/4.7.10 requirements for watertight doors are relocated to the TRM. These requirements ensure the protection of safety related equipment from the effects of water or stsam escaping from ruptured pipes or components in adjoining rooms. This specification lists several watertight doors that are required to be closed, except fgr normal. entry or exit. These watertight doors are not required to be closed to mitigato any DBE or tra_nsient, nor do they provide input to any system required to mitigate DBEs. Thus watertight doors do not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former reouirements regarding watertight doors,'as relocated to in the TRM; will. require a safety evaluation pursuant to 10 CFR 50.59. Therefore, per,the NRC Final Policy Statementithis Specification is relocated.out of ITS. Thus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.7.10 may be relocated to the TRM.

CTS 3/4.7.11.1 - Fire Suooression Water Svstem and CTS 3/4.7.12 - Penetration Fire Barriers CTS 3/4.7.11 contains Specifications for the Fire Suppression Water System (3/4.7.11.1),

Spray and/or Sprinkler System (3/4.7.11.2), Halon Systems (3/4.7.11.3), Fire Hose Stations (3/4.7.11.4), and Yard Fire Hydrants and Hydrant Hose Houses (3/4.7.11.5). These Calvert Cliffs Unit Nos.1 and 2

l DRAFT Revised: 3/9/98 Specifications along with CTS 3/4.7.12, Penetration Fire Barriers are relocated to the TRM.

The Fire Suppression Systems and the Penetration Fire Barriers are required to provide fire protection. The Fire Protection Systems and the Penetration Fire Barriers are not required to mitigate any DBA or transient., nor de they providc inpu m orm... required to in : gate

-1 BBEt While the other Fire Protection Systems do not play a significant role in the Calvert Cliffs Probabilistic Risk Assessment, the Fire Suppression Water System, and the Halon System do play a significant role in the preliminary Calvert Cliffs Fire Probabilistic Risk f

Assessment. However, these requirements are allowed to be relocated out of ITS by NRC l

Generic Letter 88-12 in.which the %e NRC determined that the Fire Protection System can be controlled outside ITS. Any changes to these former requirements regarding fire suppression water system _ and penetration fire barriers, as relocated to the TRM,will require a i

safety evaluation pursuant to 10 CFR 50.59. The 10 Crn 50.50 ensurcs that chengcs to thcas requ;rcrncnts requerc a ssfety avaiustion pursuant to 10 Crn 50.50. In addition; there are fire protection. requirements in Appendix R. Therefore, per the NRC Final Policy Statement, this Specification is relocated out of ITS. Thus sufficient regulatory controls exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.7.11.1 and 3/4.7.12 may be relocated to the TRM.

CTS 3/4.9.3 - Decav Time CTS 3/4.9.3 requirements for decay time are relocated to the TRM A decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> is currently specified oefore moving irradiated fuel in the reactor pressure vessel. It is not expected that the plant will ever be prepared to move irradiated fuel within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of subcriticality. Based on this physical feature, this specification c'oes not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former requirements regarding decay. time, as relocated to in the TRM, will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the NRC Final Policy Statement,;this Specification is relocated out.of the ITS. Thus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.9.3 may be relocated to the TRM.

CTS 3/4.9.5 - Communications CTS 3/4.9.5 requirements for communications are relocated to the TRM. Direct communications are used between the control room and refueling station personnel. This l

communication is not an initial condition. assurnptlon of any DBE or transient, nor does it provide input into any system required to mitigate DBEs. Thus this equipment does not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these formerj j

requirements regarding communication; as relocated.to in the TRMewill require a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the. NRC Final Policy Statem_enti;this Specification'is relocated out of the'lTSOThus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.9.5 may be relocated to the TRM.

Calvert Cliffs Unit Nos.1 and 2 l

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e 1*

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' DRAFT Revised: 3/9/98 CTS 3/4.9.6 - Refuelina Machine Ooerability Requirements for Refueling Machine Operability are relocated to the TRM. The Refueling Machine is used to transfer CEAs and fusi assemblies. The Refueling Machine doesinot functionja riot required to mitigate any DBE or transient, nor does it provide input into any system required to mitigate DBEs. This equipment does do not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former. requirements regarding refueling machine operability, as relocated to in the TRM, will require a safety evaluation pursuant to 10 CFR 50.59. Therefore,'per the NRC Final Policy Statement;this specification is relocated out of the ITS. Thus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.9.6 may be relocated to the TRM.

CTS 3/4.9.7 - Crane Travel - Soent Fuel Storaae Pool Buildino Requirements for the Crane Travel - Spent Fuel Storage Pool Building are relocated to the TRM. The Spent Fuel Pool crane travel deals with preventing loads in excess of 1600 pounds from traveling over fuel assemblies, unless the loads are handled by the single failure proof Spent Fuel Pool Cask Handling Crane. Control of the crane travel does not function irMtot reqtme to mitigate any DBE or transient, nor does it provide input to any system required to mitigate DBEs. This equipment does not meet any criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in TS. Any changes to these former requirements regarding crane travel-spent fuel storage pool building, as relocated to in the TRM, will require a safety evaluation pursuant to 10 CFR 50.59. Therefore, per the NRC Final Policy Statement, this _ specification is relocated out of the ITS. ;Thus sufficient regulatory controls over these requirements exist to ensure continued protection of the public health and safety. Therefore CTS 3/4.9.7 may be relocated to the TRM.

CONCLUSION The relocated CTS discussed above are not required to be in the TS under 10 CFR 50.36 and 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 curren.t specifications, information, and requirements that are being moved to licensee-controlled documents. This is the subject of a license condition established herewith.

Until incorporated in the UFSAR and procedures, changes to these specifications, information, and requirements will be controlled in accordance with the current applicable curicnt procedures that control these documents. Following implementation, the NRC will audit the i

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removed provisions to ensure that an appropriate level of control has been achieved. The l

NRC staff has concluded that, in accordance with the Final Policy Statement, sufficient i

Calvert Cliffs Unit Nos.1 and 2 l

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e i!' DRAFT Revised: 3/9/98 regulatory controls exist under the regulations, particularly 10 CFR 50.59. Accordingly, these specifications, information, and requirements, as described in detail in this Safety Evaluation, may be relocated from CTS and placed in the UFSAR or other licensee-controlled documents as specified in the licensee's letter dated December 4,1996.

F. Control of Specifications, Requirements, and Information Removed from 6he CTS The facility and procedures described in the UFSAR and TRM, incorporated into the UFSAR by reference, can only be revised in accordance with the provisions of 10 CFR 50.59, which ensures records are maintained and establishes appropriate control over requirements removed from CTS and over future changes to the requirements. Other licensee-controlled documents contain provisions for making changes consistent with other applicable regulatory requirements; for example, the Offsite Dose Calculation Manual (ODCM) can be changed in accordance with ITS 5.5.1, the emergency plan implementing procedures (EPIPs) fEPlGs) can be changed in accordance with 10 CFR 50.54(q); and the administrative instructions that implement the Quality Assurance Manual (QAM) can be changed in accordance with 10

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CFR 50.54(a) and 10 CFR Part 50, Appendix B. Temporary procedure changes are also l

controlled by 10 CFR 50.54(a). The documentation of these changes will be maintained by i

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

The licensee committed in a letter dated December 4,1996, to confirm that CTS requirements designated for placement in the UFSAR or the TRM are appropriately reflected in these

. documents, or that they will be included in the next required update of these documents.

This is the subject of a license condition established herewith. The licensee has also committed to maintain an auditable record of, and t implementation schedule for, the procedure changes assnciated with the development of ITS. The licensee will maintain the documentation of these changes in accordance with the record retention requirements in the QA plan and the TRM. The December 4,1996 letter, Attachment IV, " Application of the Technical Specification Selection Criteria (Split Report)," includes a list of the changes involving specific requirements that have been removed from the CTS. For each of these changes, Attachment IV also includes the licensee-controlled documents and the TS or regulatory requirements governing changes to these documents.

1 G. EVALUATION OF OTHER TS CHANGES INCLUDED IN THE APPLICATION FOR CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS i

ITS SR 3.4.9 - Pressurtzer The licensee is proposing to add a new surveillance requirement, SR 3.4.9.2, to the ITS which will require verification every 24 months that the capacity of each required bank of pressurizer heaters is equal to or greater than 150 kW m.,

The CTS, as well as LCO 3.4.9 m,,~,.....

&4&b in the ITS, require "two banks of pressurizer heaters operable with the capacity of I

Calvert Cliffs Unit Nos.1 and 2

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. DRAFT Revised: 3/9/98 each bank 1150 kW.", ", The CTS do not have a SR that requires verification of the heater j

capacity, but Calvert Cliffs has been performing this verification each refueling cycle.

Therefore, adding this SR constitutes a more restrictive change. Since Calvert Cliffs is going to a 24 month fuel cycle, the proposed frequency of 24 months will ensure insttre that this verification is performed each refueling cycle.

The requirement to have two banks of pressurizer heaters ensures that RCS pressure can be maintained. The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled. Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat. Safety analyses presented in the Updated Final Safety Analysis Report do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure. The genede value of 150 kW capacity per bank is derived from the use of 12 heaters rated at 12.5 kW each. The heater capacity emetet needed to maintain pressure is dependant on the ambient heat losses. The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure and provides a wide subcooling margin to saturation in the loops. Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power is the reason for their inclusion. We find the proposed addition of the new SR 3.4.9.2 acceptable, since this is a desirable improvement from the current TS, which represents the design basis at Calvert Cliffs.

ITS 3.4.10 - Pressurizer Safetv Valves Each Calvert Cliffs unit has two spring loaded pressurizer code safety valves intended whose pwpcac is to provide RCS overpressure protection. Operating in conjunction with the Reactor Protective System, two valves are used to ensure that the S3fety Limit (SL) of 2750 psia is not exceeded for analyzed transients during operation in Modes 1 and 2. Two safety valves are also used for portions of Mode 3 (Hot Standby). For the remainder of Mode 3, Mode 4, Mode 5, and Mode 6 with the head on, overpressure protection is provided by operating procedures and ITS the LCO 3.4.12, " Low Temperature Overpressure Protection (LTOP)

System." The current TS applicability requires that both safety valves be operable in Modes 1,2 and 3 and that one safety valve be operable in Modes 4 and 5. The ITS will modify these applicability requirements for Mode 3 to specify that two safety valves shall be OPERABLE with all RCS cold leg temperatures > 365* F for Unit 1 and > 301" F for Unit 2fh The LCO is not applicable in Mode 3 when all RCS cold leg temperatures are 1365 F (Unit 1) and 1 301* F (Unit 2) and Modes 4 and 5, and Mode 6 with the reactor vessel head on, because LTOP protection i:: provided. Therefore, the proposed modification in " Applicability" in the ITS for OPERAElLITY of the safety valves is acceptable.

ITS SR 3.4.11 - Pressurizer Power Ooerated Relief Valves (PORVs) l Calvert Cliffs Unit Nos.1 and 2

. DRAFT Revised: 3/9/98 The CTS require that each PORV shall be demonstrated OPERABLE by performance of a Channel Function Test once per 31 days. The licensee proposes to change the test interval from once per 31 days to once per 92 days as part of the conversion to the ITS. The PORV actuation instrumentation is the same as that used for the RPS High Pressurizer Pressure Function. The RPS High Pressurizer Pressure Function STE Surveillance Frequency was decreased from 31 days to 92 days in the RPS and ESFAS " monthly to quarterly" Technical Specification change tapproved in an NRC Safety Evaluation Report for Amendments 193 and 170 for Units 1 and 2, respectively, dated August 24, 1994). The PORV actuation and the High Pressurizer Pressure High Trip Setpoint share the same instrumentation even down to the bistables b;staplcs. The staff determined that since it was acceptable to change testing of the RPS instrumentation from monthly to quarterly, that it was also acceptable to test the PORV actuation instrumentation from monthly to quarterly because of similar. significance of.the two systems.

ITS SR 3.4.13 - Reactor C_oolant System Ooerational Leakaae Current TS 3.4.6.2.C specifies that reactor coolant system leakage shall be limited to "1 gpm total primary-to-secondary leakage through all steam generators and 100 gallon-per-day through any one steam generator." The proposed ITS LCO SR 3.4.13 eliminates the limit of 1 gpm total primary-to-secondary leakage through all steam generators and thus will only require a limit of 100 gpm through any one steam generator. The 1 gpm limit was deleted because the combined leakage from both steam generators of 200 gpm is only a fraction of the gpm limit; that is, the 1 gpm leak rate limit can never be reached. This change is acceptable because the proposed limit of 100 gallons per day from any one steam generator provides a bounding limitation on primary-to-secondary leakage.

ITS SR 3.5.2 - ECCS - Ooeratina Current TS SR 4.5.2.f.2 requires verifying at least once per REFUELING INTERVAL, during shutdown, that the High Pressure Safety injection Pump and the Low Pressure Safety injection (LPSI) Pump start automatically upon receipt of a Safety injection Actuation Test Signal. Proposed ITS SR 3.5.2.6 retains this same requirement with a specified frequency of 24 months, whi::h is equivalent to the CCNPP refueling interval. The proposed ITS will add a new SR 3.5.2.7 which requires verification that each LPSI pump stops on an actual or simulated actuation signal. Currently, upon receipt of a recirculation actuation signal, the LPSI pumps are tripped; however, there is no CTS Surveillance to verify this. This change is more restrictive since it adds a requirement to the TS. The addition of this new SR will enhance safety, because it requires verification of the effect on LPSI at.thefactuation signal,;e-SH the failure of which could adversely affect a safety function. We find the proposed SR 3.5.2.7 acceptable because this is an improvement from the current TS; which represents the design basis for Calvert Cliffs, l

ITS SR 3.7.8.3 - Control Room Emeroenev Ventilation Svstem Calvert Cliffs Unit Nos.1 and 2

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. DRAFT Revised: 3/9/98 The proposed amendment changes the surveillance interval from 18 months to 24 months (each refueling interval) for the following SR (surveillance requirement). Current TS SR 4.7.6.1.e.2 requires demonstrating that each train of the Control Room EVS (CREVS) is operable at least once every 18 months by verifying that on a control room high radiation test signal, the system automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber absorber banks and that both of the isolation valves in each inlet duct and common exhaust duct, and the isolation valve in the toilet area exhaust duct, close.

ITS SR 3J.8_.3yn 4.7.0.1.c.2 changes SR.4.7.6.1.e.2 to require demonstrating CREVS operability at least once every 24 months by verifying that each CREVS train actuates on an actual or simulated actuation signal.

The CREVS is an emergency system which may also be operated during normal operati)ns. It consists of two independent and redundant safety-related safsty iciated trains each v' 1 a-prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber abscrber section for removal of gaseous activity (principally iodine), fresh air intake dampers, a common exhaust damper, and common supply and return ducts that recirculate and filter the control room air.

The CREVS is designed to maintain a suitable environment in the control room for operating personnel and safety-related control equipment following an accident involving the release of radioactivity, chemical or toxic gases. Detection of high radiation above a the setpoint will actuate the CREVS, closing the unfiltered outside air intake and unfiltered exhaust dampers and aligning the system for recirculation of control room air through the HEPA and charcoal filters. The CREVS functions to mitigate the consequences of certain design basis accidents by pressurizing the control room and providing filtered recirculated air to control room personnel. Total system failure could result in a control room operator receiving a dose in excess of 5 rem due to radiation release.

Currently; the SR calls for verifying that a high radiation test signal will result.in_to casure that the CREVs actuating actuates as required at _least once pe_r 18, months. The proposed change calls for verifying, at least every 24 months, that CREVS starts and operates as required on either both an actual or and a simulated actuation signal.

The staff reviewed the proposed changes and the licensing basis. In addition to the license and the CTS 76, the licensing basis includes Regulatory Guide (RG) 1.52 (Rev. 2), UFSAR Section 9.8.2.3, " Auxiliary Building Ventilation Systems," and UFSAR Chapter 14, " Fuel i

Handling." Extending the surveillance intervals will not affect any accident initiators, or affect the consequences of an accident.

l lTS.SR 3.7.8.1 requires that the CREVS be demonstrated operable at least once every 31 l

days by initiation of flow through each HEPA filter and charcoal adsorber absorber train and Calvert Cliffs Unit Nos.1 and 2

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DRAFT Revised: 3/9/98 verifying that each system train operates for a minimum of 15 minutes. Also, the CREVS consists of redundant electrical and mechanical components which are operated once every 31 days to verify operability. Therefore, components of CREVS are exercised by other more frequent surveillance such that the components m...r..mm will not be experiencing a longer period of disuse or a longer period between operation. Based on the cumulative effect of the required surveillances, the staff concludes that sufficient testing is required to enable the licensee to make an adequate determination of CREVS operability on ; ccatlnue; bei.

A'so the surveillance records over the past ten. years were reviewed by the licensee. Ns failures have occurred during his time period. Therefore, the licensee concluded that there is no potential for any increases in failure rates of the components under a nominal 24-month surveillance interval.

In accordance with the GL 91-04, " Changes in. Technical Specifications Surveillance Interva!s to Accommodate a 2' Month Fuel Cycle," the affect of this change on plant safety would be very small, historical data supports this and ccnclusion, and cxccpt fcr cac devlation wah l1C 4-Mr no assumptions in the plant licensing basis would be invalidated. Therefore, this change is acceptable to the staff.

ITS SR 3.7.9.1 - Control Room Emeraency Temoerature Svstem (CRETS)

The proposed amendment will change the surveillance interval from 62 days on a staggered test bases (one train every 31 dayst to 24 months (each refueling interval) for the following SR:

Current TS SR 4.7.6.1.a requires demonstrating that each CRETS train is operable at least once every 62 days, on a staggered test basis (one train every 31 days), by: (1) deenergizing the backup Control Room air conditioner; and (2) verifying that the emergency Control Room air conditioners maintain the air temperature s 104*F for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when in the recirculation mode.

SR 4.7.6.1.a changes to ITS SR 3.7.9.1 to require demonstrating operability of CRETS at least once every 24 months by verifying each CRETS train has the capability to maintain control room temperature within limits.

The CRETS is a system that maintains the control room temperature below 104 *F during an

. emergency to maintain personnel occupancy and equipment operability requirements. It may also be operated during normal operations. It consists of two independent and redundant safety-relatedleefe;y re e:ed trains each with cooling coils, instrumentation, and controls to maintain 3rcvlde a required temperature in of the control room to ensure that this the con:rc room temperature will not exceed equipment operability requirements following isolation of the con' rol room. Demonstrating the operability of the CRETS involves verifying that the t

emergency Control Room air conditioners maintain the air temperature s104*F for at least 12 Calvert Cliffs Unit Nos.1 and 2

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. DRAFT Revised: 3/9/98 hours. CRETS operability is also demonstrated at least once per 31 days by initiation offlow through each HEPA filter and charcoal absorber train and verifying that each of the two trains operates for at least 15 minutes.

i The staff has reviewed the proposed changes with respect to GL 9'94. All actions specified in the GL were completed. Thc licensee concluded that extending the surveillance interval from 62 days to 24 months would have a small effect on safety. This is based on its review of historical data over past ten M years which indicated that the CRETS test has not had any failures and, therefore, significant degradation of the system is slow and is not expected to change during the surveillance period. This is an acceptable conclusion to the staff. The staff be.ieves that the previously established 62 day surveillance interval for the CRETS is very conservative, end notes that the operability of the system is verified every 31 days, and notes that sufficient system testing is involved to preclude undiscov > red significant oegradation of i

the system components. Also, instrument drif+ will have no r.fect on this test, in so far as there There is no instrumentation tested by th 'TS Tcchnic 1 Spccificction requirement.

Therefore, the staff finds the p*oposed change.,ceptable.

ITS SR 3.7.11.3 - Soent Fuel Pool F@,aust Ventilation Svstem (SFF6VS)

The proposed amendment will change the surveillance interval from 18 months to 24 months (each refueling interval) for the following SR:

Current TS SR 4.9.12.d requires demonstrating that the SFPEVS is operable at least once per 18 months by: (1) verifying that the pressure drop across the combined HEPA filters and charcoal adsorber escrber banks are <4 inches Water Gauge while operating the ventilation system at a flow rate of 32,000 cfm i 10%; and (2) verifying that each exhaust fan maintains the spent fuel storage pool at a measurable negative pressure relative to the outside atmosphere during syster.t operation.

SR 4.9.12.d changes to iTS SR 3.7.11.3 to require demonstrating that the SFPEVS is operable at least once per 24 months by verifying that each exhaust fan maintains the spent fuel storage pool at a measurable negative pressure relative to the outside atmosphere during system operation.

The proposed change increases the surveillance interval from 18 months to 24 months for verifying that the SFPEVS can maintain a measurable negative pressure in the spent fuel pool l

area of the Auxiliary Building. The staff reviewed the proposed changes and the licensing basis, in addition to the ITS % and the license, the licensing basis includes UFSAR Section 9.8.2.3, " Auxiliary Building Ventilation System," UFSAR Section 1418, " Fuel Handling f

incident," and RG 1.25, " Assumptions Used for Evaluating The Potential Radiological Consequencer of a Fuel Handling Accident in The Fuel Handling and Storage Facility for Boiling and F.ersurized Water-Reactors."

i Calvert Cliffs Unit Nos.1 and 2

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> DRAFT Revised: 3/9/98 i

l The SFPEVS consists of two independent, redundant exhaust fans, a high efficiency particulate air (HEPA) filter, ductwork, valves, instrumentation, and an activated charcoal adsorber abscibsr section for removal of gaseous activity (principally iodines) from the atmosphere of the fuel pool area following a fuel handling accident. The SFPEVS is capable of maintaining a negative pressure in the spent fuel pool area relative to the outside atmosphere of the Auxiliary Building during normal operations and following accident conditions. Also, instrument drift will have no affect on the test--in that there There is no instrumentation associated w35 this Technical Specification requirement. Decreasing Surveillance Frequencies constitutes a less restrictive change. The staff reviewed the proposed changes with respect to GL 91-04. All actions specified in the GL were completed. The licensee concluded that extending the surveillance interval from 18 to 24 months would have a small effect on safety.

This is based on the licensee's review of hinorical cata over the past ten 40 years which indicated that the SFPEVS test has never failed to maintain a measurable negative pressure in the spent fuel pool area (J the Auxiliary Building. Therefore, increasing the surveillance interval by six (6) months is not expected to result in any change in the licensee's ability to

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determine changes to the integrity of the SFPEVS. The staff finds the proposed change acceptable.

ITS SR 3.7.12.3 - Penetration Room Exhaust Ventilation Svstem (PREVS)

The proposed amendment will change tne surveillance interval from 18 months to 24 months (each refueling interval) for the following SR:

Current TS SR 4.6.6.1.d.2 requires demonstrating that each PREVS train is operable at least once per 18 months by verifying that the filter train starts on r Containment isolation Test Signal.

SR 4.6.6.1.d.2 changes to ITS SR 3.7.12.3 to require demonstrating operability of the PREVS at least once every 24 months by verifying each PREVS train starts on an actual or simulated actuation signal.

The proposed Enge increases the surveillance interval from 18 months to 24 months for verifying that ;he PREVS will actuate on an actual or simulated actuatien signal. The staff reviewed the proposed changes and the licensing basis. In addition to the TS and the license, the licensing basis includes UFSAR Section 6.6.2, " Containment Penetration Room Ventilation System," and S3ction 14.24, " Maximum Hypothetical Accident," and RG 1.52. Extending the surveillance interval will not affect any accident initiators, or affect the consequences of an accident.

The PREVS is an emergerecy system which consists of two independent reduMant trains composed of a prefilter, an activated charcc,al adsorber sbscrbcr section for removal of gaseous activity (principally iodine), a fan, ductwork, valves, and dampers. The system filters air from the penetration room. The PREVS is designed to actuate on a signal from Calvert Cliffs Unit Nos.1 and 2 l

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Revised: 3/9/98 containment isolation. It initiates filtered ventilation upon receipt of '.nis signal. The staff believes that an operability determination of the PREVS each 24 monfns instead of each 18 months does not provide a significant opportunity for increased undetectable system degradation.

The staff reviewed the proposed changes with respect to GL 91-04. All actions specified in the GL were completed. The licensee concludsd that extending the surveillance interval from 18 to 24 months would have a small effect on safety. This is based on de licensee's review of previously performed surveillances which indicated that no failures have tael found.

Therefore, increasing the surveDance interval by six (6) months is not expected to result in any change in the licensee's. ncer,ms ability to determine whether the equipment will actuate and operata in accordance with its design. Therefore, this change is acceptable to the staff.

IV.

STATE CONSULTATION in accordance with the Commission's regulations, the Maryland State official was notified of the proposed issuar.ce of the amendment. The State official 4r the State of Maryland had no comments.

V.

ENVIRONMENTAL CONSIDERATION Pursunt to 10 CFR 51.21,51.32, and 51.35, an environmental assessment and finding of no significent impact was published in the FederalRegister on 1998 (_ FR

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for the ITS conversion.

Accordingly, based upon the en.viron, mental assessment, the Commission has determined that issuance 'of this ITS conversion amendment will not have a significant effect.on the; quality of t.he. human environment.'

With respect;to.other3TS changes included in the application for conversion to improved lechnical. Specifications,7the, items change requirements; with. respect to: installation or use,of afacilityfoomponent located within,the restrictedLarea'as defined in30 CFILPartl20!Jhe NRC staff;has;. determined ths,t the amendment involves no_ signific. ant; increase in;the amountsiend no;significant change in the types / of any3 effluents.thatlmay be, released;offsite,=

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and:that there is no:significant increase in individual.or cumulative, occupational. radiation exposureOTheLCsr,s':[: i haslpr.eviously_ issued _a; proposed finding.that,the; amendment, involves;no;significant hazards. cor'siderationi and there-has;been po;public_ comment.on such finding (2FR A - )E According!y,.the amendment meets the eligibility criteria for. categorical l

exclusion set forth in,10 CFR 51.22(c)(9)UPursuant to 10 CFR 51.22fb) no environmental 1

l imosct.st@ ment or. environmental assessment need be prep:;3d in connection with.the issuance usthese other TSl changes included.in the amendmentt Calvert Cliffs Unit Nos.1 and 2 l-t

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Revised: 3/9/98 VI. CONCLUSION The improved CCNPP TS provide c!earer, more readily understandable requirements to ensure i

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' safe operation of the plant.' The NRC staff concludes that they satisfy the guidance in the Commission's policy statement with regard to the content of technical specifications, and conform to the model provided in NUREG-1432 with appropriate modifications for plant-specific considerations. The NRC staff further concludes that the improved CCNPP TS satisfy Section 182a of the Atomic Energy Act,10 CFR 50.36 and other applicable standardsc On

• his basis, the NRC staff concludes that the proposed improved CCNPP TS are acceptable.

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

l The Commission has concluded, based on the considerations discussed above, that: (1) there ls 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 l

Commission's regulations; and, (3) the issuance of the amendments will not be inimical to the common defense and ' security or to the health a' red safety of the public.

l Principal Contributors:

M. Reardon A. Dromerick C. Harbuck C. Schulten R. Tjader i :-

M. Weston E. Tomlinson R. Giardina R. Clark A. Chu Date:

_ 1998 l'

Calvert Cliffs Unit Nos.1 and 2'