Text

Forwards Addl Info,Requested in NRC Ltr from CE Rossi to RA Newton,To Support NRC Review of Merits Tech Specs & Bases (WCAP-12159)
	ML20244B712
Person / Time
Site:	Comanche Peak
Issue date:	05/22/1989
From:	Hinds J; WESTINGHOUSE OPERATING PLANTS OWNERS GROUP
To:	Reinhart M; Office of Nuclear Reactor Regulation
Shared Package
ML20244B548	List: ML20244B543; ML20244B712;
References
	OG-89-36, NUDOCS 8906130234
	Download: ML20244B712 (160)
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$wed sh Fower boord IG' won PDeer 0G-89-36 May 22, 1989 Mr. Mark Reinhart Senior Operations Engineer Technical Specifications Branch Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Mail Station 11 F23 Washington, D.C.

20555

Subject:

Westinghouse Owners Group MERITS Technical Specifications and Bases (WCAP-12159) Revi.ey Transmittal 2.

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Reference:

1.

NRC letter from C..E. Rossi to R. A. Newton, NRC Review of the Westinghouse Owners Group MERITS Technical Specifications, April 28, 1989, 2.

WOG letter from R. A. Newton to C. E. Rossi, MERITS Technical Specifications and Bases (WCAP-12159) Review, Transmittal 1," May 19,1909.

Dear Mr. Reinhart:

The purpose of this letter is to forward Transmittal 2, which cor.tains additional information requested in Reference 1.

Reference 1 requested this information to support the NRC's review of the MERITS Technical Specifications and Bases.

Attachment I contains additional information in response to Item 1 of Reference 1.

During the development of the roadmap, one MERITS specification was identified as containing a requirement which was approved for relocation by the NRC. This specification will be revised and submitted to the NRC at a later date.

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-Q contains the response to Item 2 of Reference 1. contains additional information in response to Item 4 of Based on the meeting held with the NRC on May 17,1989, the Reference 1.

remaining information required to respond to Item 4, including a draft Core Operating Limits Report, will be submitted to the staff for review by The MERITS Technical Specifications (TS) that will be submitted May 30, 1989.

to complete this item will result in revisions to some of the documents provided in Attachment 5 of this transmittal. contains the response to Item 5 of Reference 1.

. contains the revised versions of the specifications, bases, and The attachment also justifications changed in response to the NRC letter.

provides a table to indicate which NRC items resulted in changes to the MERITS The TS submitted, since these changes required coordination with one another.

attachment includes a List of Effective Revisions (LOER) for each specification, bases, justification, and marked up Standard Technical Specification. A revised LOER will be provided to the NRC whenever revised versions of any document are submitted.

l The software conversion requested by Item 6 will be provided by June 1,1989, as requested by the NRC.

If you have any questions on the information provided, please call me at g

(805) 595-4420.

s Very truly yours, Mb%' b h7/9 Jacqueline R. Hinds, Chairman Technical Specification Subcommittee Westinghouse Owners Group JRH/csb Attachments

cc: Mr. C. E. Rossi - NRC-Mr. J. A. Calvo - NRC Mr. J. B. George

._TU Electric Mr. M. R. Edelman - Centerior Energy Mr. T. E.-Tipton - NUMARC Mr. W. J. Mall ' - NUMARC Mr. R. A. Bernier - Arizona Public' Service Company Mr. R. R. Sgarro - Pennsylvania Power and Light-Mr. C. W. Smythe - GPU Nuclear Corporation:

Mr. S. Webster - Combustion Engineering Mr. J. P. Klaproth - General Electric Company Mr. E. J. Lozito Virginia Power WOG Primary Representatives-WOG' Technical Specification Subcommittee Mr. J. A. Triggiani Mr. W. J. Johnson Mr. R. P. DiPiazza Mr. R. L. Bencini I

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OG-89-36 May 22, 1989 ATTACHMENT 1 OVESTION SOURCE NRC letter from C. E. Rossi to R. A. Newton, NRC Review of the Westinghouse Owners Group ME"ITS Technical Specifications, April 28, 1989.

OVESTION RE0VIRED INFORMATION TO FULLY PROGRESS WITH THE REVIEW OF THE WOG'S PROPOSED NEW STS l

1.

Please provide the staff with a comprehensive, marked up version of the current Westinghouse STS which identifies the following:

a.

The specific location in the new STS of requirements to be retained from the current STS but which now reside with a different LCO. This information should be detailed; for example, show which LCO in the new STS contains each instrument in Table 3.3-6, RADIATION HONITORING INSTRUMENTATION FOR PLANT OPERATIONS from the current STS. This is an example; the staff needs this type of information for each similar table, listing, chart, etc., found in the current STS.

b.

The complete justification for not including in the new STS, requirements that were to be retained. As an example, Table 3.3-6 in the current STS contains'16 instruments; the new STS addresses 5 of O

them. The staff needs to know to where the requirements for the other 11 instruments are to be relocated and the basis for such relocation.

c.

The preferred location for requirements which have been authorized for relocation outside the new STS and which are not included in the proposed new STS. This information should be detailed.

See the example in 1.a. above.

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OG-89-36 May 22, 1989 ATTACHMENT-1

RESPONSE

Transmittal 1 provided a' comprehensive, marked up version of the current Westinghouse STS (Rev. 4a) in response to the information requested in Item 1.

This mark-up provided a detailed roadmap identifying the relationship between Sections 1, 2, 3, and 4 of the current STS and the corresponding Divisions.1, 2, and 3 of the MERITS Technical Specifications (TS). The roadmap, together with its amplifying notes, addressed the. information requested in the three sections of Item 1.

A discussion was also provided on the_ location of relocated requirements.

The roadmap identified several requirements that were inadequately addressed in-the development of the MERITS TS. The specifications associated with these requirements were revised and are listed below. The revised specifications are.

provided in Attachment 5 of this transmittal. These revisions also include changes required in response to-Item 5 in the NRC letter with regard to the "NRC Staff Review of NSSS Vendor Owners Groups' Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications."

MERITS TS Revised Due to Inadequately Addressed Requirements 3.5.1 Containment Integrity 3.5.2 Purge and Exhaust Isolation Radiation Instrumentation 3.8.4 Containment Building Penetrations This completes the information provided in response to Item 1 to allow the NRC to progress with the review of the MERITS TS.

The roadmap also identified one MERITS TS (3.6.11, Control Room Emergency Air Cleanup System) that inadvertently retained a requirement in the LC0 that was approved for relocation by the NRC. This requirement in the specification will not impede NRC progress in the review of the MERITS TS. This specification will be revised to be consistent with the NRC Split Report and submitted at a later date.

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l OG-89-36 May 22, 1989 ON:

ATTACHMENT 2 OVESTION SOURCE NRC letter from C. E. Rossi to R. A. Newton, NRC Review of-the Westinghouse Owners Group MERITS Technical Specifications, April 28, 1989.

I OVESTION RE0VIRED. INFORMATION TO' FULLY PROGRESS WITH THE REVIEW OF THE WOG'S PROPOSED NEW STS 2.

Justifications for changes to requirements which are to be retained in technical specifications must go beyond just referencing previous, plant-specific licensing actions'.

For example,-new STS LCO 3.7.1, fLECTRICAL POWER SYSTEMS, Justification for Changes, item number 7 cites only that the proposed change is "... consistent with the requirements licensed on Callaway, Diablo Canyon, and Braidwood." The staff needs to know the specific technical bases for the changes, plant specific ~

characteristics considered, and.why the WOG considers these to be generic.

RESPONSE

Revised justifications have been prepared for the MERITS TS with justifications referencing plant-specific licensing actions. The revised ~ justifications discuss the specific technical basis for the changes, identify the plant-specific characteristics considered, and why these plant-specific characteristics are considered generic, or alternatively, the MERITS specifications were justified using the current Rev. 4a STS.

The MERITS TS in Table 1 include justifications that referenced plant-specific licensing actions. Revised justifications are provided in Attachment 5.

The revised justifications include changes required in response to Item 4 in the NRC letter with regard to the Core Operating Limits Report.

Listed below are four new specifications that were added in the development of the MERITS TS. They do not have corresponding specifications in the current Rev. 4a STS. Revised justifications are included for the two Reactor Chapter specifications. Although they are different from the existing F0 and Axial Flux Difference specifications in the STS, they were justified as generic and therefore acceptable as WOG MERITS specifications.

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OG-89-36 May 22, 1989 RESPONSE (continued)

Houever, for specifications 3.6.4A and 3.6.48, there are no feedwater isolation i

specifications in the STS Rev. 4a that can be used as a reference. The Main Feedwater Isolation and Regulation Valve specifications can be considered -

generic since they can be applied to plants that must provide feedwater line isolation protection following a high energy line break. Therefore, the j

existing or subsequently revised justifications for these specifications will 1

retain their reference to plant-specific licensing actions.

MERITS TS Without Corresponding STS Rev. 4a Specifications 3.1.6B Heat Flux Hot Channel Factor - Fg(Z) (Fg Methodology) 3.1.8B Axial Flux Difference (RAOC Methodology) 3.6.4A Main Feedwater Isolation Valves 3.6.4B Main Feedwater Regulation Valves This completes the response to Item 2.

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OG-89-36 May.22, 1989 ATTACHMENT 2-Table 1

. MERITS TS ' Justifications-Revised Due To' Plant-Soecific References 3.1.6B Heat Flux Hot Channel Factor - Fg(Z) (Fg Methodology) 3.1.7 Nuclear Enthalpy Rise Hot Channel Factor - FAH 3.1.8B Axial Flux Difference (RAOC Methodology) 3.1.9 Quadrant Power Tilt Ratio 3.2.1 Reactor Trip System Instrumentation 3.2.2 Engineered Safety Features Actuation System-Instrumentation 3.2.3 Accident Monitoring Instrumentation 3.2.4 Remote Shutdown System 3.3.1 RCS Pressure Temperature,. and Flow DNB Limits 3.3.9 RCS Loop Isolation Valves 1

3.3.10 RCS Isolated Loop Startup 3.3.16 RCS Leakage Detection Instrumentation 3.4.1 Accumulators 3.4.2 ECCS Trains - T 2 350*F avg 3.7.1 AC Sources - Operating 3.7.2 AC Sources - Shutdown 3.8.5 Residual Heat Removal and Coolant Circulation - High Water Level i

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OG-89-36 May.22, 1989 ATTACHMENT 3 OVESTION SOURCE l

I NRC letter from C. E. Rossi to'R. A. Newton,'.NRC Review of the Westinghouse Owners Group MERITS Technical Specifications, April 28, 1989.

The WOG meeting with the NRC Staff on May 17, 1989 to discuss the Reactor Chapter and ' parameter. limits to be. contained in the_ Core Operating Limits Report (COLR).

.j OUESTION RE0VIRED INFORMATION TO FULLY PROGRESS WITH THE REVIEW OF THE WOG'S-PROPOSED NEW STS 4.

Associated with the example used in item 3 above, Generic Letter 88-16 is.

not intended to place Safety Limits-in a COLR, and it is not intended to place parameters which are not currently cycle specific in a COLR.

With respect to the WOG's proposed new STS,-the staff position is that the following parameters should not be relocated to a COLR since they are not cycle specific. Therefore, the WOG should provide complete technical specifications for these parameters.

a.

Safety Limit 2.1.1, Reactor Core Safety limit, b.

LC0 3.1.2, Moderator Temperature Coefficient (MTC).

c.

LC0 3.3.3, RCS Pressure / Temperature Limits.

d.

LC0 3.4.1 c, ECCS Accumulators boron. concentration, e.

LCO 3.4.5 b, ECCS RWST boron concentration, j

f.

LCO 3.8.1, REFUELING-OPERATIONS Boron Concentration.

For other parameters proposed to be included. in a COLR, the new STS must provide the specific references to NRC-approved methodologies used to develop the respective parameters.. Generic Letter 88-16 requires that the specific references " identify the Topical Report (s) by number, title, and date, or identify the staff's safety evaluation report for a plant-specific methodology by NRC letter and date " As mentioned in item 2 above, an NRC-approved, plant-specific methodology, by itself, does not constitute an NRC-approved, generic methodology. The new STS should also make it clear that individual plants have the option of retaining cycle specific parameters in their TS even if approved methodologies for their plant

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exists.

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OG-89-36 May 22, 1989 ATTACHMENT 3 00ESTION (continued)

As an alternative, it may be more practical to retain all of the parameters in the body of the new STS and to identify the COLR option for plant specific submittals. With this alternative, submittal of the COLR and referenced NRC-approved methodologies could be deferred until-the lead plant submittals.

MEETING CLARIFICATION COMMENTS A number of MERITS TS will be revised as a result of comments from the meeting with the NRC staff on May 17, 1989. The specifications which include references to the COLR will be revised to incorporate the comments from this meeting and will be submitted by the May 30, 1989 commitment date.

RESPONSE

Reference 2 provided a list of the parameters that will be provided in the COLR, along with the reference containing the associated methodology for each parameter.

Revision A of specifications 3.1.4, 3.1.5, 3.1.8A, and 3.1.8B submitted in WCAP-12159 are acceptable and based upon the discussion at the NRC meeting, no further actions are necessary at this time.

1 For specifications 3.1.2, 3.1.6A,'3.1.6B, 3.1.7, and 3.8.1, changes must be made to incorporate the comments from the NRC meeting. Revisions are being prepared and the revised specifications will be submitted to the'NRC by May 30, 1989.

The three MERITS TS with parameter no longer proposed for inclusion in the COLR, specifications 2.1.1, 3.4.1, and 3.4.5, have been revised and are included in Attachment 5.

The revised specifications also include changes required in response to Item 2 in the NRC letter with regard to justifications referencing plant-specific licensing actions.

1 Specification 3.3.3, RCS Pressure / Temperature Limits, will be revised and submitted to the NRC at a later date after a WOG decision has been.iade on the preferred mechanism to address this issue.

A draft COLR report will also be developed and submitted to the NRC by May 30, 1989.

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'0G-89-36 May 22, 1989-ATTACHMENT 4-OUESTION SOURCE 1.

l NRC letter from C. E. Rossi to R. A. Newton, NRC Review of the Westinghouse Owners Group MERITS Technical' Specifications, April-28, 1989.

OUESTIONS 1

RE0VIRED INFORMATION TO FULLY PROGRESS WITH THE REVIEW OF~THE WOG'S PROPOSED NEW STS 5.

The "NRC Staff Review of NSSS Vendor Owners Groups' Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications," dated May 9,.1988, Appendix B, Table 2, lists Westinghouse STS LCO's which may be relocated outside of new STS. However, a number of these LCO's contain qualifying notes.

It does not appear as if the WOG observed all of these notes in developing their proposed new STS. For example, the current LC0 3.6.1.7, " Containment Structural Integrity,"'

contains Note 2.

Note 2 says in part, "...if the associated Surveillance Requirement (s) is necessary to meet the OPERABILITY requirements for a retained LCO, the Surveillance Requirement (s) should be relocated to the retained LCO."

In the cited example the WOG did not include the surveillance requirements of LCO 3.6.1.7 in the new STS; neither did they provide justification as to

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why the surveillance requirements were not included. The staff decision is that the surveillance requirements for prestressed concrete containments must be relocated to the new STS LCO 3.5.1, Containment Intearity. The definition for CONTAINMENT INTEGRITY must be revised accordingly.

Another example involves Note 4 on LCO 3.4.5, Steam Generators.

i The WOG must provide complete, revised technical specifications for not only the situations described above but also for the other situations where the notes indicate that requirements are to be retained.

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OG-89-36 May 22, 1989 ATTACHMENT 4 Reference 2 provided a comprehensive, marked up version of the current Westinghouse STS (Rev 4a) in response to the information-requested in Item 1 of Reference 1.

This mark-up provides a detailed roadmap identifying the -

relationship between Sections 1, 2, 3, and 4 of the current STS and the corresponding Division?l, 2, and 3 of the MERITS Technical Specifications (TS).

This roadmap, together with its amplifying notes, identifies the locations of information and requirements that are to be retained by the qualifying notes in Table 2, Appendix B, of the "NRC Staff Review of NSSS Vendor Owners Groups'.

Application of The Commission's Interim Policy Statement Criteria To Standard Technical Specifications" (NRC Split Report).

The following pages provide the changes to the MERITS Technical Specifications, and additions / changes to Design Features and Administrative Controls, necessary to incorporate the qualifying notes in the NRC Split Report.

Only those instances where the disposition of the qualifying note was not adequately described and justified in the WOG MERITS submittal are included.

Where a particular item resulted in a decision different from or contrary to the note, justification is provided. The items are arranged by STS LCO number, and the qualifying note is restated for completeness.

The MERITS Phase 11 task did not include rewriting of the Standard Technical l

Specifications (STS) Sections 5.0, Design Features, and 6.0, Administrative Controls. Therefore, changes to the Design Features and Administrative Controls sections are provided as additions / modifications by the appropriate title currently in the STS Sections 5.0 and 6.0.

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ATTACHMENT 4 OVESTION NRC Split Report, Note 4 (LC0 3.4.5, Steam Generators)-

'This LCO may be relocated out of Technical Specifications. However, the associated Surveillance Requirement (s) must be relocated to Technical Specification Section 4.0, Surveillance Requirements.

EfSPONSE The WOG is proposing that the surveillance requirements not be relocated to the Surveillance Requirement section of the MERITS specification (equivalent to Technical Specification Section 4.0, Surveillance Requirements, in the Rev. 4a STS).

Instead, the WOG proposes that the surveillance requirements _for specification 3.4.5 be included in a Steam Generator Tube Inspection Program.

This program will incorporate the requirements presently in the surveillance requirements for LCO 3.4.5, Steam Generator. The elements of this program will be delineated by a paragraph in the Administrative Controls Division. Any changes to this program will require prior Commission approval.

Technical Specification Section 4.0, Surveillance Requirements, in the Rev. 4a STS (and the corresponding Division 3.0 in the MERITS TS) contains general statements about surveillance requirements and not the detailed requirements r-themselves. The steam generator tube inspection surveillance requirements of LC0 3.4.5 would add'a level of detail which would be inconsistent with the f(

content of the Applicability Sections.

These detailed surveillance requirements are also more detailed than those normally found in the Surveillance Requirements Section of the individual MERITS Technical Specifications. The MERITS specifications have been substantially improved to facilitate operator utilization and including these surveillance in a program would maintain the effort to orient the TS for operator use.

Therefore, the WOG has concluded that incorporating the steam generator tube inspection surveillance requirements in a program is more appropriate. The detailed requirements in LCO 3.4.5 of the Rev. 4a STS would be maintained in the program and the elements of the program are defined in the proposed addition to the Administrative Controls section that follows.

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OG-89-36 May 22, 1989 ATTACHMENT 4 RESPONSE (continued)

This is to be added to-Administrative Controls - PROCEDURES AND PROGRAMS:'

Steam Generator Tube Insoect'on Proaram A program which will ensure that the surveillance requirements for steam generators are performed to maintain their OPERABILITY.. The program shall include' the following:

1)

Steam generator sample selection 2)

Steam generator tube sample selection and inspection a)

Minimum sample size b)

Inspection result classification c)

Actions required 3)

Inspection frequencies 4)

Acceptance Criteria 5)

Reports I

Changes to the Steam Generator Tube Inspection Program shall be implemented only upon prior approval by the Commission.

1 OVESTION NRC Split Report, Note 4 (LC0 3.4.10, RCS Structural Integrity)

This LC0 may be relocated out of Technical Specifications. However, the associated Surveillance Requirement (s) must be relocated to Technical Specification Section 4.0, Surveillance Requirements.-

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RESPONSE

l The WOG is proposing to relocate the surveillance requirements for RCS l

Structural Integrity in a similar fashion to that proposed for the requirements I

for Steam Generator Tube Inspection. Therefore the RCS Structural Integrity surveillance requirements would not be relocated to the Surveillance Requirement section of the MERITS specification (equivalent to Technical Specification Section 4.0, Surveillance Requirements, in the Rev. 4a STS).

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OG-89-36 May 22, 1989 ATTACHMENT 4

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RESPONSE (continued)

Instead, the WOG proposes that the surveillance requirements for specification 3.4.5 be included in a Reactor Coolant Pump Flywheel Inspection Program and I

provides the following for consideration by the NRC staff. This program will l

incorporate the requirements presently in the surveillance requirements for LCO 3.4.10, RCS Structural Integrity.

The MERITS specifications have been substantially improved to facilitate operator utilization and including these surveillance requirements in a program would maintain the effort to orient the TS for operator use. Therefore, the WOG has concluded that their incorporation in a program is more appropriate.

The detailed requirements of the surveillance would be maintained in the program and the requirement for the program would be contained in the Administrative Controls Division, as delineated by the following paragraph.

This is to be added to Administrative Controls - PROCEDURES AND PROGRAMS:

Reactor Coolant Pumo Fivwheel Inspection Procram A program for inspecting each reactor coolant pump flywheel per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

OVESTION i

NRC Split Report, Note 5 (LCO 3.6.1.2, Containment Leakage)

This LC0 may be relocated. However, P ' l ' l, and Lt must either be l

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retained in TS cr in the Bases of the appropriate containment LCO.

RESPONSE

The WOG proposes that the leak test parameters listed in the note be retained in the Design Features Section of the STS. Since the definition of CONTAINMENT INTEGRITY was modified to require containment leakage rates within limits, SR 3.5.1.5, which requires verification of the containment leakage rates, was included in MERITS specification 3.5.1.

The details of the leak rate testing surveillance will be included in a program. The requirements of this program will be identified in the Administrative Controls Division. The following additions to the Design Features and Administrative Controls Division will provide compliance with Note 5:

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.l ATTACHMENT 4 RESPONSE (continued)

This is to be added to Design Features:

CONTAINMENT LEAK RATE TEST PARAMETERS The containment leak rate test parameters are as follows:-

a.

Calculated peak DBA interior pressure, Pa " I 3 psi 9-b.

Maximum allowable leakage rate, La = ['

.]% by weight"of the containment air.per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P -

a c.

Design leakage rate, Ld=[

]% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 'at P -

a d.

Maximum allowable leakage rate, Lt=[

]% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at-a reduced pressure of Pt[ ] psi 9

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OG-89-36 May 22, 1989 l

ATTACHMENT 4 RESPONSE (continued)

This is to be' added to Administrative Controls - PROCEDURES AND PROGRAMS:

i Containment Leak Rate Test Proaram A program which will ensure that the containment leak rate tests are performed to maintain the leak rate part of the definition of CONTAINMENT INTEGRITY. The program shall include the following surveillance in accordance with 10 CFR 100, Appendix J:

1)

Type A tests (Overall integrated containment leakage rate).

2)

Type B tests (Local penetration leak rates).

3)

Type C tests (Containment isolation valve leakage rates).

4)

Air lock seal leakage and air lock overall leakage rates.

5)

Isolation valve and channel weld pressurization system pressure verifications.

6)

[ ]-inch purge supply and exhaust leakage rates.

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OG-89-36 May 22, 1989 ATTACHMENT 4 t

OUESTION NRC Split Report, Note 2 (LCO 3.6.1.7, Containment Structural Integrity)

L This LCO may be removed from the STS. However, if the associated Surveillance Requirement (s) is necessary to meet the OPERABILITY requirements for a retained' LCO, the Surveillance Requirement (s) should be relocated to the retained LCO.

RESPONSE

The WOG has concluded that the containment structural integrity surveillance requirements are not necessary to meet the new definition of CONTAINMENT INTEGRITY for plants with reinforced concrete or steel shell (dual and ice condenser) containments. For'these containments, the Containment Leak Rate Test Program verifies the leak rate requirements for CONTAINMENT INTEGRITY.

The WOG agrees that for post-tensioned prestressed concrete containments, the surveillance requirements' for the tendons should be retained. The WOG proposes to add a Containment Structural Integrity Test Program to the~ Administrative Controls Section, as delineated in the added paragraphs below. Also, SR 3.5.1.9 will be added to MERITS specification 3.5.1. lhis SR will, for the applicable plants, require verification of containment structural integrity,-in accordance with the Containment Structural Integrity Test Program.

MERITS TS 3.5.1 has been revised to include the new SR 3.5.1.9 and the revised specification is provided in Attachment 5 of this transmittal.

This is to be added to Administrative Controls - [ UNIT REVIEW GROUP (URG)]:

RESPONSIBILITIES Review of the Containment Structural Integrity Test Program and implementing procedures, and the submittal of recommended changes to the [ Company or unit Nuclear Review and Audit Group].

This is to be added to Administrative Controls - PROCEDURES AND PROGRAMS:

Containment Structural Integrity Test Program Implementation.

The WOG proposes to maintain the definition for CONTAINMENT INTEGRITY provided in WCAP-12159. With the retention of SR 3.5.1.9 and the implementation of ~this program, the containment integrity requirements and the existing CONTAINMENT INTEGRITY definition in the MERITS TS are consistent with those in the current Rev. 4a.STS.

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OG-89-36 May 22, 1989-

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ATTACHMENT 5

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REVISED SPECIFICATIONS. BASES. AND JUSTIFICATIONS l

This attachment contains the specifications, bases, justifications, and/or l

mark-ups that have been revised to support the response to the iters in the NRC letter. The revised versions are organized by cl. apter.

Table 1 provides a listing of those technical specifications that were revised and the NRC request associated with the change.

Each specification, bases, justification, and mark-up is considered as an individual document for the purpose of revising the MERITS submittal. The revision process and the files for the individual documents have been structured so that whenever a change is made to any document, a complete draft of that document (all pages, including the revised pages with vertical lines.

indicating changes from the previous version) will be provided and all sages of the affected document will be marked with the same revision letter in tie l

footer. All revised documents will include vertical lines in the right margin for those document pages which contain changes, and the revision letter will be written next to the line.

Since the STS markup documents are not on word processor files, their current revision letter will be handwritten at the top of the marked-up page.

When a document is revised, its associated documents will not be revised or submitted to the NRC unless there are specific changes required to be made to these other documents as a result of the change to the original document.

For example, the original Revision A STS markup may not have changed when the l

associated Revision B justification was prepared in response to Item 2 of Reference 1.

Also attached is a List of Effective Revisions (LOER) for each specification, bases, justification, and STS mark-up. The current revision for each document will be provided in an updated List of Effective Revisions whenever revisions are made to the MERITS submittal. The columns in the LOER provide-the applicable revisions for each technical specification, bases, justification, and mark-up, reflecting the changes due to the associated submittal.

If there is no letter in the revision column, then the Revision A version in the MERITS submittal of March 30, 1989 is still the applicable version.

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l-OG-89-36 May 22, 1989 O

ATTACHMENT 5

,d Table 1 1

Semmary of Revised HER1TS TS NRC Item Specification Number 2.1.1 Reactor Core Safety Limits 4

3.1.2 Moderator Temperature Coefficient 4*

3.1.3 Rod Group Alignment Limits 2,

3.1.6A Heat Flux Hot Channel Factor - Fg(Z) (FXY Methodology) 4 3.1.68 Heat Flux Hot Channel Factor - Fg(Z) (Fg Methodology) 2,4*

N 2,4*

l 3.1.7 Nuclear Enthalpy Rise Hot Channel. Factor - FAH 3.1.8B Axial Flux Difference (RAOC Methodology) 2 3.1.9 Quadrant Power Tilt Ratio 2-3.2.1 Reactor Trip System Instrumentation 2

3.2.2 Engineered Safety Features Actuation System 2

Instrumentation 3.2.3 Accident Monitoring Instrumentation 2

3.2.4 Remote Shutdown System

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3.3.1 RCS Pressure,: Temperature, and Flow DNB Limits 2

3.3.9 RCS Loop Isolation Valves 2

3.3.10 RCS Isolated Loop' Startup 2

l 3.3.16 RCS Leakage Detection Instrumentation 2

3.4.1 Accumulators 2,4 3.4.2 ECCS Trains - T 2 350*F 2

avg 3.4.5 Refueling Water Storage Tank 4

3.5.1 Containment Integrity 1,5 3.5.2 Purge and Exhaust Isolation Radiation Instrumentation 1

(Atmospheric, Dual, Ice Condenser) l 3.7.1 AC Sources - Operating 2

3.7.2 AC Sources - Shutdown 2

1 3.8.1 Boron Concentration 4*

3.8.4 Containment Building Penetrations 1

3.8.5 Residual Heat Removal and Coolant 2

Circulation - High Water Level

Specification with COLR reference that will be submitted by May 30, 1989.

\\

Page 2

?

1

' MERITS TECHNICAL SPECIFICATIONS.

)

BASES. AND JUSTIFICATIONS j

i

/]

LIST OF EFFECTIVE REVISIONS O

Applicable

. Revisions

I B

d M

1.0 'USE AND APPLICATION 1.1 DEFINITIONS...............................................

l 1.2 LOGICAL' CONNECTORS........................................

l

- 1.3 COMPLETION TIMES..........................................

1 1

1.4 LEGAL CONSIDERATIONS......................................

2.0 SAFETY LIMITS' 2.1 ' REACTOR SAFETY LIMITS 2.1.1 Reactor Core Safety Limits.......................

B B

2.1.2 Reactor Coolant System Pressure Safety Limit.....

3.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS - APPLICABILITY.................

3.1 REACTOR 3.1.1 Shutdown Margin.......................-...........

3.1.2 Moderator Temperature Coefficient................

{

f 3.1.3

-Rod Group Alignment Limits.......................

B i

(

3.1.4 Shutdown Bank Insertion Limit....................

3.1.5 Control Bank Insertion Limits....................

3.1.6A Heat Flux Hot Channel Factor - F (Z).............

1 g

(F y Methodology)

)

X 3.1.6B Heat Flux Hot Channel Factor - F (Z).............

B g

3 (F Methodology) 1 Nuklear Enthalpy Rise Hot Channel Factor - FAH.. ~

B B

3.1.7 3.1.BA

. Axial Flux Difference (CAOC Methodology).........

3.1.BB Axial Flux Difference (RA0C Methodology).........

.B 3.1.9 Quad ra n t Powe r T il t R a t i o............................

B B

3.1.10 Rod Position Indication..........................

3.1.11 Mode 1 Physics Tests Exceptions..................

3.1.12 Mode 2 Physics. Tests Exceptions...................

4 i

Rev. A if unmarked

-I - Technical Specification B - Bases J - Justification t

M - Marked-up STS j

WOG-MERITS 5/22/B9 Page 1

?

MERITS TECHNICAL SPECIFICATIONS.

BASES. AND JUSTIFICATIONS a

l D

LIST OF EFFECTIVE REVISIONS d'

f

. Applicable i

Revisions

1H J

M

'3.2 INSTRUMENTATION j

3.2.1 Reactor Trip System Instrumentation..............

B 3.2.2 -

Engineered Safety. Features Actuation System Instrumentation..;......................-..........

B 3.2.3 Accident Monitoring Instrumentation..............

B 3.2.4 Remote Shutdown System...........................

.B.

3.3 REACTOR COOLANT SYSTEM (RCS)-

3.3.1 RCS Pressure, Temperature, and Flow DNB Limits..

B B

3.3.2 RCS Minimum Temperature for Criticality...........

3.3.3 RCS Pressure / Temperature Limits..................

3.3.4 RCS Loops - Modes 1 and 2........................

3.3.5 RCS Loops - Mode 3...............................

3.3.6 RCS Loops - Mode 4...............................

3.3.7 RCS Loops - Mode 5, Loops filled.................

3.3.B RCS Loops - Mode 5, Loops Not filled.............

3.3.9 RCS Loop Isolation Valves........................

B

.B 3.3.10 RCS Isolated Loop Startup.........................

B B

3.3.11 Pressurizer......................................

3.3.12 Pressurizer Safety Va1ves........................

3.3.13 Pressurizer Power-0perated Relief Valves.........

3.3.14 RCS Operati onal Leakage...........................

3.3.15 RCS Pressure Isolation Valve Leakage.............

3.3.16 RCS Leakage Detection Instrumentation............

B 3.3.17 RCS Specific Activity............................

3.3.18 Cold Overpressure Prevention.....................

3.3.19 RCS Loops - Test Exceptions......................

3.4 EMER6ENCY CORE COOLING SYSTEM (ECCS) 3.4.1 Accumulators.....................................

B B

B i

.]

3.4.2 ECCS Trails - T 2 350*F.......................

B' B

avg 3.4.3 ECCS Trains - T,yg < 350*F.......................

3.4.4 Seal Injection Flow..............................

3.4.5 Refueling Water Storage Tank..................... B

.B B

3.4.6 Boron Injection Tank..............................

s Rev. A if unmarked I - Technical Specification B - Bases

'a J - Justification M - Marked-up STS WOG-MERITS 5/22/89 Page 2 h

-l 1

1 MERITS TECHNICAL SPECIFICATIONS 1 BASES. AND JUSTIFICATIONS

- LIST OF EFFECTIVE REVISIONS Applicable Revisions

I.

Bd 5

3.5 CONTAINMENT' SYSTEMS 3.5.1 Containment' Integrity............................

B' B

B B

3.5.2 Purge and Exhaust Isolation Radiation Instrumentation..................................

B B

(Atmospheric,. Dual, Ice Condenser)-

3.5.3A Containment Spray System (Ice Condenser).........

3.5.3B. Quench Spray System.(Subatmospheric).............

3.5.3C Containment. Spray and Cooling Systems.............

(Atmospheric, Dual)

(Credit taken for iodine removal) 3.5.30 Containment Spray and Cooling Systems............

(Atmospheric, Dual)

(No credit taken for iodine removal)-

3.5.4 Recirculation Spray System (Subatmospheric)......

3.5.5 Air Return System (Ice Condenser)................

3.5.6 Spray Additive System............................

3.5.7 Vacuum Relief Va1ves.............................

3.5.8 Ice Bed (Ice Condenser)..........................

3.5.9 Ice Condenser Doors-(Ice Condenser)..............

3.5.10 Divider Barrier Integrity (Ice Condenser)........

3.5.11 Containment Recirculation Drains (Ice Condenser).

3.5.12A Containment Internal Pressure....................

(Atmospheric, Dual, Ice Condenser).

3.5.12B Containment Internal Pressure (Subatmospheric)...

3.5.13A Containment Air Temperature (Ice Condenser)......

3.5.13B Containment Air Temperature (Atmospheric, Dual)..

3.5.13C Containment Air Temperature (Subatmospheric)....

3.5.14 Iodine Cleanup System............................

3.5.15 Penetration Room Exhaust. Air Cleanup System......

3.5.16 Shield Building Air Cleanup System...............

(Dual, Ice Condenser) 3.5.17 Hydrogen Monitors................................

3.5.18A Hydrogen Recombiners (Internal)..................

3. 5.18B Hydrogen Recombiners (External)..................

3.5.19 Hydrogen Mixing System.............................

3.5.20 Hydrogen Ignition System (Ice Condenser).........

l I

l

Rev. A if unmarked I - Technical Specification B - Bases J - Justification 5 - Marked-up STS

' WOG-MERITS 5/22/39 Page 3 1

l l

l i

MERITS TECHNICAL SPECIFICATIONS.

i BASES. AND JUSTIFICATIONS

/O LIST OF EFFECTIVE REVISIONS

'Y Applicable Revisions

I B

J B

3.6 PLANT SYSTEMS i

3.6.1 Main Steam Safety Valves - Thermal Power 2 10% of RTP...............................

3.6.2 Main Steam Safety Valves - Thermal Power < 10% of RTP...............................

3.6.3 Main Steam Line Isolation Valves.................

3.6.4A Main Feedwater Isol ation Val ves..................

3.6.4B Main Feedwater Regul ation Val ves.................

3.6.5 Auxiliary Feedwater System.......................

3.6.6 Condensate Storage Tank..........................

3.6.7 Secondary Coolant Specific Activity..............

3.6.8 Component Cooling Water System...................

3.6.9 Service Water System.............................

3.6.10 Ultimate Heat Sink...............................

3.6.11 Control Room Emergency Air Cleanup System........

3.6.12 ECCS Pump Room Exhaust Air Cleanep System........

3.6.13 fuel Storage Pool Water Level....................

3.6.14 Fuel Building Ai r Cleanup System.................

/7 3.7 ELECTRICAL SYSTEMS i

n 3.7.1 AC Sources - 0perating...........................

8 3.7.2 AC Sources - Shutdown............................

B 3.7.3 DC Sources - Operating...........................

3.7.4 DC Sources - Shutdown............................

3.7.5 Distribution Systems -

Operating.................

3.7.6 Distribution Systems -

Shutdown..................

I 3.8 REFUELING OPERATIONS j

j 3.8.1 Boron Concentration..............................

)

3.8.2 Unborated Water Source Isolation Valves..........

~

3.8.3 Nuclear Instrumentation..........................

3.8.4 Containment Building Penetrations................

B B

3.8.5 Residual Heat Removal and Coolant I

Circul ation - High Water level...................

B I

3.8.6 Residual Heat Removal and Coolant Ci rcul ation - Low Water Leve1....................

3.8.7 Refueling Cavity Water Level - High..............

3.8.8 Refueling Cavity Water Level - Low...............

3.8.9 Decay Time.......................................

, Rev. A if unmarked I - Technicai Specification H - Bases I

J - Justification

(

B - Marked-up STS WOG-MERITS 5/22/89 Page 4 i

Reactor C re Safety Li;its 2.1.1 2.1 SAFETY LIMITS 2.1.1 Reactor Core Safety Limits SL 2.1.1 The combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer 6

pressure shall not exceed the Safety Limits shown in Figure 2.1.1-1.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE----------

Required Action must be completed whenever Condition A is entered.

Safety Limit exceeded.

A.1 Be in MODE 3.

I hour-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY None.

CROSS-REFERENCES l

TITLE NUMBER l

Safety Limit Violation 5.7

(

\\

l Unit Name 2.1-1 Amendment WOG-MERITS Rev. B t

i

Reactor Core Safety Li:its 2.1.1 O

680 00 NOT OPERATE

' 2385 PSIG IN THIS AREA g-N g

This Figure For 640 Illustration Only s N Do NOT use for 2235 PSIG N+

Operation

=

620

\\

'd 198 0

n N

N s

en 600 b"

168B PSIG #

D, Q>

580 3

E2

\\

[

560 u

ACCEP"ABLE OPERAT:ON 540 s

520 0

20 40 60 80 100 120 Percent of RATED THERMAL POWER Figure 2.1.1-1 (Page 1 of 1)

Reactor Core Safety Limits Unit Name 2.1-la Amendment WOG-MERITS Rev. B f

Reactor Core Safety Limits B 2.1.1 B 2.1 REACTOR B 2.1.1 Reactor Core Safety Limits BASES BACKGROUND The restrictions of this safety limit prevent overheating of the fuel and possible cladding failure which would result in the release of fission products to the reactor cool ant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the cladding surface temperature is slightly above the coolant saturation temperature.

The proper functioning of the Reactor Protection System and steam generator safety valves prevent violation of the Reactor Core Safety Limits.

APPLICABLE Operation above the upper boundary of the nucleate boiling SAFETY ANALYSES regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter, Q

however THERMAL POWER and Reactor Coolant Temperature and j

Pressure have been related to DNB through the [

]

l correlation (Ref.1). The [

] DNB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

Automatic enforcement of these Reactor Core Safety Limits I

are provided by the following functions:

a.

High Pressurizer Pressure Reactor Trip b.

Low Pressurizer Pressure Reactor Trip c.

Overtemperature Delta-T Reactor Trip d.

Overpower Delta-T Reactor (rip e.

Power Range High Neutron Flux Reactor Trip l

f.

Steam Generator Safety Valves General Design Criteria 10 (Ref. 2) requires that the minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients be limited to [

).

This value corresponds to a 95 (continued)

O Unit Name B 2.1-1 Revision WOG-MERITS Rev. B

Reactor Core Safety Liaits B 2.1.1 BASES APPLICABLE percent probability at a 95 percent confidence level that SAFETY ANALYSES DNB will not occur and is chosen as an appropriate margin (continued) to DNB for all operating conditions.

The limitation that the average enthalpy in the hot leg be equal to the enthalpy of saturated liquid also ensures that the " Delta-T" measured by instrumentation (used in the protection system design as a measure of the core power) is proportional to core power.

It is implicitly assumed in the generation of the core -

limits and in the operation of the protection functions that the minimum RCS flow requirement of LCO 3.3.1 is satisfied.

The Reactor Core Safety Limits represent a design requirement for establishing the protection system trip setpoints identified above. They are not as restrictive as the conditions of LCO 3.3.1, RCS Pressure, Temperature, and Flow DNB Limits, or the assumed initial conditions of the safety analyses (as indicated in Chapter [15] of the FSAR).

Figure B 2.1.1-1 provides a conceptual view of the protection provided to prevent violation of the core limits.

O SAFETY The curves provided in Figure 2.1.1-1 show the loci of LIMIT points of THERMAL POWER, Reactor Coolant System pressure, and average temperature for which the minimum DNBR is not lessthan[

), or the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or the exit quality is within the limits defined by the [

-]

q DNBR correlation.

I The curves are based on enthalpy hot channel factor limits provided in the CORE OPERATING LIMITS REPORT. The dashed line of Figure B 2.1.1-1 shows an example of a limit curve at 2235 psig. In addition, it illustrates the various prote: tion system functions that are designed to prevent the unit from reaching the limit.

The heat flux Safety Limit is higher than the limit calculated when the AXIAL FLUX DIFFERENCE (AFD) is within.

the limits of the f (AI) function of the Overtemperature 3

Trip. When the AFD is not within the tolerance, the AFD effect on the evertemperature AT trips will reduce the fatpoints to provide protection consistent with core safety limits.

(Ref. 3,4)

O (continued)

Unit Name B 2.1-2 Revision WOG-MERITS Rev. B i

l Reactor' Core Safety Lisits B 2.1.1 BASES (continued)

APPLICABILITY Safety. Limit 2.1.1 only applies in MODES 1 and 2 because these are-the only MODES in which the reactor is critical.

Automatic protection functions are required to be OPERABLE during MODES 1.and 2 to assure. operation within the core limits. The. steam generator safety valves or. automatic j

protection actions serve to_ prevent Reactor Coolant System '

heatup to the core limit conditions or to initiate a reactor trip function (which forces the unit into MODE 3).

Setpoints for the reactor trip functions are specified in LCO 3.2.1 and LCO 3.2.2..

i ACTIONS L1 I

When the Reactor Core Safety Limits are exceeded the unit must be brought to MODE 3 where the safety limit is not-applicable. Operation above the upper boundary of the j

nucleate boiling regime could result in excessive cladding j

temperatures because of the onset of departure from nucleate boiling and lead to cladding failure. Similarly, operation with bulk hot leg boiling could limit the effectiveness of the overtemperature/ overpower AT protection O

system. A Completion Time of one hour is adequate to -

accomplish an orderly shutdown.

SURVEILLANCE Surveillance Requirements are met through executing REQUIREMENTS the Surveillance Requirements specified for the Reactor Protection System and steam generator safety valves.

If maintained OPERABLE, the Reactor Protection System and steam generator safety valves will not permit violation of Reactor Core Safety Limits.

REFERENCES 1.

[ North Anna Updated] FSAR, Section [4.4]

2.

Title 10 Code of Federal Regulations (10CFR), Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, 1988.

3.

WCAP 8746-A, Design Bases for the Overtemper-ature AT and the Overpower AT Trips, March 1977 i

(continued) i 1

O l

Unit Name B 2.1-3 Revision I

WOG MERITS Rev. B i

Reactor Core Safety Limits B 2.1.1 BASES REFERENCES (continued) 4.

WCAP 9273-NP-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.

O 1

e O

Unit Name B 2.1-4 Revision WOG-MERITS Rev. B

Reactor Core Safety Liaits B 2.1.1 f

680 OT DELTA T OP DELTA T UNACCEPTABLE

/

N OPERATION N

NUCLEAR OVERPOWER 2235 psig

_/

\\

640 3

' m

\\

N N

~

Y s

x, 600 7

[

S/G SAFE" W

580 J

m n

u e

i 560 THIS FIGURE FOR ILLUSTRATION ONLY. DO NOT USE FOR OPERATION 540 ACCEPTABLE OPERAT: ON 520 0

20 40 60 80 100 120 Percent of RATED THERMAL POWER Figure B 2.1.1-1 (Page 1 of 1)

Reactor Core Safety Limits vs Boundary of Protection O

Unit Name B 2.1-5 Revision WOG-MERITS Rev. B

-___-m_.m.- _ __ -.. _ _ - _ _ _ _ _ _ _ _ _ _ _ -. _ _ _.

Reactor Core Safety Licits 2.1.1 I

i JUSTIFICATION FOR CHANGES TO W-STS REV. 5 i

The following justifications are provided to explain the reasons for the differences between Desk Reference, W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each justification corresponds to a number appearing on the attached, marked up 1

Specification 2.1.1, Safety Limits - Reactor Core.

1.

The modifier " Reactor Coolant System" is inserted for clarity.

l Reference to N-1 loop operation is not required, since the MERITS Technical p 2.

Specification will not cover N-1 loop operation.

3.

Specific' reference to reporting requirements has been deleted from the action section of the specification since it does not enhance the safety of the plant. The reporting requirements are required by the, " Safety Limit Violation," section of the Technical Specifications. General reference to

.i reporting requirements is, however, still retained in the CROSS-REFERENCES l

of the MERITS specification.

A figure is also provided with the MERITS specification bases as an example g 4.

of how the limits are typically specified.

1 i

WOG MERITS Page 1 of 1 Rev. B

Rev, g 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE gg c,,g,_3 The combination of THERMAL POWER, pressurizer pressure, and the\\p highest 2.1.1 operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figures 2.1-1 and 2.1 2 foi s. end a i h;;; ;=tf::, n:;tetive'yh@

b

~

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within I hour, 2nd ::O y eft' *l e - Si r -

Of !;;;f f f::tt:n 5.'.1. g

nt:

REACTOR COOLANT SYSTEM PRE'SURE S

2.1.2 The Reactor Coolant System pressure shall not exceed [2735] psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded [2735] psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded [2735] psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

1 l

l 1

s' CDMANCHE PEAK - UNIT 1 2-1

+ G 9'

. Re v, 9 e

e t

O t

t 4

e 4*

4 FIGURE 2.1-1 REACTOR CORE SAFETY. LIMIT - FOUR LOOPS IN OPERATION G

COMANCHE PEAK - UNIT 1 2-2

____m.

e

~

O 9

9 e

' 9 e

e 9

e O

~

O

~

I 4

e FIGURE REACTOR CORE AF IMIT - TH LOOPS IN OPERATION

~

O COMANCHE PEAK - UNIT 1 2-3

_______.__.m__

O O

i i

4 8

1 i

i.

,5 t

i J

'd I

1 I

p '

i I

I 1

I I

J l

1 I

O 1

l.

i n

1 l

- _ _ = _ _ _ _ - - _ _ _ _ _ _ _

I

Rod Group Alignment Limits 3.1.3 JUSTIFICATION FOR CHANGES TO M-STS REY. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference, W-STS Rev. 5, the base Standard Technical Specification, and this MERITS specification. The number for each written ju!.tification corresponds to a number appearing on the attached, marked up Specification 3.1.3.1, Movable Control Assemblies - Group Height.

1.

The term Rod Cluster Control Assembly (RCCA) will be used along with

" rod" to refer to an' individual control element. Often the term " Control Rods" is used to refer to a control bank of RCCAs and at other times

" Control Rod" may-refer either to a control bank or a single RCCA.

Without the modifier of " shutdown" or " control", an RCCA can be in either i

a shutdown bank or a control bank.

2.

All use of the term full-length has been removed since Westinghouse plants no longer have part-length RCCAs.

3.

The asterisks next to the 1 and the 2 of the APPLICABILITY section and the corresponding explanation at the bottom of the page have been replaced by a Cross Reference to these test exceptions.

O 4.

The requirements of the W-STS Specification 3.1.1.1 SHUTDOWN MARGIN in l V MODES 1 and 2 have been relocated to administrative controls outside of Technical Specifications.

In addition, the MERITS version of.the LCOs for control bank insertion limits and shutdown bank insertion limit ensure the existence of adequate SHUTDOWN MARGIN. Because the requirement for verifying SHUTDOWN MARGIN per specification 3.1.1.1 is-not applicable in these MODES, the conditions and the betions required to ensure adequate SHUTDOWN MARGIN is available are contained within this specification.

5.

This section corresponds to Condition A of the MERITS Specification.

6.

This section corresponds to Condition E of the-MERITS Specification.

7.

This section corresponds to Condition B of the MERITS Specification.

8.

Permission for continued power operation is implied by completing the Required Actions within their specified Completion Times.

9.

These limits will be provided in the CORE OPERATING LIMITS REPORT (COLR).

't WOG-MERITS Page 1 of 2 Rev. B l

L

f t

Rod Gr up Alignment Liaits 3.1.3-10.

The MERITS. version contains a condition which allows. continued power operation with more than one RCCA inoperable but trippable. This condition is most likely to exist.due to a control system malfunction and would not prevent the control banks from dropping into the core in the event of an accident requiring a reactor trip. This' generic change to-NUREG-0452 Rev,4 Specification 3/4.1.3, Movable Control Assemblies and its associated Bases, were previously submitted to the NRC by Westinghouse through letter NS-NRC-84-2990, December 21, 1984.

Subsequently, the NRC has approved the change to several-plant technical specifications including Millst)ne Unit 3, Seabrook,' Vogtle, Comanche g

Peak, and South Texas.

This corresponds to Condition C of the MERITS Specification.

11.

The surveillance requirement has been reworded.to include the limits and to correct errors.

12.

Surveillance requirements from W-STS Rev. 5, 3.1.3.4 Rod Drop Time, have been relocated into this specification in accordance with the split decision criteria for the MERITS progrcm; certain LCOs may be removed from Technical Specifications, but their corresponding Surveillance are to be retained.

O O

WOG-MERITS Page 2 of 2 Rev. B I

Heat Flux Hot Channel Factor - Fq(Z) 3.1.6B

(

JUSTIFICATION FOR CHANGES TO SPECIFICATION FOR CATAWBA The following justifications are provided to explain the reasons for the I

differences between the specification for Catawba, which is the base for this-particular specification, and this MERITS specification. The number for each written justification corresponds to a number appearing on the attached, marked up specification 3.2.2 Heat Flur. Not Channel Factor - F (Z).

q There are currently in use two methodologies for the verification that the total power peaking factor, Fg(Z), is not exceeded. For completeness both methodologies have been included in the MERITS Technical Specifications.

One methodology entails the surveillance of the height dependent radial peaking factor, Fxy(Z), as verification that operation will not cause the i

Fn(Z) limit to be exceeded. MERITS Specification 3.1.6A uses a Fxy Surveillance and is based on STS Rev. 4a.

The second methodology entails the measurement of the steady state F (Z) and g

adjustment of the measured values by a factor, W(Z), which accounts for plant maneuvers within the restrictions on axial flux difference and rod insertion limits. This particular specification, denominated F (Z) Surveillance, is g

l MERITS specification 3.1.6B.

To show the Fg(Z) Surveillance, as currently I

applied on operating plants, the Technical Specification for a plant I

utilizing this methodology had to be used, and the Catawba technical specification was selected. The parts of the plant specific Catawba specification pertaining to Sase Load (BL) have been deleted for simplicity.

Documentation on the methodologies was subcitted to the Nuclear Regulatory Commission (WCAP - 10216-P-A, under cover letter NS-EFR-2649) in August 1982.

In February 1983 Westinghouse received from the NRC office of Standardization &

Special Projects Branch, Division of Licensing a confirmation that the report was approved and that the F (Z) Surveillance Methodology was accepted for q

referencing in license applications. A total of twelve plants use F (Z) g Surveillance type of technical specification, including Catawba and McGuire.

S l

f WOG-MERITS Page 1 of 3 Rev. B

Heat Flux Hot Channel Factor' Fg(Z) 3.1.6B O

I.. The limits for Fq(Z) will be given in the CORE OPERATING LIMITS REPORT, including the K(Z) curve.

2.

In this base specification the term Fg(Z) represents the meas'ured value of the core. peaking factor under steady-state condition adjusted to allow fordesinandmeasurementuncergainties. The steady-state adjusted-peaking actor is denominated Fg in the MERITS,;which is the factor-l compared to the limits-to assess whether a violation has occurred.

g 3.

The cause of the out-of-limit condition should be. corrected if it can be identified. However, some causes, e.g., a core loading which gives peaking factors outside limits, may be identified but cannot be easily corrected.

In either case, the restrictions on THERMAL POWER if limits are exceeded.

i are sufficiently conservative to warrant continued operation.

An action has been added. the MERITS Required Action A.4 requires flux mapping to demonstrate that Fg(Z) is within its limits before increasing THERMAL POWER.

4.

In addition to checking for limit violations under steady state operation (see item two above), it is also necessary to evaluate the potential to exceed peaking factor limits in the event of a normal load transient. A factor W(Z) is analytically determined. This factor includes all changes

-)

in core power distribution caused by control rod insertion, power level j

O changes and axial zenon transients. The product of the steady-state, adjusted peaking factor and the derived factor W(Z) is denominated-W Fg (Z), which is the factor compared to the limits to check compliance wightransientslimitations. SR 3.1.6.2 addresses the verification that Fq (Z) is within limits.

'g 5.

A Required Action has been added to allow the plant to go to MODE 2 if the requirements for power reduction and resetting Overtemperature and Overpower setpoints are not met in the required Completion Time. The LCO is not applicable in MODE 2.

j 6.

This particular specification is remunerated in the MERITS format and is l

SR 3.0.4.

The nonapplicability of this SR appears in a general j

surveillance note.

4 7.

The requirement for THERMAL POWER at which flux maps are taken for an

.)

F (Z) evaluation has been simplified to be MODE I.

l g

The requirement to increase the._ measured by [

] for manufacturing tolerances and by [

] for measurement uncertainties has been simplified to require multiplying the measured value by [

).

The above requirements appear in a NOTE in the MERITS specification.

j l

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l WOG-MERITS Page 2 of 3 Rev. B j

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Heat Flux Hot Channel Factor - Fg(Z) 3.1.6B 7

(

8.

These limits will be provided in the COLR.

9.

These requirements appear in a general surveillance note.

10. The functions K(Z) and W(Z) will be provided in the COLR.

11. A requirement has been added to verify that Fg is within limits prior to exceeding 75% of RATED THERMAL POWER after each refueling. This requirement has been included in the frequency column of SR 3.1.6.1

12. Specifics for calculating the percent by which Fg exceeds its limit will be provided in the COLR.

13. Requiring completion within 15 minutes is not consistent with the reality that it takes anywhere from 6 to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , or longer to process the results of a flux map. The MERITS specification requires adjustment of the AFD limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and after that, the AFD specification spells out requirements for controlling axial power distribution, they need not be repeated in this specification.

14. This is redundant to the LC0 requirements and previously stated Required Action:.

15. This item refers to Base Load operation and is deleted.

16. Identification of the applicable core planes appears in a surveillance note for SR 3.1.6.2 in the MERITS specification.

17. This requirement is contained in a general surveillance note.

%/

WOG-MERITS Page 3 of 3 Rev. B

Nuclear Enthalpy Rise Hot Channel Factor - FAH 3.1.7 O

JUSTIFICATION FOR CHANGES TO W-STS Rev. 5 The following justifications are provided to explain the rsasons for the differences between the Desk Reference, W-STS Rev. 5, the base Standard Technical Specification, and this MERITS specification. The number for each written justification corresponds to a number appearing on the attached, marked-up Specification 3.2.3, RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor.

This W-STS specification has been modified to incorporate the latest licensed methodology in which the limits of F N AH are specified in the LCO replacing the total flow rate and R criteria. This methodology eliminates the use of a rod bow penalty by reduction in the available DNBR margin. Typical values of DNBR margins of 9.1% completely offset typical rod bow penalties of 1% to 3%. Upon removal of the rod bow penalty, R becomes a constant (1) and Figure 3.2-3 degenerates into a simple rectangle where R and RCS flow are independent. Thus, Figure 3.2-3 may be deleted and the RCS flow requirement moved into the Specification for DNB parameters, MERITS specification RCS Pressure, Temperature, and Flow DNB Limits, LCO 3.3.1.

This change will simplify the requirements and make it much easier to determine the proper and conservative corrective measures specified by the Required Actions, in the event of an out of limit flow condition. The technical basis for this methodology is described in detail in the MERITS Bases and in Westinghouse WCAP 8691, Rev.1, Fuel Rod Bow Evaluation, dated July 1979. This approach has been licensed at Vogtle, Farley, South Texas, Byron, and Millstone, and submitted for Comanche Peak.

Further, this simplified specification is applicable to all plants on a generic basis, since the rod bow penalty can be eliminated on all plants. The ability to eliminate the rod bow penalty is based on reactor physics principles rather than plant specific equipment or a

configurations.

J O

WOG-MERITS Page 1 of 4 Rev. B 1

N Nuclear Enthalpy Rise Hot Channel Factor - FAH 3.1.7 1.

The W-STS LCO which specifies the RCS total flow and R requirement, N

including Figure 3.2.3, has been replaced by the F g g

requirement.

N The limits for FAH will be given in the CORE OPERATING LIMITS REPORT.

N 2.

The condition has been revised to specify "FAH outside limit" (new Condition A) consistent with the LCO change in methodology.

N 3.

Action a.1 has been revised to reflect the change to FAH and is specified in Required Action A.I I.

The Completion Time has been changed from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time provides time to restore N

FAH to within its limits. This restoration could, for example, involve re-aligning any misaligned rods or reducing power enough to bring N

Fgg within its power dependent limit. When the FAH limits are exceeded, it is unlikely that the DNBR limit would be violated

[

in steady : late operation, since events which could significantly perturb the F N AH value (e.g., static Control Rod Misalignment) are considered in the safety analyses; however, the DNBR limit may be violated if a DNB limiting event were to occur. Thus, the Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months represents a short time period during whir.h the likelihood of a DNB limiting event occurring is insignificant.

4.

Action a.2 has been specified in Required Action A.I.2.1.

The Completion Time has been changed from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time

]

provides time to reduce THERMAL POWER to < 50% of RATED THERMAL POWER (RTP) by boration or movement of the Control Rods. Due to the functional N

relationship between power level and FAH, the limit will increase with power level reduction. This is acceptable since the DNB margin

{

l increases with a reduction in power level. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l

represents a short time period during which the likelihood of a DNB limiting event occurring is insignificant.

B O

WOG-MERITS Page 2 of 4 Rev. B

1 Nuclear Enthalpy Rise Hot Channel Factor - FAH 3.1.7 -

O 5.

The first part of Action b. has been revised to specify the verification of N

FAN and'it appears in Required Action A.2.

6.

The second part of Action b. has been specified in new Condition.B.

Condition B requires the plant be placed in MODE 2 if the applicable Required Action (s) of Condition A could not be completed within.their-Completion Time. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is typical of the allowance in the STS for placing the plant in MODE 3 from normal power operation.

7.

The Action c. requirement to " identify and correct the cause of the out-of-limit condition" has been deleted, since identification of the cause is not always possible and the other specified Actions are adequate to 1

assure safe operation of the plant.

i 8.

The requirements for Actions c.1,c.2, and c.3, have been revised to require verification that F N AH is within its limits prior to increasing THERMAL POWER above a nominal 50% of RTP, a nominal 75%, and within 24:

O-3 hours of attaining 1 95% of RTP are required via a surveillance frequency for each out-of-limit occurrence. These requirements have been specified in new Required Action A.3.

9.

Surveillance Requirement 4.2.3.1 provides an exception to the STS specification 4.0.4.

This exception has been changed to a general surveillance note which references the MERITS equivalent requirement in SR i

3.0.4.

This exception has been retained since, the plant must be in MODE 1 j

to perform the subject surveillance which demonstrates that FAH is within its limits.

The surveillance requirement to increase the measured FAH by 4%

for measurement uncertainty has been added as a general surveillance note.

j Previously this uncertainty was taken into account in the Figure 3.2-3 j

criteria.

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WOG-MERITS Page 3 of 4 Rev. B l

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Nuclear Enthalpy Rise Hot Channel Factor - FAH 3.1.7

(

10. Surveillance Requirement 4.2.3.2 has been revised to specify verification N

of F y to within limits.

g

11. The Surveillance related to measurement of flow have been removed from this specification. Disposition of these surveillance is discussed in the B

MERITS specification for DNB parameters, LCO 3.3.1.

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l WOG-MERITS Page 4 of 4 Rev. B

EEV. b POWER DISTRIBUTION LIMITS 3/4.2.3 RCS Flow RATE AND NUCLEAR ENTHALPY RISE HOT CHANN,EL FACTOR LIMITING CONDITION FOR OPERATION

-9

.3 The coreinstion of indicated Reactor Coolant System (RCS) total flow rat nd R shall be maintained within the region of allowable operation wn on Figu 3.2-3 for four loop operation.

Where:

N.

Fas 1.49 LL.

0.2 (1.0 - F))

i THERMAL POWE

, and b*

P =

RATED THERMAL POWER Fh=Measuredval fFhobtaine y using the novable incere c.

detector obtain a power distrib on map. The measured ofFhshallbeusedtocalculate since Figure 3.2-3 val i

neludes penalties for undetected feedwater v ri fouling of I

[0.1K and for measurement uncertainties of (2.1%

flow and 4% for incere measurement of Fh.

APPLICABILITY: MODE 1.

f ACTION:

g F~g a co7 5stE L E tn sT~

with t% r efn:tfen Of "C Mut fka rou ;rd R ;;uf t: th; regi;n of h

uti: ;;rHf:n :h;;r On Ff;= 2.2-2;-

4 h

a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1.

Resto I"tAe-comb'netten of M ! t ul f! : r: u =d : to within the above limits, or 2.

Reduce THERMAL POWER to less than 505 of RATED THERMAL POWER I

and reduce the Power Range Neutren Flux - High Trip 5etpoint to less than or equal to 55% of RATED THERMAL POWER within the g I

next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

H Fsn sk~tt \\;e w$n 1he uah s opensriac uinirs nekar (coin) pxxiect, tje gawQ CDMANCHE PEAK - UNIT 1 3/4 2-8 i

._____-.__a

REs/. S Y

me e

e-e e

e 9

e e

O t

QF Gsta osssreo)

V G w, -,b Wes y i

n.........-,.... -,,..........

?

(

C08%N;HE PEAK

UNIT 1 3/4 2 9 4

A G

9

Decket No. 30 445 August 34. 8987 POWER D15TRIBUT10N LTMITS LTNITIE C0CITION FOR DPERATION ACTION (Continued) b.

Within to hours of initially being'evtside the above limits, vertfy4

!=:= : ::;ir :: '.

221 % = 2

-;T4:= :::

W:

i="S: :" S cf J' tut "'r --2 r. n:t--f t within S;^

_the abeve iteits.elreduca THEDEL POWEA la less than EE of arED Mlpnt POWER,within the eart+ hours.

A

>.. m.a _ m... c-- _

a.

to increa POWER above the reduced THEIDE sit seguired Wy ON a..

above-POWER OPERATION

/7%

may proceed provided that the -

..._- R and indicated RCS VJ tatal flow rete are 6, thropph inhcd and 805 total esapartsen am he within eAn emeten e' la.

en shown en Fiaure 3.1-$f prior to eacteding the fellering B

MDut POWER levels:

g,,,, y g],, y,,,,,,,,,,,,,,,,,,,,,,,,

1.

A arminst SOK ef RATD TEDEL POWER,

~

2.

A meninal 7EE of RATED THEDEL POWER and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than er equal ta 95% of RATD THEWnt POWER.

SURVEILLaw t REQUlttMINTs 3

W C 7'E 5

frht The previsions of betticatten 4.0.+sre met (cable.

~

ru pues.aso us sTr s sm se smaanue a v er.

.mem asews wear **v.

4. I. *w. I IM A T 174".$K. Of fi.IN^UI D UNI NW ^6U Zd $ NOII DO NUI d=fe S MLW S n;h. Of :. n;tek ;;ntS Of ff;.. 3."-3:

Prior ta operation above 7EE of RATE THESEL POWER a ve u r r n w a m m as.

a.

levitar,and 6.

At laast ance per 31 Effective Fv11 Power Days.

4.1.3.3 The indicated tes total fler rate shall be verified to be within the Q/t regten of acceptable operatten of Fleure 3.2-3 at least once per 12 heves when the most recently attained salue of R ettained per Specificatten 4.1.1.1.18 l

sisumed la amist.

4.1.3.4 The RCI total flev rata indicatars shall be stjected ta a OMANNEL.

l CALIBRATION at least once per is senths. The esasurement instrumentatten l

shall be cattbrated within 7 days prier la the performance of the salerimetric flev esasurement.

4.1.3.5 The R $ tata1 flow reta shall be estatstned by pr,ctaten heat balance

~

esasurement at least once per 18 months. ~

COPEN:NE PEAK - UNIT 1 1/4 2 10 h.

AFD 3.1.8B JUSTIFICATION FOR CHANGES TO SPECIFICATION F0F NORTH ANNA t'

The following-justifications are provided to explain the reasons for the'-

differences between the specification For North Anna, which is the base for this particular specification, and this MERITS specification. The number for l

each written justification corresponds to a number. appearing on the attached, marked up Specification 3.2.1, Axial Flux Difference (AFD).

l There are currently in use two standard methodologies for the control of axial power distribution. For completeness both methodologies have been. included in the MERITS Specifications. One methodology is the Constant Axial Offset-l Control (CAOC) proced"re. This procedure requires that the axial offset be kept within a band about i target value during normal plant operation, including power change maneuvers, in order to easure that power shapes -are acceptable.

Using the Standard Technical Specification, Rev. 4a (Comanche Peak'- Unit 1) as the base, this particular procedure has been addressed in MERITS specification 3.1.8A. For some plants, the Relaxed Offset Control (RAOC) procedure is employed. This particular procedure.has been addressed in MERITS specification 3.1.88. To show the RAOC procedure, as currently applied on the operating plants, the Technical Specifications for a: typical-plant had to be used, and North Anna was selected.

Documentation on the Westinghouse methodology for RAOC was submitted to the Nuclear Regulatory Commission-(WCAP - '10216-P-A, under cover letter NS-EPR 2649)inAugust1982. In February 1983 Westinghouse received from the NRC-office of Standardization & Special Projects Branch,' Division of Licensing a o

confirmation that the report was approved and that the_RAOC procedure was accepted for referencing in license applications.

The RAOL procedure is being used on a total of thirteen plants, including Catawba and McGuire.

g 1.

This addition establishes that the term " flux difference units" is used for specifying the target flux difference and the target band. The units are J

"%-delta-flux". Axial flux difference is the normalized excore current difference between the top and bottom excore section. Converting to per cent yields whole number units.

2.

The values of AFD limits are removed from the LCO, including Figure 3.2-1, and will be provfded 4 the CORE OPERATING LIMITS REPORT (COLR) which is produced for each cyt.lc. The AFD limits may change from cycle to cycle and therefore are pore repropriately placed in the COLR which will be described in the Administrate Controls of the Technical Specifications.

3.

Action b is not required. SR 3.0.4 will cover this.

In addition, if AFD is not within its limits at 150% of RATED THERMAL POWER, the AFD alarm will indicate that AFD is outside its limits.

O WOG-MERITS Page 1 of 2 Rev. B I

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~

AFD 3.1.8B 4.

The phrase "above 50% of RATED THERMAL POWER" is not required. The Surve111ances support the LCO Applicability.

5.

'If the AFD Alarm is inoperable, monitoring at a frequency of I hour is consistent with the wider AFD operation space which should be infrequently violated.

Logging of values will be accomplished as necessary but need not

'i be specified as a technical specification requirement.

If monitoring of AFD demonstrates that AFD is outside its limits, Required Actions apply immediately.

B i

l 6.

Surveillance 4.2.1.2 has been changed into a note pertaining to the LCO.

It defines the requirements for meeting the LCD.

^

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WOG-MERITS Page 2 of 2 Rev. B v

QPTR 3.1.9 O

JUSTIFICATION FOR CHANGES TO W-STS REV.5 The following justifications are provided to explain the reasons for the differences between the Desk Reference, W-STS Rev. 5 the base Standard Technical Specification, and this MERITS specification. The number for each written justification corresponds to a number appearing on the attached, marked 8

up Specification 3.2.4, Quadrant Power Tilt Ratio.

This revised form of the technical specification on Quadrant Power Tilt Ratio (QPTR) significantly simplifies the requirements and better addresses the primary concern for quadrant power tilts, to assure that core peaking factors remain within their limits while the core is in the tilted condition.

g 1.

The footnote referencing Special Test Exceptions 3.10.2 appears in the Cross References of the MERITS Technical Specification.

2.

The limit of 1.09 is not necessary since the 1.02 limit will be reached prior to 1.09 and the action statements already in progress will address the problem. With action statements to reduce THERMAL POWER and setpoints, surveillance on F N and F have been added to verify that AH g

core peaking factors remain within limits when the QPTR is greater than 1.02. The Action 4 requirement to " identify and correct the cause of the out-of-limit condition

has been deleted, since identification of the-O cause is not always possible and the other specified Actions are adequate to assure safe operation of the plant.

B i

3.

Since the QPTR alarm would already be in its alarm state, additional changes in the QPTR would be detected by requiring a check of the QPTR once i

per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

If the QPTR has increased, THERMAL POWER and setpoints would have to be decreased accordingly.

4.

A requirement has been added to evaluate the accident analyses if the QPTR remains outside its limit. Although power reduction, setpoint reduction, N

and measurement of FAH and Fg will address the impact on tilt, other accident parameters are a function of the power distribution.

This evaluation should confirm the acceptability for continued operation at RATED THERMAL POWER with the actual incore tilt outside its limit.

Following a successful reevaluation, the excore detectors should then be adjusted to zero out the tilt. The point of this entire specification is to have a means of knowing if the gross radici power distribution of the core has changed. The QPTR provides an estimate of what could be happening I

to the core power distribution, although an incore map is required in order to determine if there ex3sts a real tilt in the core power distribution.

Zeroing out the tilt allows the QPTR monitor to detect changes from the tilted, but acceptable condition.

1 WOG-MERITS Page 1 of 2 Rev. B

l 5.

This statement is redundant with the APPLICABILITY statement.

6.

The requirement to perform the surveillance with one NIS excore channel inoperable with THERMAL POWER > 75% of RATED THERMAL POWER' appears as' a NOTE in the frequency column for.SR 3.1.9.2.

l I

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l WOG MERITS Page 2 of 2 Rev. B

ktv. O.

tocket No. 50' 445 POWER DISTRIBUTION LIMITS 089851 I4' IIII 6

3/4.2.4 0UADRANT POWER TILT RATIO l

LIMITING CONDITION FOR OPERATION j

3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: ICDE 1, above 50% of RATED THERMAL POWE ACYiON:

WiththeQUADRANTPOWERTILTRATIOdetarsinedtoexceed1.02'deth l

--er-1... u.. = x.=: e 1.ee:

l 1.

O g ' ;.{ tt.e 0"P2P.'M Z P, TILT "' TIC t ?:::t :- : ?:7 5:2-f n.,g = q = = = u =:_=d= e 3

e

-L;

P"."L ".'

P,i; nd=

d O 1::: thr 5"*" Of **?EU " E*??L tond onw(w h hun f4we<f &.

Ar. -4r-Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months I

?:f :: th: 0"*raar M ", TILT "."TIO O.iu.:.. ne-

?!:;f t, r 3

t; Reduce THERMAL POWER at least 35 from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron 1

ypy Flux-High Trip 5etpoints within the next 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sor c.sd enns.pt.s- /2 kaars f[LetL/ko=.

h 3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-Migh Trip 5etpoints to less than er equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.

Identify and correct the cause of the out-of-liait condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least i

once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

i or steater RATED THERMAL POWER.

5ee Special Test Exceptions Specification 3.10.2.

COMANCHE PEAK - UNIT 1 3/4 2-11 e

R^v.B b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 7 days thereafter perform surveillance Requirement 4.2.2.2 and 4.2.3.2.

Reevaluate each safety analysis in Table.... and confirm analysis c.

results remain valid for duration of operation under this condition at RATED THERMAL POWER.

After this confirmation, calibrate excore detectors to show zero QPTR prior to increasing thermal power to RTP.

/

O 4

h4d b =

Docket No. 50 445 i

POWER DISTRIBUTION LIMITS August 14. 1987 d

LIMITING CONDITION FOR OPERATION l

ACTION (Continued) i b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to i

misalignment of either a shutdown or control rod:

l i

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER 4s reduced to less than 50% of RATED THERMAL l

POWER.

2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for.

each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 3D sinutes; 3.

Verify that the QUADRANT POWER TILT RATIO is within its lisit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next O

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and, reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equel to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.

Identify and correct the cause of the out-of-limit condition l

prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or. greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

O i

COMANCHE PEAK - LHIT 1 3/4 2-12'

R u /3 j

^

Docket No. 50-445 POWER DISTRIBUTION LIMITS EUIt 8'

O O

LIMITING CONDITION FOR OPERATION I

ACTION (Continued)

Reduce THERMAL POWER to less than 50% of RATED THERMAL thin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron

-High Tr eints to less than or equal to 55%

TED THERMAL POWER wi the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 3.

Identify and corre ca the out-of-lisit condition prior to increasing TH R; subsequent POWER OPERATION above 50% of RATE tMAL y proceed provided that the.

QUADRANT LT RATIO is verif thiri its limit at least once p!

ur for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until ver at 95% or greater HERMAL POWER.

gi d

he provisions of Specification 3.6.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be detemined to be within the li jt d= 5.5 :f ".".T 0 T" F."A ^%':r, b. g Calculating the ratio at least once per 7 days when the alarm is a.

SPERABLE, and b.

Calculating the ratio operation when the ala.at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state m is inoperable.

@ 4. 2. 4. 2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the novable incere detectors to confim that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or full-core flux map, is consistent with the indicated QUADRANT POWER 1

TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l i

O CDMANCHE PEAK - UNIT 1 3/4 2-13 L

RTS Instrumentati;n LCO 3.2.1 r

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JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification,. and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specifications 2.2.1, Reactor Trip System. Instrumentation Setpoints, and 3.3.1, Reactor Trip System Instrumentation.

1.

The Limiting Safety System Settings have been combined with the LCO to form a single LCO with all channel, train, and interlock requirements contained within LCO 3.2.1, Reactor Trip System (RTS) Instrumentation.

Specifications 2.2 and 3/4.3.1 are now combined into a single LCO since all the information from these two Specifications deals with the same instrumentation.

2.

A new Condition A addresses one or more channels for one or more functions -

from Table 3.2.1-1 inoperable. The Required Actions are to refer to the table and then perform the Required Actions as referenced in the appropriate Conditions from the table. This Condition directs the user to the table to determine which Condition (s) to enter and then to follow the prescribed Required Actions. This Condition also allows more than one channel / train / interlock to be inoperable and still have a prescribed course n

of action to follow. This is consistent with the STS but is now clearly stated for the reader.

Action a. has been deleted from the specification.

If the measured Trip Setpoint exceeds the Allowable Value, the channel is declared inoperable and the unit is in an Action Statement (Condition). This is a conservative modification to the Westinghouse STS and is implemented because it reflects current industry practice.

3.

With the setpoint less conservative than the Allowable Value, the associated function must be declared inoperable and the referenced Required Actions for the appropriate Condition of Table 3.2.1-1 must be performed.

4.

The performance of Equation 2.2-1 for evaluating Total Allowance has been deleted from the specification. 'This is based on current industry practices.

5.

The reorganization of the tables now has All of the requirements for gash function contained within the single Table 3.2.1-1.

6.

The TOTAL ALLOWANCE (TA), Z value, and SENSOR ERROR (S) are no longer required because equation 2.2-1 is no longer calculated. See also Item 4.

7.

The f(AI) function has been added to the table as a line item to more correctly identify the Surveillance Requirements (SR) associated with this function. Previously, these SRs were identified with the Power Range p

Neutron Flux, High Setpoint reactor trip and were not really associated with that function.

WOG-MERITS Page 1 of 7 Rev. B

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RTS Instrumentation LCO 3.2.1 p

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specifications 2.2.1, Reactor Trip System Instrumentation Setpoints, and 3.3.1, Reactor Trip System Instrumentation.

8.

The plant specific loop design flow values are now located in the Bases section. The table now uses % flow as this is what the control room operator monitors. This was done to be consistent with the Writers Guide-and to use units that are consistent with the meter indications on the control board.

9.

This function has been added to the table to include as many Westinghouse reactor trip functions as reasonably possible. Numerous Westinghouse plants have this reactor trip function.

10. The reactor trip breaker undervoltage and shunt trip mechanisms have been-called out as a separate line item in the table to more easily facilitate the treatment of the inoperability of these functions separately from the inoperability of the reactor trip breakers (RTBs). This is consistent with GL 87-09.

11. The equations were modified to be more technically correct ustog the 2 and s symbols instead of - symbols.

12. The wording has been modified to also accomodate plants that have removed the bypass redistance temperature detector manifolds and now use the loop thermowells.

13. The response time testing of the RTS functions is now described in the Surveillance Requirements section. Response Time Table 3.3-2 has been deleted but the requirement to perform response time testing has been retained in the Surveillance Requirements section.

Improvement programs at the Seabrook and Vogtle nuclear plants have already accomplished this change to Technical Specifications and have relocated the resp 0nse time table to other controlled documents.

There are no plant-specific differences which would make this change not applicable at any other Westinghouse plant.

It is merely the relocation of the Table containing the actual response times to another licensee controlled document. This is an administrative change and does not involve plant-specific differences. Therefore, this change is generic to all Westinghouse plants.

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I O WOG-MERITS Page 2 of 7 Rev. B j

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4 RTS Instrumer.tation LCO 3.2.1 l

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JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain.the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard i

Technical Specification, and this MERITS Specification.- The number for each

-l written justification corresponds to a number appearing on the attached, marked-up, Specifications 2.2.1, Reactor Trip System Instrumentation Setpoints, and 3.3.1, Reactor Trip System Instrumentation.

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14. The TOTAL NO.- 0F CHANNELS column of' Table 3.3-1 has been deleted. Only-the l

MINIMUM CHANNELS OPERABLE column is included and it is now the total number of channels for that RTS function. This is based on a change to the tech spec' format. The CHANNELS TO TRIP column of Table 3.3 has been deleted.

i The trip actuation logic for a particular RTS function is now found in the Condition Statements of the specification. The MINIMUM CHANNELS OPERABLE-of Table 3.2.1-1 is now the same as the TOTAL NO. OF CHANNELS from Table i

3.3-1.

With less than the total channels available, a Condition, with associated Required Actions, is entered.

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15. The Action Statement has been expanded to more clearly state the intent of the action. Separate Conditions now address inoperable intermediate range l

channel (s) when < the P-6 setpoint, and when > P-6 but < P-10.

If power is

> P-6 but <: P-10, the operator can now reduce power to < the P-6 setpoint-

'y p_t increase power to 2 the P-10 setpoint as the intermediate range channels are not required to be operable in either of these conditions. Note that i

the safety analyses take no credit for this reactor trip function. Note i

also that the Nuclear Instrumentation System (N!S) intermediate range channels must be OPERABLE prior to reducing power to < P-10 if the operator chose to increase power to > P-10.

The existing Action Statement would f

have you fix the inoperable intermediate range channel' prior to exceeding the P-10 setpoint. This could leave you without the required i

instrumentation for the existing unit condition until the channel was fixed.

A new Condition was added to address the case of two inoperable intermediate range channels if < P-10 and > P-6.

The actions are to suspend positive reactivity additions and to reduce power to less than the P-6 setpoint. Suspending positive reactivity additions precludes power-escalation and reducing power to < the P-6 setpoint places the NIS source range channels in operation. The actions associated with LCO-3.0.3 are not appropriate for the condition of two inoperable chanr.els > P-6 and < P-10.

The existing action for an inoperable NIS intermediate range channel when <

l P-6 has been expanded to cover one or two inoperable NIS intermediate range l

channels when < P 6.

The intent here was to provide a course of action in i

the event that both intermediate range channels were inoperable. Note that the safety analyses take no credit for the intermediate range channels.

The actions associated with LCO 3.0.3 are not appropriate for the condition of two inoperable channels < P-6.

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WOG-MERITS Page 3 of 7 Rev. B

l RTS Instrumentation 1

LCO 3.2.1 l

JUSTIFICATION FOR CHANGES TO W-STS REV. S' The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specifications 2.2.1, Reactor Trip System Instrumentation Setpoints, and 3.3.1, Reactor Trip System Instrumentation.

16. A new condition has been added to address two inoperable NIS source rangel channels when < P-6.

The actions are to suspend positive reactivity-additions and to open the RTBs. Suspending positive reactivity-additions.

precludes power escalation. Opening the RTBs transfers actions to Condition K.

Here, positive reactivity additions are suspended, all valves that could permit unborated water to enter the Reactor Coolant System are closed, and the SHUTDOWN MARGIN must be verified to. ensure reactor stability is maintained.- The actions associated with LCO 3.0.3 are not appropriate for the condition of two inoperable channels < P-6.

The MODE 3, 4, and 5 requirements have been separated into:.a) with the RTBs open, or b) with the RTBs closed and the Rod Control System capable of rod withdrawal. These actions now address one inoperable NIS source range channel and two inoperable source range channels with the RTBs closed.

With two inoperable source range channels, all operations involving O

positive reactivity additions must be suspended and the RTBs opened and' Condition K entered. With the RTBs open,.only one source range channel is required to be OPERABLE. This is licensed at the Diablo Canyon Plant and the North Anna Plant. The actions associated with LCO 3.0.3 are not appropriate for the condition of-two inoperable channels.

The STS was not clear as to whether the RTBs were open or closed.

Engineering judgement would indicate that with the RTBs closed, the possibility for a positive reactivity addition accident.is much greater than with the RTBs open. With the RTBs open, only a boron dilution accident or uncontrolled RCS cooldown could add positive _ reactivity.

These are both relatively slow transients and.would allow operator detection and associated actions to mitigate these transients. Therefore, only one NIS source range channel is' required to perform the monitoring functions. The new MERITS specification has appropriate actions which the STS does not have to address the loss of one or two NIS' source range channels in MODES 3, 4, or 5.

The design of the RTS's at Westinghouse plants is essentially the same with regards to the protection afforded by the NIS source range in MODES 3, 4, and 5.

Therfore, this change is generic to all Westinghouse plants.

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17. The MERITS Technical Specifications do not address (n-1) loop operation as so few plants have this option licensed.

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RTS Instrumentation LCO 3.2.1 1

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JUSTIFICATION FOR CHANGES TO W-STS REY. 5 The following justifications are provided to explain the reas'ons for the differences between the Desk Reference W-STS Rev. 5, the base Standard-Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specifications 2.2.1, Reactor Trip System Instrumentation Setpoints, and 3.3.1, Reactor Trip System Instrumentation.

18. This reactor trip function is automatically blocked below P-7 and automatically enabled above P-7.

This has been added to the Applicability statement for clarification.

19. This trip function is automatically blocked below P-7 (Two Loops),

automatically enabled above P-7 and below P-8 (Two Loops), automatically blocked below P-8 (Single-loop), and automatically enabled above P (Single Loop). The Applicability section now clearly addresses these design features.

20. A line item has-been added to the table to address plants that have only three channels per steam generator because they also' use the Function-16.

reactor trip on Steam Generator Water Level--Low Coincident With Steam /feedwater Flow Mismatch.

21. The turbine trip functions are automatically enabled above [P-9],

automatically blocked below [P-9], and this is now clearly stated in the Applicability statement. The exact permissive varies from plant to plant and will be the correct one for that plant when the specification is made plant specific. This is licensed at the Diablo Canyon Power Plant.

All Westinghouse turbine trip functions are related to, and interlocked with, one of the Permissives, P-7, P-8, P-9, or P-10. The change to the specification now clearly states this in the Mode Applicability Column of the Table. This is not a technical change, only an editorial change, and is a generic item applicable to all Westinghouse plants.

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22. The Catawba Unit was recently fined for not having four channels operable because this function requires 3/4 channels to be < P-10 in order for the function to be in the required state. Going below P-10 automatically enables the Power Range, Neutron Flux--Low Setpoint reactor trip and the Intermediate Range, Neutron Flux reactor trip. Therefore, the Minimum.

Channels Operable has been increased to four. The Action has also been modified to have the operator verify the status of the interlock within 15 minutes after going below P-10.

The current action of verifying the status of the interlock within one hour does not really meet the intent when the concern is when going below P-10.

WOG-MERITS Page 5 of 7 Rev. B j

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RTS Instrumentation LCO 3.2.1-p JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain-the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard.

1 Technical Specification, and this MERITS Specification. The number for each

.j written justification corresponds to a number appearing on the attached, 1

marked-up, Specifications 2.2.1, Reactor Trip System Instrumentation Setpoints, l

and 3.3.1, Reactor Trip System Instrumentation.

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23. The Action Statement has been expanded to open the RTBs within the following hour. This is a conservative modification and places the unit in l

a safe condition with the manual trip switch (es) inoperable.

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24. The change in allowed time is justified in WCAP-10271, Supplement 1-P-A, Evaluation of Surveillance Frequ"encies and Out of Service Times for the Reactor Protection System, Supplement 1, May 1986.

25. The Action Statement has been expanded to-allow placing the inoperable channel in bypass while resetting the NIS power range Trip Setpoints.

If the channel is not placed in bypass while performing this action, a reactor trip will result when the OPERABLE channels' Trip Setpoints are reduced.

26. How.the determination of whether the interlock is in the required state for I

O the existing plant condition should not be limited to observation of the annunciator window. This phrase was removed to allow the unit more flexibility as to how the determination is made.

27. The MODES FOR WHICH SURVEILLANCE IS REQUIRED column of Table 4.3-1 has been j

deleted as this is the same MODE applicability as already exists in Table 3.2.1-1.

28. The performance of an ANALOG CHANNEL OPERATIONAL TEST prior to reactor startup has been added to ensure that this function is OPERABLE to provide q

protection during reactor startup. The safety analyses take no credit for the NIS source and intermediate range reactor trips. This function provides core protection during startup and must be verified OPERABLE prior i

to startup.

WOG-MERITS' Page 6 of 7 Rev. B

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RTS In:.trumentation LCO 3.2.1 i

V JUSTIFICATION FOR CHANGES-TO W-STS REV. 5 The folicwing justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specifications 2.2.1, Reactor Trip System Instrumentation Setpoints, and 3.3.1, Reactor Trip System Instrumentation.

29. The frequency has been changed to every 31 effective full power days

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(EFPD). This is the intent, to calibrate the excore NIS power range channels as they vary with core burnup and flux distribution. Basing the surveillance on core burnup is more reflective of the changes occurring.

This is also licensed at the Diablo Canyon Power Plant.

30. The frequency has been changed to every 92 EFPD. This is the intent, to calibrate the excore Nls power range channels as they vary with core burnup and flux distribution. Basing the surveillance on core burnup is more reflective of the changes occurring. This is also licensed at the Diablo Canyon Power Plant.

The intent of the previous two surveillance is to periodically verify the accuracy of, and calibrate, the NIS power range channels as the core is depleted as determined by the more accurate incore system. The current STS requirements are performed every calendar month or every three calendar months. The new requirement to perform the surveillance every 31 or 92 EFPD is more conservative if the unit is at high power. The intent of these surveillance is to adjust the NIS power range channels as the flux patterns vary with core burnup.

It is therefore more logical to base the surveillance frequency on core burnup rather than calendar months.

In the case where the unit is at low power levels, the new surveillance frequency still meets the intent, and simultaneously eliminates unnecessary testing at power.

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31. Prior to placing the bypass breakers in service, they must meet the same Surveillance Requirements as the reactor trip breakers. This is also noted 1

in the appropriate SR.

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32. The Westinghouse neutron detectors are supplied with a fixed voltage power source, with a minor adjustment capability.

It is therefore not possible to obtain a true plateau curve for these detectors. The OPERABILITY of the 1

detectors is checked daily during performance of SR 3.2.1.2, CHANNEL j

CALIBRATION.

Inability to adjust the output from the detects. s to meet the requirements of the SR would be the primary indication ths.t'the detector was failing.

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1 WOG-MERITS Page 7 of 7 Rev. B l

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ESFAS Instrumentation LCO 3.2.2 JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specification 3.2.2, Engineered Safety Features Actuation System Instrumentation.

1.

The reorganization of Tables 3.3.3, 3.3.4, 3.3.5, and 4.3.2 now has all of the requirements for each function contained within the single Table 3.2.2-1.

l 2.

A new Condition A addresses one or more channels for-one or more functions from Table 3.2.2-1 inoperable. The Required Actions are to refer to the table and then perform the Required Actions as referenced in the appropriate Conditions from the table.

This Condition now directs the user to the table to determine which Condition (s) to enter and then follow the prescribed Required Actions. This Condition also allows more than one channel / train / interlock to be inoperable and still have a prescribed course of action to follow. This is consistent with the STS but is now clearly stated for the reader.

Q Action a. has been deleted from the specification.

If the measured Trip Q

Setpoint exceeds the Allowable Value, the channel is declared inoperable and the plant is in an Action Statement (Condition). This is a conservative modification to the Westinghouse STS and is implemented because it reflects current industry practice.

3.

With the setpoint less conservative than the Allowable Value, the associated function must be declared inoperable and the referenced Required J

Actions for the appropriate Condition of Table 3.2.1-1 must be performed.

4.

The performance of Equation 2.2-I for evaluating Tott.1 Allowance has been deleted from the specification. This is based on current industry i

practices.

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l WOG-MERITS Page 1 of 3 Rev. B L

ESFAS Instrumentation' LCO 3.2.2 t

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specification 3.2.2, Engineered Safety Features Actuation System Instrumentation.

5.

The response time testing of the Engineered Safety Features Actuation System (ESFAS) functions is now described in the Surveillance Requirements section. Response Time Table 3.3-5 has been deleted but the requirement' to perform response time testing has been retained in the Surveillance Requirements section. Improvement programs at the Seabrook and Vogtle nuclear plants have already accomplished.this change to Technical Specifications and have relocated the response time table to other controlled documents.

There are no plant-specific differences which would make this change not applicable at any other Westinghouse plant.

It is merely the relocation of the Table containing the actual response times to another licensee controlled document. This is an administrative change and does not involve plant-specific differences. Therefore, this change is generic to all Westinghouse plants, a

6.

The. TOTAL NO. OF CHANNELS Column and CHANNELS T0 TRIP Column of Table 3.3-3

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have been deleted. Only the MINIMUM CHANNELS OPERABLE Column is included and it is now the total number of channels for that ESFAS function. This is based on a change to the tech spec format.

7.

The ESFAS specification has been expanded to include two-loop, three-loop, and four-loop plants.

8.

The MINIMUM CHANNELS OPERABLE is now the TOTAL NO. OF CHANNELS from Table 3.3-1.

With less than the total channels available, a Condition, with associated Required Actions, is entered. The column entry has been expanded to be more definitive.

9.

This information has been removed from the table as it was not complete and it is not the purpose of Tech Specs to provide system descriptions.

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10. The MERITS Technical Specifications do not address (n-1) loop operation as j

so few plants have this option licensed.

11. This function has been added to the table to include as many Westinghouse.

ESFAS functions as reasonably possible. Numerous Westinghouse plants have this ESFAS function.

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l WOG-MERITS Page 2 of 3 Rev. B

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ESFAS Instrumentation LC0 3.2.2 O

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specification 3.2.2, Engineered Safety Features Actuation System Instrumentation.

12. The requirement to have the individual main steam line isolation valve switches as part of the ESFAS functions has been deleted. This is a manual actuation and the safety analyses do not model manual actuation for these valves. This function is covered in LCO 3.6.3, Main Steam Isolation Valves.

13. The requirement to have the individual auxiliary feedwater pump switches as part of the ESFAS functions has been deleted. This is a manual actuation and the safety analyses do not model manual actuation of auxiliary feedwater. This function is covered in LCO 3.6.5, Auxiliary Feedwater System.

14. The safeguards 4 kV bus undervoltage and degraded voltage functions only start the emergency diesel generators. These functions have been relocated to LCO 3.7.1, AC Sources - Operating.

15. Control room emergency ventilation functions have been added to cover as j

many Westinghouse ESFAS functions as reasonably possible'.

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16. The change in allowed time is justified in WCAP-10271, Supplement 2, Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System, Supplement 2, February' 1986.

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17. How the determination of whether the interlock is in the required state for the existing plant condition is made should not be limited to observation I

of the annunciator window. This phrase was removed to allow the plant more flexibility as to how the determination is made.

18. There was no completion time associated with this action and four hours was selected based on, and consistent with, LC0 3.5.1, Containment Integrity.

19. This function is only required in MODES 1, 2, and 3.

Action 19 places the unit in MODE 5.

The Required Action has been changed to # 23 (Condition 1) which places the unit in MODE 4.

20. If the P-12 and/or P-14 functions are inoperable, other functional requirements are contained in the respective functional section.

Observation of the alarm window is not the only response that should be taken.

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21. The MODES FOR WHICH SURVEILLANCE REQUIRED column of Table 4.3-2 has been -

deleted as this is the same MODE applicability as already exists in Table 3.2.2-1.

WOG-MERITS Page 3 of 3 Rev. B

Accident Monitoring Instrumentation.

LCO 3.2.3 JUSTIFICATION FOR' CHANGES TO W-STS REY.:5 The following justifications are provided to explainfthe reasons for th'e differences between the Desk Reference W-STS Rev. 5, the base Standard' l-Technical Specification, and this MERITS Specification._ The number for each written justification corresponds to a number appearing on the attached,.

1 marked-up, Specification 3.3.3.6, Accident Monitoring Instrumentation.:

l.

All required information ~is now contained.in a single Table 3.2.3-1.

The information from Tables 3.3-10;and 4.3-7 has been combined.into a single Table.3.2.3-1.

2.

A new Condition' A addresses one-or more channels for one or more, functions from Table 3.2.3-1 inoperable. The Required Actions are to

_l' refer to the table and then perform the Required Actions as referenced in the. appropriate Conditions from the table. This Condition directs the user to the table to determine which Condition (s) to enter and -

'then to-follow the prescribed. Required Actions. This Condition also allows more than one-channel to be inoperable and still have. a prescribed course of action to follow. This is consistent with the-STS but is now clearly stated for the reader.

3.

The new table of functions and requirements-does not include these-I functions.

I 4.

These requirements have been relocated to the Administration sntion.

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of the tech spets.

5.

The surveillance are now found in the Surveillance Requirements section as SR 3.2.3.1 and SR 3.2.3.2.

6.

The TOTAL NO. OF CHANNELS column and the MINIMUM CH' NNELS OPERABLE

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A column of Table 3.3-10 have been deleted. The total number of-channels for that accident monitoring function is:now the REQUIRED NUMBER OF CHANNELS. This is based on a change to the. tech' spec format. With less than the REQUIRED NUMBER OF CHANNELS available, a l

Condition, with associated _ Required Actions,-is entered.

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7.

This table is for illustration purposes only.

It does not attempt to encompass every Type A, Category 1 variable used at every plant. This table contains the types of variables commonly found and includes variables that utilize each of the prescribed-Required Actions.

Plant-specific lists are in the plant-specific FSAR and will be used when the specification is made plant-specific.

O WOG-MERITS Page 1 of 2 Rev. B

Accident Monitoring Instrumentation LCO 3.2.3 p

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number'for each written justification corresponds to a number appearing on the attached, marked-up, Specification 3.3.3.6, Accident Monitoring Instrumentation.

8.

The Required Actions now clearly address the condition where one of the channels for a "per unit" function, steam generator level for example, is inoperable. The Completion Times are on a per steam generator basis. This is different from the normal Completion Time requirements of the MERITS LCOs.

a 9.

Additional Conditions and associated Required Actions have been added to address units that have more than two channels per function. These Required Actions are the same as licensed at the Vogtle Nuclear Power i

Plant, copy attached.

The Vogtle requirements are not plant-specific requirements. Any Westinghouse plant with more than two channels per function would have the same requirements.

Increased outage times are typically allowed if there are more channels to rely upon and this change is an

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addition to the STS. Therefore, this change is generically (Q

applicable to all Westinghouse plants with more than two channels per function.

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(O) v WOG-MERITS Page 2 of 2 Rev. B

Remote Shutdown System LCO 3.2.4 f%

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specification 3.3.3.5, Remote Shutdown System.

1.

All required information is now contained in a single Table 3.2.4-1.

The information from Tables 3.3-9 and 4.3-6 has been combined into a single Table 3.2.4-1.

2.

A new Condition A addresses one or more channels for one or more functions from Table 3.2.4-1 inoperable. The Required Actions are to refer to the Table and then perform the Required Actions as referenced in the appropriate Conditions from the Table. This Condition directs the user to the Table to determine which condition (s) to enter and then to follow the prescribed Required Actions. This Condition also allows more than one channel / transfer switch / control circuit to be inoperable and still have a prescribed course of action to follow.

This is consistent with the STS but is now clearly stated for the reader.

r 3.

The allowed outage time has been increased to 30 days. This is I

currently licensed at the North Anna,-D.C. Cook, and Trojan power

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plants (see attached copies of these specifications).

The instrumentation channels used at the three referenced plants are similar to the instrumentation channels used at all Westinghouse plants. The surveillance requirements at the three referenced plants are similar to those at all Westinghouse plants. Many unit years of operating experience have shown that the 30 day outage time is acceptable. This change is generically applicable to Westinghouse plants and should become a part of the MERITS specifications.

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4.

The preparation of the Special Report is covered in the Administrative Controls Chapter of Tech Specs and is not required to be in the Action Statements.

1 WOG-MERITS Page 1 of 2 Rev. B I

l Remote Shutdown System LCO 3.2.4 JUSTIFICATION FOR CHANGES TO W-STS REY. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up, Specification 3.3.3.5, Remote Shutdown System.

5.

This table is for illustration purposes only.

It does not attempt to encompass every variable used at every plant. This table contains the types of variables commonly found and includes variables that utilize each of the prescribed Required Actions. Plant-specific lists will be used when the specification is made plant-specific.

6.

The TOTAL NO. OF CHANNELS column and the MINIMUM CHANNELS OPERABLE column of Table 3.3-10 have been deleted. The total number of channels for that remote shutdown monitoring function is now the REQUIRED NUMBER OF CHANNELS. This is based on a change to the tech spec format. With less than the REQUIRED NUMBER OF CHANNELS available, a Condition, with associated Required Actions, is entered.

7.

The list of controls, power, and transfer switches has been deleted and replaced with the second portion of the LCO. This approach has been licensed at the Vogtle and Shearon Harris plants (see attached copies of these specifications).

This change is not a technical change.

It is an administrative change that reformats the long list of switches and control circuits.

The actual switches and control circuits are plant-specific. This l

change merely reformats the method by which the individual switches and control circuits are presented. This change is generically applicable to all Westinghouse plants.

g 8.

The Required Actions now clearly address the condition where one of l

the channels for a "per unit" function, steam generator level for example, is inoperable. The Completion Times are on a per steam generator basis. This is different from the normal Completion Time requirements of the MERITS LCOs.

WOG-MERITS Page 2 of 2 Rev. B

RCS Pressure, Temperature, and Flow DNB Limits 3.3.1

!O JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the ' Desk Reference" W-STS Rev. 5, which is the base Standard Technical Specification, and this rewritten MERITS Specification. The number for each justification corresponds to a number appearing on the attached, marked-up Specification 3.2.5 from W-STS Rev. 5.

The changes indicated below in items 1 through 6 to this specification may be applied to all Westinghouse Pressurized Water Reactors. DNB is a function of heat transfer and fluid flow characteristics.

It is not a function of how the heat is generated.

1.

LCO 3.2.5 has been relocated to the Reactor Coolant System chapter in the MERITS version and exists as LCO 3.3.1.

In addition, requirements for RCS Flow rate that existed in STS specification 3.2.3, RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, have been incorporated as required into the MERITS DNB specification. The disposition of requirements that existed in the STS specification 3.2.3 and not incorporated into the MERITS DNB specification are explained in item 5 below.

2.

The LCO is rewritten to incorporate the parameters including, RCS flow rate, such that Table 3.2-1 may be deleted.

\\

l 3.

Instead of within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months are given for placing the plant in MODE 2.

LCO 3.0.3 gives 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to get to MODE 3.

Recognizing that time wise, little difference exists between going to MODE 3 or MODE 2, because most of the time is spent in reducing power to < 10% of RTP. Therefore, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable amount of time to reach MODE 2 without challenging the operators or safety systems.

4.

As discussed in item 1 above, surveillance requirements and related notes for RCS Flow are added as follows:

The RCS total flow rate shall be determined by precision heat balance measurements et least once per 18 months. Within [ ] days prior to i

performing the precision heat balance flow measurement, the instrumentation used for performing the precision heat balance shall be calibrated. The provisions of specification SR 3.0.4 (4.0.4) are not applicable for verification that RCS flow is within its limit.

- Includes a [ ]% flow measurement uncertainty.

B 1

i (

1 WOG-MERITS Page 1 of 2 Rev. B 05/13/1989

RCS Pressure, Temperature, and Flow DNB Limits 3.3.1 5.

The surveillance requirements for verifying RCS flow every 31 days and prior to exceeding 75% RATED THERMAL POWER (RTP) are 'not included in the MERITS DNB specification. These surveillance requirements were initially in place when the specification for F N AH and the specification for RCS flow were combined into one specification. The value of F N AH must be verified prior to exceeding 75% of RTP, and because of the combination, the same surveillance frequency was assigned for RCS flow verification. In MERITS, the specification for N

FAH now stands alone in LCO 3.1.7, Nuclear Enthalpy Rise Hot N

Channel Factor - Fgg, and verification of FAH prior to exceeding 75% RTP is retained there. Therefore, the surveillance on RCS flow every 31 days and prior to exceeding 75% RTP is no longer required.

If there are any significant RCS flow irregularities, they will show up via flow meter indications and RCS temperature indications.

The surveillance requirement for verifying the RCS flow instruments are

(/

calibrated every 18 months is not included in the MERITS DNB specification because the requirement already existed in the STS Instrumentation i

specification and will exist in the MERITS specification LCO 3.2.I, RTS Instrumentation.

6.

This note is moved to the bottom of the specification page since the table is being deleted.

3 6

i i

WOG-MERITS Par,e 2 of 2 Rev. B 05/13/1989

Rev,a h

Docket No. 50 445 14, 1957 3,3 5"" 0 r?":5'r!0" '!"F5 Reuhr C,rlet S sb,(Rcs) August y

,]

3.3.1 -:n. :. : ese - t"rEu RCS Pressore, Tempenha cnd Fl<w DNB Lids

"!"!T!"* 0^" P!0" "5" 0"E***!0" LCO 3.3,1 l

.:.5 The followin{DNE-related pa"ameters shall be maintained within the Sfafd limits ; b r. er,T c.,...,..

Resetee<T:ge, temperature.

RCS cVtra...r.: :y. -- T,,, =: 6 E J,F #g s.

b.

Pressurizer Pressure.

it [ ] Si C,

RCS teta flcw rsie, F#-

~~ {

3fM APPLICABILITY: ICDE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next't hours.

SURVEILLAN:E REQUIREMENTS

\\

$bcWn C,be, 4.2.5 Each of the parameters of Tdu.*-: shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . -

@f@

g COMANCHE PEAK - UNIT 1 3(42-14 I

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RCS Lcop Isolation Valv::s 3.3.9

[

JUSTIFICATION FOR CHANGES TO W-STS REY.5

%J The following justifications are provided to explain the reasons for the differences between the

Desk Reference" W STS, Rev. 5, which is the base Standard Technical Specification and this rewritten MERITS Specification.

The number for each justification corresponds to a number appearing on the attached, marked-up Specification 3.4.1.5 from the W-STS, Rev. 5.

1.

The current STS specifications for plants with loop isolation valves are written with the assumption that the plants may be operated with N-1 loops. The current expectation is that very few plants will be licensed for N-1 operation.

Therefore, the proposed MERITS version of this specification is written such that isolated loop operation is not permitted in MODES 1 through 4, and should inadvertent isolation of a loop occur, the unit would be forced into MODE 5 or 6 where continued operation with the isolated loop would be permitted.

In addition, the LCO requirement to verify boron concentration of the isolated loop in MODES 1 through 4 is deleted since the proposed LCO will not permit loops to be isolated in MODES 1 through 4.

The proposed LCO will read as follows:

Each RCS hot and cold leg loop isolation valve shall be open with power removed from each isolation valve operator.

2.

In accordance with item 1 above, the new action statements would read as follows:

With power available to one or more loop isolation valve operators, a.

remove power from all loop isolation valve operators within 30 minutes.

NOTE All required actions of b.) must be completed whenever action b.) is entered.

b.

With one or more RCS loop isolation valves closed, immediately isolate each affected loop by closing, as required, the hot or cold leg isolation valve and be in MODE 4 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.

In accordance with item I above, the surveillance requirement reads as follows:

At least once per 31 days, verify each RCS loop isolation valve open and power removed from each loop isolation valve operator.

e i

WOG-MERITS Page 1 of 1 Rev. B 05/18/1989

Ru. B andet No. 50 445

~

REACTOR C00LAKT SYSTEM g

ten,,.ern enne i ne t.., au.. s gc3 g,o0f yS0}op40n yalas m.

LIMITING CONDITION FOR OPERATION u

u_.

.u,,

u n.e...,

2 Ol

.....n

---. -- r-_. _ -, -

..........m,........

s.

.u

....4m m_

y...

.,_.....u

.u.

r-APPLICABILITY: MODES 1, 2, 3, 4. end4, ACTION:

1

__.4..,4.s j

.u.m..

m_.._6-.. u._..4..-__-.........

7....,.

.u u_.

@.m

.m_

a...m.,......21.

4....,m_..m.,.

...s

.u n...,.........,....

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unt 1

u....

.. u 4.

.u n

are m....

w ua_

....,.u

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a..._m..........

u....

.u

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i I

SURVEILLANCE REQUIREMENTS

.....e

, u...

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1 1

COMANCHE PEAK - UNIT 1 3/4 4-7 i

i i

l RCS Isolated Ltop Startup J

-3.3.10 i

i

(

j

\\

l \\

JUSTIFICATION FOR CHANGES TO W-STS REY.5 j

The following justifications are provided to explain the reasons for the differences between the " Desk Reference" M STS, Rev. 5, which is the base Standard Technical Specification and. this rewritten MERITS Specification.

The number for each justification corresponds to a number appearing on the attached, marked-up Specification 3.4.1.6 from the M-STS, Rev. 5.

'I.

As a result of the changes described in the justification'for changes to s

LCO 3.3.9 (STS 3.4.1.5), the proposed MERITS version of this specification

-j is written such that isolated loop operation is only permitted.in MODES 5 and 6.

The proposed LCO addresses the status of loop isolation valves. As such, the existing part a.) of the STS specification becomes a prerequisite 1

(surveillance requirement).to the proposed LCO.

In addition, the proposed LCO does not need to address the reactor being suberitical since in MODES 5 and 6. LCO 3.1.1 (MODE 5) and LCO 3.8.1 (MODE 6), already cover the required shutdown margin for these MODES. The proposed LCO will read as follows:

Each RCS isolated loop shall remain isolated with:

i o

a.

The hot and cold leg isolation valves closed if boron concentration of the isolated loop is < boron concentration' of the. operating _ loops,.and b.

The cold leg isolation valve closed if the cold les temperature of the isolated loop is > [

)*F be' low the highest cold leg temperature of the operating loops.

2.

In accordance with item 1 above, the new action statements would read as follows-I With an isolated loop hot or cold leg isolation valve opened prior to i

meeting the LCO requirements, immediately close the hot and cold leg isolation valves if the boron concentration requirement is not met, or immediately close the cold leg isolation valve if the temperature requirement is not met.

3.

In accordance with item I above, shutdown margin and the surveillance requirements for verifying shutdown margin are covered by other LCO's.

Therefore, this surveillance is replaced with the following wording:

)

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to opening the hot or cold leg isolation valve in an isolated loop, verify boron concentration of isolated loop,1 the boron concentration of the operating loops.

O WOG-MERITS Page 1 of 1 Rev. B 05/22/1989 L_-__ _

1

i kV. b Docket 460. 50-445 REACTOR COOLANT SYSTEM 1

ISOLATED LOOP STARTUP [0PTIONAL) l LIMITING CONDIT!0'N FOR OPERATION 3.4.1.6

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d APPLICABILITY:,444 MODES. 5 u. 6 ACTION:

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,,.s,

_2 u_ t.. r..

3...s.

r-2

~

SURVEILLANCE REQUIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be detemined to be l

within 20'F of the highest cold leg temperature of the operating loops within l

30 minutes prior to epening the cold leg stop valve.

1 4.4.1.6.2 Th: r:::t:r :h:li 5: it: :b::' t: 5: d: 't'::1 by :t h::t 3

E

^.F.'E etth'- 30 tnet:: ;-!:r t :;;rfa; th: :: S h; :t:; ;; h:.

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CDMANCHE PEAK - UNIT 1 3/4 4-8

i 1

kdV.

A jm REACTOR COOLANT SYSTEM LOOP !$0LAT!0N VALVES l

LIMITING CONDITION FOR OPERATION shall rima,n i3rMd with:

Euh m

3.4.1.5.2 44-en RCS loop 4e-isolated, eo4atete het leg and cold leg stop a

valves closed m644) le%

e.'if The boron concentration of the isolated loop is eneter than er ::::?

r 40-the boron concentration of the operating loops, and Ths ctlL li9 stktatn,ala clts f 1p 3r.diy &n l

The tasperature of the cold leg ok the isolated loop is e64Ma 20*F h t #

l b.

j

+8 the highest cold leg temperature of the operating loops.

l l

1 APPLICABILITY: M ES 5 and 6.

r: t Of t M 2 :: : = f'i ntfer. et : tfeffed, t e t t

ACTION: S'th '2: n; r eit'- *M "t M; e-teld le; Ste; "@;er.

l 2.

)

SURVEILLANCE REQUIREMENTS i\\

4.4.1.5.2.1 The isolated loop cold leg temperature shall be detemined to be within 20*F of the highest cold leg temperature of the operating loops within l

J 30 minutes prior to opening the cold leg stop valve.

4.4.1.5.2.2 The boron concentration of an isolated loop shall be determined to j

be greater than or equal to the boron concentration of the operating loops l

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to opening either the hot leg er cold leg stop valves of an isolated loop.

6 8AAIDWD00 - UNITS 1 & 2,

3/4 4-8

[

q RCS '.:.akage Detection Instrumentation J

3.3.16 JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the i

differences between the " Desk Reference" W-STS, Rev. 5, which is the base Standard Technical Specification, and this rewritten MERITS Specification. The-i number for each justification corresponds to a number appearing on the attached marked-up Specification 3.4.6.1 from the W-STS, Rev. 5.

1.

Currently, the STS lists three independent detection systems;' containment atmosphere [ gaseous or particulate) radioactivity monitoring system; containment pocket sump level and flow monitorin L

containment atmosphere [ gaseous or particulate) g system; and either the radioactivity monitoring 1

system or the [ containment air cooler condensate flow rate) monitoring system. For the majority of Westinghouse Pressurized Water Reactors, the containment atmosphere particulate system and the gaseous system are a common system utilizing the same power source, sample point and various other common components.

The indicated change is needed to clarify that primary system leakage detection is monitored by two independent techniques and not three. Two of the three listed are a common system sharing common components. The only independence is that there are two detectors with associated electronics; one looking at a particulate filter and the other at a gas O

Should one of the common components in the system fail, both chamber.

systems will fail, thereby placing the unit in an action statement requiring plant shutdown within six hours. Plant shutdown is unnecessary in this case, since adequate capability still exists to detect primary system leakage. The containment sump monitoring capabilities ar1 still j

available and containment atmosphere airborne levels will be determined i

using grab samples. The intent of Regulatory Guide 1.45 is met by employing separats detection methods for monitoring airborne radioactivity and sump level. Airborne radioactivity is monitored using j

the containment atmosphere gaseous and particulate radioactivity-monitoring system, and the liquid volumes are monitored using the sump 3

i level and flow monitoring system. The loss of one technique is acceptable for 30 days provided the other-technique is available.

The proposed changes to the STS allows greater flexibility with both the particulate and the gaseous radioactivity monitors inoperable. The thirty day action statement allows adequate time to repair or replace the inoperable components. The proposed action statement for the inoperable airborne monitors also requires more frequent grab samplin twelve hours as opposed to once every twenty-four hours). g (once every The proposed action statement for the inoperable sump level and flow monitors has an added requirement to perform an RCS inventory balance once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

6 2.

The SURVEILLANCE REQUIREMENTS section has been changed to the new MERITS column format with stated SR number and description and the required O

FREQUENCY arranged in order of descending frequency. The new format eliminates the need for Table 4.3.3 related to this technical specification.

WOG-MERITS Page 1 of 1 Rev. B 05/13/1989

Accumulators 3.4.1 3.4 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.4.1 Accumulators LCO 3.4.1 Each ECCS accumulator shall be OPERABLE with:

a.

The isolation valve open and power to the valve removed, I

b.

A contained borated water volume 1 [

] and S [

]

gallons, c.

A boron concentration 1 [

[ and s [

] ppm, and l8 d.

A nitrogen cover-pressure 2 [

] and 5 [

] psig.

APPLICABILITY:

MODES I and 2, MODE 3 with pressurizer pressure > [

] psig.

ACTIONS CONDITION REQUIRED ACTION-COMPLETION TIME O

i A.

One accumulator A.1 Restore accumulator to I hour I

inoperable for reasons OPERABLE status.

Other than boron i

concentration outside limits.

}

B.

One accumulator B.1 Restore accumulator 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to to within boron boron concentration concentration limits.

outside limits.

1 C.

Required Actions not C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

A!!D C.2 Reduce pressurizer 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months pressure to s[

] psig.

Unit Name 3.4-1 Amendment WOG-MERITS Rev. B i

u______________

l-Accumulators 3.4.1-l O SURVE7tLANCE REQUIREMENTS SURVEILLANCE FREQUENCY' SR 3.4.1.1 Verify each accumulator isolation 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months valve fully open.

3 SR 3.4.1.2 Verify contained borated water volume-24 hours' in each accumulator within limits.

SR 3.4.1.3 Verify nitrogen cover-pressure in each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months accumulator within limits.

SR 3.4.1.4 Verify buron concentration for each 31 days accumulator within limits.

AND Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after solution volume changes l

>[-

) gallons SR 3.4.1.5 Verify power to each accumulator 31 days isolation valve operator. removed.

CROSS-REFERENCES - None.

l O

Unit Name 3.4-2 Amendment WOG-MERITS Rev. B f

1 Accumulators 3.4.1 JUSTIFICATION FOR CHANGES TO W-STS REV. 5 b/

The following justifications are provided to explain the reasons for the differences between the " Desk Reference" W-STS REV. 5, the base Standard Technical Specification used for MERITS, and this MERITS Specification.

The number for each written justification corresponds to a number appearing on the attached, marked up Specification 3/4.5.1 Accumulators from " Desk Reference" W-STS REV. 5.

1.

The plant specific numbers in the LCO were replaced with empty brackets to make the spec generic.

2.

The note applicable to MODE 3 was incorporated as per the MERITS program format.

3.

The phrase "and power to the valve removed" was added for clarity.

4.

The Required Action time for opening the isolation valve of "immediately" was changed to I hour in order to quantify the term "immediately". This time is consistent with LC0 3.4.2 ECCS Trains -

T 2 350*F, Condition C which permits both trains of ECCS ayg to be inoperable for I hour.

5.

ACTION A.1 was written to combine all conditions except for boron concentration into one condition since the Completion Times are the same.

6.

The Completion Time for restoring boron concentration was changed from 5N 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months based on the South Texas Project Technical Specifications. In the initial injection phase of the calculated response to Loss Of Coolant Accident (LOCA), the boron content of the l

accumulators is not specifically evaluated since it is not of concern during this phase of the accident. During the recirculation phase, when the ECCS is taking suction from the sump, the boron content of the accumulators is considered but the contribution is small when its volume is compared with the total volume of the RCS and the Refueling Water Storage Tank. Also, if the boron content of the accumulators is found to be in violation of the requirement, it is not possible to change the boron concentration and confirm that the new concentration meets the LCO limit in I hour.

For these reasons the NRC agreed to extend the allowed outage time for this condition from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in the South Texas Project Technical Specifications. This change is applicable to all Westinghouse plants since the cited O

reasons are generic.

7.

The term " HOT STANDBY" was changed to MODE 4 as per the convention adopted by the owners groups via the MERITS program format.

l 8.

ACTION C was written addressing the first two to eliminate duplication of ACTIONS.

)

WOG-MERITS Page 1 of 3 Rev. B

l Accumulators 3.4.1 1

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 O

i V

9.

The SURVEILLANCE REQUIREMENTS were reordered to comply with the MERITS program format.

1

10. SRs 3.4.1.1, 3.1.1.2 and 3.4.1.3 were written as separate activities for clarity.and consistency with the LCO. The Completion Times were changed from 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s-to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months based on the South Texas Project Technical Specifications and the explanation in item 6.

l

11. The surveillance activities which required verifying that the 1

isolation valves would open automatically was eliminated since this

.j LCO requires that the isolation valves are open and the power is removed. There is a surveillance, SR 3.4.1.1, to verify the open valve position. The verification of the valves being able to open -

automatically would violate this LCO.

12. The phrases "by the. absence of alarms" was eliminated from SR 3.4.1.2 and SR 3.4.1.3 since this is specifying a method of determining accumulator pressure and volume. The. method of determining these parameters is utility specific and should not be dictated by Tech Specs.

13. Surveillance Requirement 4.5.1.2 required a monthly ANALOG CHANNEL OPERATIONAL TEST to be performed on the accumulator water level and pressure channels. The deletion of this requirement is based on an NRC decision on the Callaway plant.

Prior to Callaway obtaining their license a meeting was held on November 14-16, 1983 where one of the points of discussion was the surveillance on the accumulator water level and pressure channels.

The surveillance consisted of an ANALOG CHANNEL OPERATIONAL TEST to be performed every 31 days. The NRC agreed that this surveillance could be deleted from the specification since its performance required g

either the lifting of leads or entering containment and. draining the reference leg.

The accumulator water level and pressure channels are control not protection functions in Westinghouse plants. Since they are control parameters they are not required to have a test switch.- The absence of the test switch can require the utility to lift leads to perform the ANALOG CHANNEL OPERATIONAL TEST. The NRC is opposed to the periodic and repetitive lifting of leads and therefore permitted the -

deletion of this surveillance from the specification for Callaway as part of their original license. Accumulator level and pressure are monitored to ensure that the accident analysis has not been violated.

In addition, accumulator pressure is monitored since.it is a j.

Regulatory Guide 1.9? Category 2 variable.

'O WOG-MERITS Page 2 of 3 Rev. B f

Accumulators 3.4.1 o

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The deletion of this surveillance is justified generically due to the fact that accumulator level and pressure are control functions on Westinghouse plants and, as previously discussed, may require the lifting of leads.

In addition the LCO addresses the parameters, not the instrumentation utilized to monitor the parameters.

If the purpose of the LCO were to ensure the operability of the instrumentation channels, the LCO would include the channels and an action statement. Similar situations exist for other parameters, such as RWST level and temperature, where the parameter is required to be verified to be within limits but the instrumentation is not subject to a specification dictated surveillance requirement.- These parameters are also monitored to verify that accident analysis assumptions are not violated and there is no surveillance on the instrumentation monitoring them. Verification of the parameters of interest cannot be verified unless the instrumentation monitoring them is OPERABLE. The operability of the instrumentation is ensured by the normal utility instrumentation calibration and QA programs..

The reasons presented for deletion of the ANALOG CHANNEL OPERATIONAL TEST can also be extended to the CHANNEL CALIBRATION surveillance.

For these reasons and since the method of verifying accumulator pressure and level is a utility specific item, not to be dictated by.

tech specs, the CHANNEL CALIBRATION surveillance was also deleted.

14. Deleted.

l8

15. The phrase " increase of greater than or equal to [1% of tank volume]"

was replaced with " changes 1 [

] gallons" to require verification whenever a volume change had taken place, either an increase or decrease. The percentage was changed to gallons to be consistent with analyses units.

1

.1 j

i l

l l

l WOG-MERITS Page 3 of 3 Rev. B i

t

7ev.3 EMERGENCY CORE COOLING $Y$7 EMS lh 3/a.5.1 ACCUMUL'aTORS COLD LEG INJECTION gar j

LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumul r shall be OPERA 8LE wi :

& m tok &m non*

The Wsenadisolation val open,wP2 :n :r.--. 4 a.

4 vo une e b we U [785 8

6-L [8q1) saiions, Q'

**"'*ia'd 6'r**'8 *a c.

A boron concentration of between (19002 and [2100) poe M 1

d.

A nitrogen cover pressure gLo, etJwe [385] M 481) psig.

?_

t

)

APPLICABILITY: 9CDES 1, 2, and 3".

ACTION:

s.

With one cold leg injection accoulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within I hour or be in at leastan :muen within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer e

re to 'ess than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 7

b.

With one cold leg injection accumu or r

1/ due to the D

isolatipn valve being closed.

erliamediatelylopen the isolation valve or be in at leastlHOT STAN08YIwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 2000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold leg injection accoulater shall be demonstrated.

-0PERA8LE:

~

At least once per h urs by:

I2 a.

1)

Verifying once of alams,' the contained borated water volume an kurs in the tank.l. and 2)

Verifying that each cold leg injection accumulator isolation valve is open.

l l

=Pressuria.r pressure above 1000,sig.

COMANCHE PEAK - UNIT 1 3/4 5-1 l

l

I Ed.S

)

I y

EMERGENCY CORE COOLING $YSTEMS

\\

l SURVEILLANCE REQUIREMENTS (Continued) gepafts3[ ]VW

. At.least once ner.31 days and[hin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution vol-

)

b.

use_ increase of greater than or equal to CJ of tank volumeEby veri-l f

rying the baron concentration of the solvt:

on in the water-f t11ed j

accumulator; c.

At least gnce per 31 days when the RCS pressure is above 1000 psig by l

verifying that power to the isolation valve operator is disconnected i

by removal of the breaker from the circuit.

1 d.

At least once per 18 months by verifying that each accumulator isola-tion valve opens automatically under each of the following conditions:

1)

Nn an actual or a simulated RCS pressure signal exceeds the

~

P-11 (Pressurizer Pressure Block of Safsty Injection) setpoint, and 2)

Upen receip't of a Safety Injection test signal.

4.5.1.1.2 Each accumulator water level and pressure channel shall be demon-1 strated CPERABLE:

)

At least once per 31 days be the perf'orsancq of an ANALOG CHANNEL a.

OPERATIONAL TEST, and b.

At least once per 18 months by the performance of a CHANNEL CALIBRATION.

j I

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t e

e G

(

CDMANCHE PEAK - UNIT 1 3/4 5-2

ECCS Trains - Tavg 1 350'F 3.4.2 3.4 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.4.2 ECCS Trains - T,yg 2 350*F LCO 3.4.2 Two ECCS trains shall be OPERABLE, with each train comprised of:

i a.

An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection (SI)-

Signal and [ automatically] transferring suction to the containment sump on recirculation, b.

One OPERABLE centrifugal charging pump,

[c. One'0PERABLE SI pump,]

[d. One OPERABLE Residual Heat Removal (RHR) pump, and]

[e. One OPERABLE RHR heat exchanger.]

i

..........................--N0TE----------------------------

In MODE 3 both SI pump flow paths may be isolated by closing the isolation valves for a period s 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform

$$55$.. "$$. $I "!.S$5..

8 1

t APPLICABILITY:

MODES I, 2, and 3.

NOTE----------------------------

LCO 3.0.4 and SR 3.0.4 are not applicable for entry into MODE 3 for' the centrifugal charging pumps and thi SI pumps -

declared inoperable pursuant to LCO 3.3.18, provided the centrifugal charging and SI pumps are restored to OPERABLE status within at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following entry into MODE 3 or prior to the temperature of one or more of the Reactor Coolant System cold legs exceeding 315'F, whichever comes first.

i Unit Name 3.4-3 Amendment WOG-MERITS Rev..B

)

f

a ECCS Trains'- T 1 350'F ayg 3.4.2 i

('"'N ACTIONS

'--)

j CONDITION.

REQUIRED ACTION COMPLETION TIME j

A.

One ECCS train A.1 -

Restore train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I

inoperable.

' OPERABLE status.

l

1 I

B.

One or more com)onents

.B.I.

Restore components to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in each train ()ut not OPERABLE status.

the same component) 8 inoperable, but 100% of i

required flow available.

I J

C.

Two ECCS trains C.1 Restore one ECCS train I hour inoperab7.e, except to OPERABLE status, as specified in Condition B.

1 D.

Required Actions not D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Times.

AND a

l D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I

\\

l 1

J l

l i

l 4

1 N

Unit Name 3.4-4 Amendment WOG-MERITS Rev. B l

l ECCS Trains - T 2 350*F avg 3.4.c U]

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify the following valves are in the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months listed position with power to the valve operator removed.

Valve Number Position Function

[

]

[

]

[

]

[

]

[

]

[

]

[

]

[

.]

[

]

SR 3.4.2.2 Verify ECCS piping is full of water by 31 days venting ECCS pump casing and accessible piping high points.

b)

V SR 3.4.2.3 Verify each ECCS manual, power-operated, 31 days and automatic valve (excluding check valves) in the flow path, that is not locked, sealed or otherwise secured in position, is in its correct position.

I i

SR 3.4.2.4 Verify each ECCS pump's developed head at As specified l

the test flow point is 2 the required by SR 3.0.5 developed head.

l SR 3.4.2.5 Verify each ECCS train automatic valve 18 months (excluding check valves) in the flow path

)

actuates to its correct position on (SI actuation and [ automatic] switchover to containment sump) signal.

(continusd)

O Unit Name 3.4-5 Amendment WOG-MERITS Rev. B

_.m_______

ECCS Trains - T 1 350'F ayg 3.4.2 p

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY:

SR 3.4.2.6 Verify, for each ECCS throttle valve 18 months listed below, each position stop is in its correct position.

Valve

~ Valve Number Position

[

]

[

]

[

]

[

]

1

[

]

[

]

i SR 3.4.2.7 Verify each ECCS pump starts automatically 18 months on receipt of an SI Signal.

SR 3.4.2.8 Verify ECCS flow by performing a flow During V

balance test for each pump separately,

shutdown, as listed below:

following modifications Centrifugal charoino lines - sinale oumo:

that could alter flow a.

The sum of the injection line flow rates, excluding the highest flow rate, is 2 [

] gpm.

b.

Total pump flow rate is s [

] gpm.

SI lines - sinole Dumo:

c.

The sum of the injection line flow rates, excluding the highest flow rate, is 1 [

] gpm.

d.

Total pump flow rate is s [

] gpm.

RHR lines - sinole Dumo:

e.

The sum of the injection line flow rates is 1 [

] gpm.

g Unit Name 3.4-6 Amendment WOG-MERITS Rev. B I

ECCS Trains -. T 1 350*F ayg 3.4.2 I

CROSS-REFERENCES TITLE NUMBER Containment Integrity 3.5.1 1

1 I

J O

l I

i O

Unit Name

4. -

Amendment WOG-MERITS Rev B

l ECCS Trains -'T 1.350*F ayg 3.4.2 i

i JUSTIFICATION FOR CHANGES TO W-STS REV. 5~

The following justifications are provided to explain the reasons for the differences between the " Desk Reference" W-STS REV. 5, the base Standard.

Technical Specification used for MERITS, and this MERITS Specification.

l The number for each written ~ justification corresponds.to a number appearing on the attached, marked up Specification 3/4.5.2 ECCS Trains;- Tayg 2;

.]

350*F from " Desk Reference" W-STS REV.--5.

1.

The note:in the LCO was'added to allow Pressure Isolation. Valve (check-valve) leak testing in MODE.3 as per SR 3.3.15.1.

The flow path may-I be isolated for a short period, under controlled conditions,:to-

' perform tests ensuring the continued availability of the, flow path.

j The flow path is readily restorable from the control room, and the -

9 benefits (ECCS injection) of. ensuring the 0PERABILITY of the flow path -

l outweigh'the decreased availability during this period.,

During discussions with the NRC staff regarding this proposed change, a' question was raised regarding the ability of the ECCS to respond to'.

a LOCA while check valve leakage surveillance was being performed.

During this surveillance, the unit would be in MODE 3 at a reduced pressure of approximately 1000 psi with selected SI valves closed.

This configuration would limit the available ECCS. flow to mitigate-l effects of a LOCA.

A large spectrum of LOCA break sizes has been evaluated in the' FSAR accident analysis for MODE 1 conditions. However, MODE 3 conditions O-at approximately 1000 psi are far below the conditions for.which the reactor coolant system has been designed that a large LOCA 'is unlikely -

and, for all practical purposes, can be assumed not to occur.-

8 Engineering studies, leak-before-break analysis and operating-experience have shown through wall' cracks in the RCS Class 1 pressure-i boundary piping, greater.than six inches in diameter, are highly unlikely. Therefore, for purposes of this discussion, a credible LOCA of piping less than six inches was evaluated.

A shutdown LOCA scoping study was performed by Westinghouse for a-standard four loop' plant after a reactor trip.L Based on this study, it was estimated that at least 20 minutes would be.available to initiate SI flow from a centrifugal charging pump to prevent-j significant core uncovery for breaks up'to three inches in diameter.

Initiation of SI flow within this time may not preclude core uncovery, but is expected to limit the. fuel. cladding heatup to.less than full l

power small break LOCA results provided in-the FSAR. For breaks i

larger than three inches and up to six inches in diameter, operator 1

action to initiate SI from a centrifuga1' charging pump-is estimated to be required within approximately 10 minutes. Additional operator action may be required, depending upon break size, within one hour of the event initiation to start an additional centrifugal charging pump or safety injection pump or depressurize the RCS using the steam generators and to start an RHR pump.

O i

WOG-MERITS Page 1 of 4 Rev. B

't i

l ECCS Trains - T,yg 2 350*F l

3.4.2 i

m JUSTIFICATION FOR CHANGES TO W-STS REV. 5 I

i I

Several indications are available to the operator to identify that a' LOCA is in progress..These include: loss of pressurizer level,.RCS u

pressure decrease, loss of RCS subcooling, radiation alarms inside containment, containment pressure increase and sump water level increase. When a LOCA has been identified, the operator would initiate a SI as required by the emergency procedures..The scoping study assumed. flow from only one centrifugal charging pump and one RHR pump to the cold legs (even through the RHR pumps would be aligned to all four cold legs in MODE 3), so it is evident that a substantial amount of additional ECCS charging flow would be available.

In addition, the emergency response procedures direct the operator to-verify ECCS flow. Since these valves could only be closed for up to a two hour period, the operator would be aware these valves are closed f

and would open them to restore all availabh ECCS flow. Simple manual i

action by the operator in the control room can reopen the closed valves.

The scoping study assumed the LOCA occurs two hours after a reactor trip. Typically, this check valve surveillance will be performed after a refueling outage or after the plant has been in MODE 5 for longer than three days, much longer than the two hours assumed in the scoping study. Therefore, the surveillance would be conducted when the initial fuel rod temperature and decty heat levels are less than CN those assumed in the scoping study. This would also extend the times Q

before operator action would be required to longer than the times-resulting from the scoping study.

Normal operating pressure in MODE 1 serves as a more severe condition which demonstrates that pipe ruptures below normal operating pressure are highly unlikely since additional margins of safety exist at lower i

pressure. The condition that could lehd to a pipe rupture, a large through wall crack, would be identified during operation. However, even with the presence of.such a crack, the piping system would remain stable and a pipe rupture would be unlikely at the reduced RCS pressure.

l In summary, a LOCA occurring during MODE 3 conditions at reduced RCS pressure is highly unlikely. A LOCA occurring in MODE 3 during the short period of time when the SI valves are closed concurrently is even more unlikely. However, if a LOCA did occur, the preceding discussion demonstrates that sufficient time would exist for operator response given the plant configuration.

r This allowance was licensed at the Byron and Seabrook units. Similar i

type studies for 2 and 3 loop plants are anticipated to provide-similar results which would permit check valve testing as described by this Note.

((

WOG-MERITS Page 2 of 4 Rev. B

.1 ECCS TrainsL-Tayg L 350*F 3.4.2 JUSTIFICATION FOR CHANGES TO W-STS REV.

S'-

- 2. -

The note in the Applicability Section was-included to allow pump.

operability to change when cressing the MODE 3 boundary, when this boundary temperature is at or near the Cold Overpressure Prevention (COP)l arming. temperature.

It is impossible for an. operator to change pump operability for centrifuga1' charging pumps and safety-injection pumps the moment the RCS T,yg is greater.than 350*F.

The note 1s. specific to those plants:that have the MODE 3 boundary ~

~

temperature either below or near C0P armi9g temperature.

When the' MODE 3 boundary temperature is above and significantly -

different than the C0P arming temperature, the need for.this allowance.

on changing pump operability is nonexistent since the operator ~would be afforded sufficient. time.to change the pump from inoperable to OPERABLE status.

1 During ascensions from MODE 4 to 3, all but the allowable pump (s) is-1 required to be inoperable per-LCO 3.3.18, Cold Overpressure Prevention, at_or below a specified temperature. When this temperature, referred to as the C0P arming temperature, sis.above_or nearly the same as the MODE 3 boundary temperature the operator must-g instantly change the pump from inoperable to OPERABLE.to be in,

compliance with LCO 3.4.2.

Such a change in operational status'is -

l impossible to perform instantaneously. _ The subject note permits the operator to make the operational status change'in an orderly manner.

I The limits imposed by the note are a time limit and a temperature limit above the mode boundary temperature. The time limit ensures that the unit would not be configured indefinitely with only one pump-OPERABLE. The temperature limit ensures that the unit would not be l

much beyond the mode boundary temperature with only one OPERABLE pump.

1 The health and safety'of the public would not be' jeopardized since the probability of a LOCA during the short time period w1en two pumps are required but only one is OPERABLE is very remote.

In addition, the temperature limit _is not significantly different than the MODE 4 temperature where only one pump is required to be OPERABLE.

The note was licensed at Callaway, Seabrook, Vogtle, South Texas and Comanche Peak and should _be applicable to all plants that' have their C0P arming temperature either above or nearly.the same as the MODE 3-boundary temperature.

3.

The ECCS actuation. deportability has been transferred to Section-5.9.2.

4.

Condition B was added to allow operability when 100% of required flow is available even with inoperable components in separate trains. This condition permits more flexibility in operation when the system is not more severely degraded than Condition A.

)

5.

Conditions C and D replace LCO 3.0.4, which was the only alternative in the original spec.

WOG-MERITS Page 3 of 4 Rev. B I

ECCS Trains.- T 2 350"F.

ayg 3.4.2 JUSTIFICATION FOR CHANGES TO W-STS REV. 5 6.

Surveillance reordered in the order.of their frequencies as per the MERITS program format. -

7.

" Subsystems" changed to " Trains" for consistency.

8.

This surveillance will be relocated since it is used to protect the RHR system and has nothing to do with tile ECCS.

9.

The word " discharge" was deleted to avoid restricting the surveillance to only discharge piping. Suction piping is to be included since the issuance of an NRC Bulletin on Gas Binding of Pumps at Farley.

10. Surveillance verifying the absence _ of_ loose debris were deleted since these activities are housekeeping requirements and should not be part of tech specs. These visual inspections are part of procedures to establish CONTAINMENT INTEGRITY.

11. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months surveillance, g.1, was deleted since these requirements are part of maintenance activities. The 18 month frequency was retained and should provide adequate assurance that incorrect positions will be detected. The phrase " electrical and/or mechanical" was deleted since it is unnecessary and the surveillance reworded slightly _for clarity.

O

12. The word " independent" was deleted since the some of the trains are capable of crossing over to each other.

13. The phrase "dur"'ig shutdown" was deleted so as to not require a shutdown to perform the surveillance. The phrase "each pump.

separately" was added to make it clear that each pump should be verified separately.

14. Surveillance Requirement 3.4.2.4 provides' verification that the pump's developed head is greater than or equal to the required developed head at one point on the pump's required head curve. This is an i

alternative to the STS, Rev. 5 which required that the pumps be tested i

at the recirculation flow point. However, the same.information is obtained by testing at any point along the pump's required head curve. This developed head technique is proposed to allow more j

flexibility in testing. An example of this would be testing in a normal operational alignment, rather than a recirculation alignment.

Neither the. intent nor the meaning of the existing specification is changed. Testing at any one flow point.is sufficient to warn of 1

abnormal pump performance. Under ASME requirements, a pump head curve is established and any change in this established pump head curve is monitored by periodic ASME-Section XI pump testing. Allowable pump 1

degradation is limited by the ASME Section XI requirements or the pump l

operating performance assumptions used in the accident analyses, whichever is greater.

15. The surveillance to inspect the containment sump for debris was i

O deleted since these activities are housekeeping requirements and should not be part of tech specs but rather plant procedures.

WOG-MERITS Page 4 of 4 Rev. B

RWST 3.4.5 l

3.4 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.4.5 Refuelina Water Storace Tank (RWST)

LCO 3.4.5 The RWST shall be OPERABLE with:

I a.

A deliverable borated water volume 2 [

] gallons, b.

A boron concentration 1 [

] and 5 [

] ppm, and-l8 c.

A borated water temperature 1 [ ] and i [

] *F.

I APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RWST inoperable.

A.1 Restore RWST to I hour OPERABLE status.

B.

Required Action not B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

Ag B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months O

l Unit Name 3.4-12 Amendment WOG-MERITS Rev. B

RWST 3.4,5

(

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify deliverable borated water volume 7 days in RWST within limit.

SR 3.4.5.2 Verify boron concentration in RWST 7 days within limits.

SR 3.4.5.3 Verify RWST borated water temperature

NOTE-----

within limits.

Only required

.if outside air temperature is < [ ]'F or > [

]'F l

l 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months I

1 l

CROSS-REFERENCES 1

TITLE NUMBER I

1 ECCS Trains - T 2 350'F 3.4.2 l

avg ECCS Trains - Tayg < 350*F 3.4.3 i

l Containment Integrity 3.5.1-I

[ Containment Internal Pressure]

[3.5.12) j J

i I

1

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Unit Name 3.4-13 Amendment WOG-MERITS Rev. B l

j RWST 3.4.5 JUSTIFICATION FOR CHANGES TO W-STS REV. 5 l

The following justifications are provided to explain the reasons for the differences between the " Desk Reference" W-STS REV. 5, the base Standard Technical Specification'used for MERITS, and this MERITS Specification.

The number for each written justification corresponds to a number appearing on the attached, marked up Specification 3/4.5.5 Refueling Water Storage f

Tank from ' Desk Reference" W-STS REV. 5.

J l

t 1.

-The actual values were replaced by empty brackets to make the spec generic.

2.

The times of 6 and 30 were changed to 6 and 36 to show cumulative time

'for the action.-

3.

The terms HOT STANDBY and COLD SHUTDOWN were replaced with MODE 3 and MODE 5 as per the MERITS program format.

4.

The surveillance were written separately to be consistent with the LCO.

5.

Deleted.

l8 O

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O WOG-MERITS Page 1 of 1 Rev. B j

l l

1

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%.B 50RON INJECTION SYSTEM

,p M, &

7/4.5.5 REFUELINGWATERSTORAGiTANK

~

LIMITINGCONDITIONFOROPERATiON 3.5.5 The refueling water storage (RWST) shall be OPERABLE with:

a.

A minimum contained bor d water volume of

pa11ons, b.

A boron concentrationIof between [2000) and [2200) ppe of boron, 8

~

c.

A minimum solution temperature of [353'F, and d.

A maximum solution temperature of {100]'F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the INST inoperable, restore the tank to OPERABLE status within I hour or be in at least NOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

D D

o SURVEILLANCE REQUIREMENTS 4.5.5 The RwST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water volume in the tank, and 2)

Verifying the boron concentration of tho' wetar.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside) air temperature is less than (353'F or greater than

[100'F.

COMANCNE PEAK - UNIT 1 3/4 5 13 i

Containment Integrity 3.5.1 3.5 CONTAINMENT SYSTEMS 3.5.1 Containment Intecrity LCO 3.5.1 CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more containment A.1 Restore containment

NOTE-----

isolation devices isolation devices to Completion inoperable on one or OPERABLE status.

Times for all more penetrations.

Required Actions of.

Condition A are on a per penetration basis

[

I hour B

A.2

NOTE----------

A check valve, with flow through the valve secured, is a deactivated automatic valve secured in the closed position.

Isolate each affected I hour penetration with a locked closed manual valve, a deactivated automatic valve secured in the closed position, a blind flange, or a closed system.

B (continued)

Unit Name 3.5-1 Amendment WOG-MERITS Rev. B I

l Containment Integrity 3.5.1 ACTIDNS CONDITION iRmfrRED ACL10N COMPLETION TIME A.

(continued) 13.3 Wer-iTv euch affected I hour 2

mliwt.isn is isolable pr with sne atrismatic iisrAartun. valve.

f A3.2.11 % stone rentainment 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months isn'imiinn devices to i

mr.twcBliE status.

E x 3., p..y

..-.__. -- NDT E - - - - - - - - -

A chett walve, with j.

ifla>w through the li yaQur secured, is a atriac1Tiws11'ed automatic l

mD.w: sfcured in the rilmsud pos.ition.

'I hsfluir earh affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

penrtrzt-ion vith a ilst4rd cinsed manual va'Ine,, a tiftactivated audmatir ulve senurad in the closed

'l mnritiivn., a blind

l ifl'amgt,, or a closed

~

r,pmer.

(continued)

Unit Name

3. :5-7 Amendment Rev. B WOG-MERITS

Containment Integrity 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One air lock door

NOTE----------

NOTE------

or one interlock Entry and exit through Completion mechanism inoperable on closed or locked doors.

Times for all one or more air locks.

when the inner air lock Required door is inoperable, is Actions of l

permissible for Condition B l

performance of air lock are on a per repairs.

air lock basis i

B.1 Verify an OPERABLE 15 minutes air lock door is closed in each affected air lock.

E 9

B.2 Restore air lock doors 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and interlocks to OPERABLE status.

B.3.1 Close and lock an 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE air lock door in each affected air lock.

E B.3.2 Verify an OPERABLE Once per air lock door in each 31 days affected air lock is locked closed.

(continued)

Unit Name 3.5 3 Amendment WOG-MERITS Rev. B l

Containment Integrity 3.5.1

.n ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

One or more air locks C.1 Verify an air lock door

NOTE-----

inoperable for reasons closed.

Completion Times for all other than Condition B.

Required' Actions of Condition C are on a per air lock basis 15 minutes E

C.2 Maintain an air lock Until air lock I

door closed, is restored to l

OPERABLE i

status

{

AND C.3 Restore each affected 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months air lock to OPERABLE status.

1 D.

CONTAINMENT INTEGRITY D.1 Restore CONTAINMENT 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

l not maintained for INTEGRITY.

reasons other than Conditions A, B, or C.

E.

Required Actions not E.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 1

met within required Completion Time.

E E.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months L

J l

I Unit Name 3.5-4 Amendment WOG-MERITS Rev. B

Containment Integrity 3.5.1 SURVEILLANCE REQUIREMENT.S SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify penetrations not capable of being 31 days isolated by OPERABLE automatic containment isolation valves, and required to be isolated during accident conditions, are isolated outside containment by locked closed manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

SR 3.5.1.'

Verify the stroke time of each power-As specified operated or automatic containment by SR 3.0.5 isolation valve (excluding check valves) is within limit.

SR 3.5.1.3 Verify penetrMions not capable of being Within 92 days isolated by C h RABLE automatic containment prior to isolation valves, and required to be entering e

isolated during accident conditions, are MODE 4 isolated inside containment by locked from MODE 5 closed manual valves, blind flanges, deactivated automatic valves secured in their closed positions, or a closed system.

SR 3.5.1.4 Verify each air lock door interlock 6 months mechanism OPERABLE.

[18 months for subatmospheric)

(continued)

O Unit Name 3.5-5 Amendment WOG-MERITS Rev. B

Containment Integrity 3.5.1 SURVEILLANCE REQUIREMENTS (continued) l.

SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify containment leakage rates in

NOTE-----

accordance with [10 CFR 50, Appendix J and SR 3.0.2 is approved exemptions or the Containment not applicable 4

Leak Rate Test Program).

[As specified by 10 CFR 50, Appendix J and approved exemptions or the Containment Leak Rate Test Program)

SR 3.5.1.6 Verify each [ Phase 'A"] isolation valve 18 months actuates to its isolation position on a

[ Phase "A") Isolation Signal.

SR 3.5.1.7 Verify each [ Phase "B") isolation valve 18 months

\\

actuates to its isolation position on a

[ Phase *B") Isolation Signal.

SR 3.5.1.8 Verify each purge supply and exhaust 18 months isolation valve actuates to its isolation position on a Containment Purge Isolation Signal.

l

[SR3.5.1.9)

Verify containment structural integrity As specified in accordance with the Containment by the Structural Integrity Test Program.

Containment Structural Integrity 1

Test Program.

I l

(continued)

I v

Unit Name 3.5-6 Amendment WOG-MERITS Rev. B

Centsinment Intcgrity 3.5.1

(

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

[SR3.5.1.10)

Cycle each weight or spring loaded check

[92 days) valve testable during plant operation, 1 through one complete cycle of full travel and verify that each check valve remains closed when the differential pressure in the direction of flow is s [

] psid and opens when the differential pressure in the direction of flow is 2 [-

] psid and < [

] psid.

g

[SR 3.5.1.11]

kycle each weight or spring loaded check

[18 months) valve not testable during plant operation, through one complete cycle of full travel and verify that each check valve remains closed when the differential pressure in the direction of flow is s [

] psid and opens when the differential pressure in the direction of flow is 2 [

] psid and < [

] psid.

8

,\\

CROSS-REFERENCES TITLE NUMBER RCS Pressure Isolation Valve Leakage 3.3.15 ECCS Trains - T 2 350 *F 3.4.2 ayg 3.4.3 ECCS Trains - Tavg < 350 *F

[ Purge and Exhaust Isolation Radiation Instrumentation]

[3.5.2)

[ Containment Spray System]

[3.5.3]

[ Recirculation Spray System]

[3.5.4)

(Vacuum Relief Valves]

[3.5.7)

[ Hydrogen Monitors]

[3.5.17]

[ Hydrogen Recombiners)

[3.5.18)

[Hain Steam Isolation Valves]

[3.6.3]

[ Main Feedwater [ Isolation) Valves)

[3.6.4]

[ Auxiliary Feedwater System]

[3.6.5)

LComponent Cooling Water System]

[3.6.8]

Service Water System]

[3.6.9) f (w

Unit Name 3.5-7 Amendment WOG-MERITS Rev. B I

C*ntainment Integrity B 3.5.1 B 3.5 CONTAINMENT SYSTEMS B 3.5.1 containment Intearity BASES BACKGROUND CONTAINMENT INTEGRITY is established when:

a.

Penetrations required to be isolated during accident conditions are either:

1. Capable of being isolated by an OPERABLE containment automatic isolation system, or

2. Isolated by locked manual valves, blind flanges, deactivated automatic valves secured in their closed positions, or a closed system, or

3. Opened under administrative control on selected penetrations (Ref.1);

b.

All equipment hatches are closed;

[c. All shutdown purge valves are closed;]

d.

Both doors of an air lock are closed except for entry and exit, then at least one air lock door shall be closed; and e.

The containment leakage rates are within their required limits.

CONTAINMENT INTEGRITY requires the existence of a structurally sound containment such that CONTAINMENT INTEGRITY aill limit the leakage of f$ssion product radioactivity from containment to the environment under the pressure and temperature conditions that may exist during a Design Basis Accident (DBA).

The containment is a reinforced concrete structure with a l

cylindrical wall, a flat foundation mat, and a dome roof.

The foundation slab is reinforced with conventional mild-steel reinforcing. The inside surface of the containment is lined with a carbon steel. liner to ensure a high degree of leak tightness.

[ Dual' and Ice Condenser containments utilize an outer concrete building for shielding and an inner steel containment' for leak tightness.)

(continued)

Unit Name B 3.5-1 Revision WOG-MERITS Rev. B

1 C:ntainment Integrity I

B 3.5.1 l

BASES BACKGROUND Eauioment Hatch (continued)

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment during unit shutdown. The equipment hatch is secured in 1

position by swing bolts positioned around its periphery.

The equipment hatch has been designed _and tested to certify j

its ability to withstand a pressure in excess of the s

maximum peak pressure resulting from the CONTAINMENT INTEGRITY. limiting DBA [ Loss of Coolant Accident (LOCA)],

such that its closure ensures CONTAINMENT INTEGRITY.

Containment Shutdown Purae System The containment shutdown purge system operates to supply outside air into the containment for ventilation and j

cooling or heating, and may-also be used to reduce the

]

concentration of noble gases within the containment prior 4

to and during personnel access. The containment isolation i

valves for this system are [normally) maintained closed in MODES 1 through 4 to ensure leak tight penetrations.

1

(

[ Containment Miniource System The containment minipurge system operates: a) to reduce the l

concentration of noble gases within containment prior to l

and during personnel access, and b) to equalize internal l

and external pressures. Since the valves used in the minipurge system are designed to meet the requirements for automatic containment isolation valves, these valves may be I

opened in MODES 1 through 4.]

Containment Isolation Devices In order to minimize containment leakage, penetrations not required for accident mitigation during a DBA are provided with isolation devices. These isolation devices consist of either passive devices or active automatic devices. Locked closed manual valves, deactivated automatic valves secured in their clused position (including check valves with flow I

through the valve secured), blind flanges, and closed systems are considered passive devices. Closed systems are i

those systems designed in accordance with General Design Criterion 57 (Ref. 2) and are identified as such in FSAR, Section[

] (Ref. 3). Check valves, or other automatic l

/

(continued)

(

1 Unit Name B 3.5-2 Revision l

WOG-MERITS Rev. B l

Containment Integrity B 3.5.1-BASES BACKGROUND valves designed to close' following an accident without (continued) operator action,,are considered active devices. Two-isolation devices are provided in_' series for each mechanical penetration, such that no single credible failure or malfunction of an active component can cause a loss of isolation, or result in a leakage rate that exceeds limits assumed in the safety analyses.

Automatic isolation signals are produced during accident conditions.- Containment' [ Phase "A"] isolation occurs upon receipt of a (Safety Injection) Signal. The [ Phase "A")

Isolation Signal isolates nonessential process lines in order to minimize leakage of fission product radioactivity. Containment [ Phase 'B") isolation occurs-upon receipt of.a [ Containment' Pressure--High-High] Signal and isolates the remaining process lines, except systems required for accident mitigation.

[Inadditiontothe isolation signals above, the purge and exhaust valves.

receive an isolation signal on [ Containment Radiation-- _

High). As a result, the containment isolation valves (and blind' flanges) help ensure that the. containment atmosphere will be isolated from the environment in the event of a release of fission product radioactivity to the containment atmosphere as a result of.a DBA.

Personnel Airlocks

[

] containment air locks, which are part of the containment pressure boundary, provide a means.for personnel access during all' MODES of unit operation. The air lock doors have been designed and certified' capable of withstanding a pressure in excess of the maximum peak pressure resulting from the CONTAINMENT INTEGRITY. limiting DBA, such that closure' of a single door assures ~

CONTAINMENT INTEGRITY. Each of the doors is provided with double gasket seals to provide pressure integrity.

An air lock door interlock mechanism prevents simultaneous opening of both doors and, therefore, ensures CONTAINMENT INTEGRITY is maintained throughout periods of access to-containment. During periods of unit shutdown when CONTAINMENT INTEGRITY is not required, the door interlock mechanism may be disabled,'. allowing both doors of an air lock to remain open _for extended periods when frequent containment entry is necessary.

[Each personnel air lock is provided with limit switches on both doors that provide control room indication of door position. Additionally,-

(continued)

Unit Name B 3.5-3 Revision WOG-MERITS Rev. B

C ntainment Integrity B 3.5.1 BASES BACKGROUND control room indication is provided to alert the operator (continued) whenever an air lock door interlock mechanism is defeated.]

Maintaining CONTAINMENT INTEGRITY limits leakage of fission product radioactivity from containment to the environment.

Loss of CONTAINMENT INTEGRITY could cause SITE BOUNDARY doses, in the event of a DBA, to exceed values given in 10 CFR 100 (Ref. 4).

APPLICABLE The safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate, such that, in conjunction with the other containment systems and Engineered Safety Feature Systems, the release of fission product radioactivity subsequent to a DBA will not result in doses in excess of the guideline values-specified in 10 CFR 100.

l The DBAs which result in a challenge to CONTAINMENT INTEGRITY from high pressures and temperatures are a LOCA, a Steam Line Break (SLB), and a Rod Ejection Accident t

(REA).

In addition, release of significant fission product

\\

radioactivity within containment can occur from a LOCA or a q

REA.

In the DBA analyses it is assumed that CONTAINMENT INTEGRITY is intact at event initiation, such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of [

] percent'of containment volume per day. This leakage rate, used in the evaluation i

of offsite doses resulting from accidents, is defined in i

10 CFR 50, Appendix J (Ref. 5), as L,; the maximum

]

allowable containment ' leakage rate at the calculated maximum peak containment pressure (P ) resulting from the a

limiting DBA [LOCA). The allowable leakage rate represented by L forms the basis for the acceptance j

a criteria imposed on all containment leak rate testing. For this unit, La=[ ] and P, - [

).

Satisfactory leak rate test results are a requirement for the establishment of CONTAINMENT INTEGRITY. The acceptance (continued)

O, Unit Name B 3.5-4 Revision j

WOG-MERITS Rev. B J

l

l C:ntainrent Integrity B 3.5.1 BASES APPLICABLE criteria applied to accidental releases of fission product SAFETY ANALYSES radioactivity to the environment are given in terms of.

(continued) total radiation dose received by: 1) a member of the general public who remains at the exclusion area boundary for two hours following onset of the postulated fission product radioactivity release, or 2) a member of the general public who remains at the low population zone for the duration of the accident. The limits established in 10 CFR 100 (Ref. 4) are a whole body dose of 25 Rem, or a i

300 Rem dose to the thyroid from iodine exposure. The worst case two-hour dose anticipated at the exclusion area boundary occurs following the postulated worst case DBA.

The worst case DBA is an overly conservative analysis of the ILOCA) event for which a significant instantaneous release of fission product radioactivity from the core is postulated.

CONTAINMENT INTEGRITY involves structures, systems and components that are part of the primary success path and which function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. As such, it satisfies the requirements of Criterion 3 of the NRC O

Interim Policy Statement (Ref. 6).

LCO compliance with this LCO ensures a containment configuration that limits containment leakage rates to less than the values assumed in the safety analyses. As a result, radiation exposures at the SITE BOUNDARY will be maintained within the limits of 10 CFR 100 following the limiting DBA.

f APPLICABILITY CONTAINMENT INTEGRITY must be maintained in MODES 1, 2, 3 and 4, when a DBA could cause a substantial increase in fission product radioactivity in containment.

In MODES 5 or 6, CONTAINMENT INTEGRITY is not required because of the l

l pressure and temperature limitations in these MODES.

LCO 3.8.4, Containment Building Penetrations, contains requirements for maintaining containment closure during CORE ALTERATIONS or movement of irradiated fuel within containment in MODE 6.

l (continued)

)

l Unit Name B 3.5-5 Revision l

WOG-MERITS Rev. B l

.l

C ntainment Intcgrity B 3.5.1 BASES (continued)

ACTIONS A.1. A.2. A.3.1. A.3.2.1. and A.3.2.2 Required Actions A.1, A.2, A.3.1,'A.3.2.1 and A.3.2.2 q

address penetrations that are provided with 2 isolation devices. Should one or more devices become inoperable on q

one or more penetrations, the potential exists that allowable leakage rates may be exceeded in the event of a DBA. Preferably, the isolation device or devices will be restored to an OPERABLE status in accordance with Required Action A.I.

The Completion Time of I hour for Required Action A.1 is based on engineering judgment consistent with the severity of a loss of CONTAINMENT INTEGRITY.

If the inoperable device or devices cannot. be restored to an OPERABLE status within the Completion Time of I hour, either Required Action A.2 must be completed to preclude exceeding allowable design leakage rates, or Required Actions A.3.1 and A.3.2.1 or A.3.2.2 may be perfonned to allow an additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months before action must be taken to reduce power.

Required Action A.2 ensures that the penetration is i

isolated by a passive isolation device as previously l

described in the Background section. The Completion Time of I hour is based on engineering judgment consistent with J

the severity of a loss of CONTAINMENT INTEGRITY.

i Required Action A.3.1 establishes that penetrations with inoperable isolation devices are still capable of being isolated by an automatic isolation device. With the i

provision of Required Action A.3.1. established within the Completion Time of I hour, an additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is permitted for restoring the device or devices per Required Action A.3.2.1 or for isolating the affected penetrations per Required Action A.3.2.2.

The Completion Time of. I hour for Required Action A.3.1 is based on engineering judgment consistent with the severity of a loss of CONTAINMENT INTEGRITY. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Required Action A.3.2.1 or A.3.2.2 is based on the provision that automatic containment isolation is available per Required Action A.3.1.

The note in the Completion Time allows multiple entries into Condition A provided the Required Actions are completed relative to the time of discovery for each affected penetration.

(continued)

O Unit Name B 3.5-6 Revision WOG-MERITS Rev. B

J Containment Integrity l

B 3.5.1 BASES ACTIONS B.1. B.2. B.3.1. and B.3.2 (continued) i Air locks are provided with two doors, each of which is j

designed to seal against the maximum containment pressurr.

resulting from the limiting DBA. Should an air lock become inoperable as a result of an inoperable air lock door or an 1

inoperable door interlock, power operation may continue provided that at least one OPERABLE air lock door is closed within the Cropletion Time of 15 minutes. The subject Completion Time is considered the shortest practical time for an operator to evaluate the situation and take the 4

Required Action.

In addition, either the inoperable j

equipment must be repaired within a Completion Time of i

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per Required Action B.2 or the OPERABLE air lock door must be closed and locked within the Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per Required Action B.3.1 and verified locked closed at least every 31 days per Required Action B.3.2..

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Times of Required Actions B.2, B.3.1 and B 3.2 are based on engineering judgment that reasonable time be allowed to perform Required Actions without prolonging the time that CONTAINMENT INTEGRITY is not ensured. The note allows access through an outer door that is used to satisfy the subject Required Actions. However, 3

k this access is only permitted for repair of inoperable air i

lock equipment. The note in the Completion Time allows multiple entries into Condition B provided the Required Actions are completed relative to the time of discovery for each affected air lock.

If the Required Actions cannot be completed in the allotted time, Condition E requires that j

the unit be placed in a MODE in which the LCO does not 1

apply.

C.1. C.2. and C.3 OPERABILITY of air locks is required to ensure that CONTAINMENT INTEGRITY is maintained. Should an air lock become inoperable for reasons other than stated in i

Condition B, the air lock leak tight integrity must be j

restored within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or actions must be taken to place the unit in a condition for which the CONTAINMENT INTEGRITY 4

LCO does not apply. Required Actions C.1 and C.2 ensure that at least one air lock door is closed and maintained l

closed until the air lock is restored to an OPERABLE status per Required Action C.3.

The Completion Time of 15 minutes per Required Action C.1 is considered the shortest practical time for an operator to evaluate the situation l

i (continued) fi Unit Name B 3.5-7 Revision WOG-MERITS Rev. B i

C:ntainment Intcgrity B 3.5.1

'/

(,,/

BASES ACTIONS and take the Required Action. The Completion Time of (continued) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment that c reasonable time period should be allowed to restore the air lock to OPERABLE status without unnecessarily prolonging the time that CONTAINMENT INTEGRITY is not ensured. The note in the Completion Time allows multiple entries into Condition C provided the Required Actions are completed relative to the time of discovery for each affected air lock.

D.d Should conditions occur other than addressed by Condition A i

through C above, that cause a loss of CONTAINMENT INTEGRITY, steps must be taken to restore CONTAINMENT INTEGRITY within the Completion Time of I hour. The Completion Time of I hour is adequate to perform minor repairs or to prepare the unit for a shutdown in accordance with Condition E.

E.1 and E.2 i

The unit must be placed in a MODE in which the LCO does not

,A apply if the respective Required Actions for a particular Condition cannot be completed within their required Completion Times. This is done by placing the unit in 1

MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to reach MODE 3 is a reasonable time, 1

based on operating experience to reach MODE 3 from full power without challenging safety systems and operators.

Similarly, the 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months to reach MODE 5 from MODE 3 is reasonable, considering that a unit can easily cooldown in such a time frame on one safety system train.

In MODE 5 CONTAINMENT INTEGRITY is not required due to the pressura and temperature limitations of MODE 5.

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS This surveillance verifies that all piping penetrations not capable of being isolated by OPERABLE automatic containment isolation devices, and required to be isolated during accident conditions, are isolated outside containment by locked closed manual valves, deactivated automatic I

isolation valves secured in their closed positions, or blind flanges. The surveillance ensures that at least (continued)

D Unit Name B 3.5-8 Revision WOG-MERITS Rev. B

Containment Integrity B 3.5.1' BASES SURVEILLANCE the outside isolation devices are capable of parforming REQUIREMENTS their safety function following a DBA. The frequency of 1

(continued) 31 days was selected based on engineering judgment l

considering the importance of verifying CONTAINMENT j

INTEGRITY.

l SL 3.5.1.2 This surveillance supplements the other surveillance described, such that the containment isolation valves l

should close in a specified time (Ref.1) following a.DBA.

The required frequency is established as part of the units inservice inspection program. SR 3.0.5 provides more information related to the inservice inspection program.

j l

i SR 3.5.1.3 This surveillance verifies that all piping penetrations not capable of. being isolated by OPERABLE automatic containment isolation devices, and required to be isolated during accident conditions, are isolated inside containment by i

I locked closed manual valves, deactivated automatic isolation valves secured in their closed positions, blind O.

flanges, or a closed system. The surveillance ensures during MODE 5 that the inside isolation devices are capable i

of performing their safety function. The intent of the l

surveillance is to verify the status of penetration l

isolation devices inside containment that were not capable of being verified in MODES 1 through 4.

l The frequency of 92 days was selected based on engineering judgment.

In practice, the surveillance would be perfonned some time close to the MODE 4 entry. Provided the surveillance is within the frequency of 92 days, multiple transitions between MODE 4 and 5 would be permitted.

E!L11dd The air lock door interlock is designed to prevent simultaneous opening of both doors in a single air lock.

Since both the inner and outer doors of an air lock are designed to withstand the maximum peak post-accident l

containment pressure, closure of either door will maintain CONTAINMENT INTEGRITY. Thus, the door interlock feature ensures that CONTAINMENT INTEGRITY is maintained while the air lock is being used for personnel entry and exit.

Periodic testing of this interlock provides assurance that (continued) l Unit Name B 3.5-9 Revision WOG-MERITS Rev. B s

Containment Integrity B 3.5.1

,m BASES SURVEILLANCE the interlock will func' tion as designed, and that REQUIREMENTS simultaneous inner and outer door opening will not (continued).

inadvertently occur. Due to the purely mechanical. nature of this interlock, a frequency of 6 months [18 months for subatmospheric containments) is considered adequate to detect degradation. As such, the frequency is based on engineering judgment SR 3.5.1.5 Periodic leak rate testing is performed in accordance with j

[the Containment Leak Rate Test Program or as required by 10 CFR 50, Appendix J (Ref. 5)]..The leak rate tests verify that the actual leakage rates from containment are less than' or equal to that assumed in the limiting containment DBA analysis..

SR 3.5.1.6. SR 3.5.1.7. and SR 3.5.1.8 Containment isolation valves that are open during unit operation are designed to close following a DBA. These surveillance ensure that upon receipt of an isolation signal the respective isolation valves close to minimize the potential leakage from containment. The frequency of Ci IB months is based on engineering judgment and considers Q

such factors as 1) the ability to perform the surveillance with the unit at power, 2) the design.and reliability of the valves, 3) the probability that a DBA will occur, and

4) the importance of CONTAINMENT INTEGRITY.

SR 3.5.1.9 4

SR 3.5.1.9 assures that for post-tensioned prestressed concrete containments the surveillance specified in the containment Structural Integrity Test Program are per-formed. This assures that a minimum acceptable level of containment structural integrity is maintained during a DBA. The requirements Dr the test program are specified-in the Administrative Controls Division.

j B

(continued)

Unit Name B 3.5-10 Revision WOG-MERITS Rev. B

)

Containment Intsgrity B 3.5.1 O

V BASES SURVEILLANCE SR 3.5.1.10. and SR 3.5.1.11 REQUIREMENTS (continued)

In subatmospheric containments there are check valves in -

certain fluid lines that must close to provide containment isolation against flow in the reverse direction when the -

line is' inactive. Because of the subatmospheric pressure in containment, the check valves are weight or spring loaded to assist closing against reverse flow.. These surveillance assure the proper operation of the valves for opening and closing functions.

y REFERENCES 1.

(Unit's controlled document containing containment isolation valve list, including those which may be opened under administrative control.]

2.

Title 10 Code of. Federal Regulations, Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants", No. 57 " Closed System Isolation Valves",1988.

3.

(Unit Name) FSAR Section [

).

4.

Title 10 Code of Federal Regulations, Part 100.11

" Determination of Exclusion Area Low Population Zone and Population Center Distance",1976.

5.

Title 10 Code of Federal Regulations, Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors",1986.

6.

52FR3788, " Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors,"

United States Nuclear Regulatory Commission, February 6, 1987.

O Unit Name B 3.5-11 Revision WOG-MERITS Rev. B

Containment Intcgrity 3.5.1 p

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the " Desk Reference" W-STS, Rev. 5, the base Standard Technical Specifications and this MERITS Specification.

The number for each written justification corresponds to a number appearing on the attached, marked up Specifications 1.7, 3.6.1.1, 3.6.1.2, 3.6.1.3, 3.6.1.4, 3.6.1.8, and 3.6.4, from the " Desk Reference" W-STS, Rev. 5.

Consistent with the practice of identifying requirements for a system or component in only one LCO, the new CONTAINMENT INTEGRITY Specification 3.5.1 has been revised to include the requirements of the STS specifications listed above. All of the specifications listed above are directly related to maintaining a leak tight containment and as such have been combined to minimize duplication and confusion as to what is the real LCO (ie. maintaining containment penetrations in a condition that will limit the release of radioactivity following a design basis accident).

In addition, the definition of CONTAINMENT INTEGRITY has been revised as a result of the changes mentioned above.

1.

The list of containment isolation valves previously found in a table of specification 3.6.4 will exist outside of technical specifications.

This will eliminate the need for license amendments every time the list is modified. Since the table has been relocated outside of

/O technical specifications, it has become necessary to modify the

(

definition of CONTAINMENT INTEGRITY to provide for the exception that previously existed on the table.

2.

The definition has been modified to include " closed system" as an appropriate isolation device. This change is consistent with the criteria given in General Design Criteria 57.

3.

The definition has been modified as a result of combining the air lock specification into the new MERITS Containment Integrity specification. The first part (a) of the STS LC0 3.6.1.3 is covered by item d of the new CONTAINMENT INTEGRITY definition and the second part (b) of the STS LCO 3.6.1.3 is covered by item e of the new C0h.AINMENT INTEGRITY definition.

In addition, Condition B and C and SR 3.5.1.4 and SR 3.5.1.5 of the new MERITS specification address the necessary actions and surveillance for meeting the requirements on air locks 4.

The requirements of specification 3.6.1. 2 are being relocated outside I

of technical specifications. As such, the reference to specification l

3.6.1.2 in the definition is no longer appropriate, and has been deleted.

In addition, SR 3.5.1.5 of the new MERITS specification requires verification of satisfactory leak rate testing. Also, specification 3.6.1.4, and identified sections of 3.6.1.3, have been l

relocated since the systems addressed and the requirements are related to satisfying leak rates, and as such will be located with the leak

,m

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WOG-MERITS Page 1 of 3 Rev. B l

)

1

Containment Integrity 3.5.1

,tQ JUSTIFICATION FOR CHANGES TO W-STS REY. 5 rate test program. However, note that P ' l ' l, and Ld Will' a

a t

be provided in the Design Features section, 4.1, " Containment".

5.

The CONTAINMENT INTEGRITY definition has been modified as a result of combining the containment ventilation system specification into the new MERITS CONTAINMENT INTEGRITY specification. The first part of the STS LCO 3.6.1.8 required that the "42-inch" shutdown purge valves be closed and sealed. The " closed" requirement is addressed by item c of the new definition and the " sealed" requirement is addressed by item e of the new definition. The second part of the STS LCO 3.6.1.8 required that only one pair of the "8-inch" mini-purge valves be open at a time and that the total time any of the valves may be opened, be limited to 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months in a calendar year. These requirements are addressed in the new definition as follows:

1)

The mini-purge valves are no different than any other containment-isolation valve and as such are covered by item a of the new definition.

2)

It is not necessary to specifically limit the time these valves may be opened since the effluent through these valves is monitored for radioactivity and the' amount of time the valves may (q

be opened is a function of the actual release If the d

radioactivity being released is within the allowable 10 CFR guidelines for continuous release, the valves could be left open indefinitely.

In addition, if the radioactivity being released exceeded allowable limits, the valves would receive a signal to close.

6.

Item e of the STS definition is deleted. The sealing mechanisms on penetrations function to maintain leakage rates within required limits and as such are required to be OPERABLE as necessary to meet item e of the new definition.

7.

The word " primary" has been deleted from the CONTAINMENT INTEGRITY specification to make the specification also apply to dual containment plants.

The ACTION statement under LCO 3.6.1.1 is now the MERITS spec Condition D.

Surveillance Requirement (SR) 4.6.1.1 (a) under LCO 3.6.1.1 is now the MERITS SR 3.5.1.1.

The footnote applicable to SR 4.6.1.1 (a) located at the bottom of the STS spec page is now the MERITS SR 3.5.1.3.

O 1

i WOG-MERITS Page 2'of 3 Rev.-B j

C ntainment Integrity' 3.5.1-i l'~)

\\

()

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 8.

Post maintenance testing activities (as in STS Surveillance 4.6.4.1) are being removed from technical specifications. Post maintenance testing ensures that equipment meets all surveillance requirements prior to restoring the equipment to an OPERABLE status. The' requirement to perform post maintenance testing.is implicit in the definition of OPERABILITY and as such does not need to be restated separately in the surveillance requirement section of any technical speci fica', jon.

~

9.

Deleted caution. Operators are trained such that they would be l

cognizant of the effects on systems from isolation of penetrations.

10. In ACTION statements for STS Specification 3.6.1.3 the statements under (a) are now in the MERITS spec as Conditions B and E.

Part B under ACTIONS of STS Specification 3.6.1.3 are represented by-Conditions C and E in the new MERITS spec..

11. SR 4.6.1.3 (a) and (b) including the applicable footnote at the bottom i

of the page are now in the new MERITS SR 3.5.1.5.

SR 4.s.l.3 (c) is

{

covered by the new MERITS SR 3.5.1.4.

l

12. In LCO 3.6.4 the ACTION statements (a), (b), and (c) are incorporated in the MERITS specification Condition A.

Part'(d) is Condition E in j

the MERITS spec.

13. STS SR 4.6.4.2 (a) is now found in the MERITS specification as STS SR 3.5.1.6.

STS SR 4.6.4.2 (b) is now MERITS SR 3.5.1.7.

STS SR 4.6.4.2 (c) is now MERITS SR 3.5.1.8.

STS SR 4.6.4.3 is now MERITS SR 3.5.1.2.

14. SR 3.5.1.9 assures that for post-tensioned prestressed concrete containments the surveillance specified in the containment structural integrity test program are performed. This assures that a minimum acceptable level of containment structural integrity is maintained during a DBA. The test program itself is specified in Administrative Controls.

SR 3.5.1.10 and SR 3.5.1.11 are check valve surveillance in subatmospheric containments where there are check valves in certain fluid lines that must close to provide containment isolation against flow in the reverse direction when the line is inactive. Because of

)

l the subatmospheric pressure in containment, the check valves are l

weight or spring loaded to assist closing against reverse flow. These j

surveillance assure the proper operation of the valves for opening l

' and closing functions.

8

/N i

WOG-MERITS Page 3 of 3 Rev. B I

i I

5 R e.v. e k,

DEFINITIONS k

CONTAINMENT INTEGRITY 1.7 CONTAIMENT INTEGRify shall exist when:

I All penetrations required to be closed during accident conditions a.

are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, er 2)

Closed by manual valves, biind flanges, or deactivate valees secured in their closed gettions.px.;'. :. ;d automatic n t: ' u or a, closed s em3 3) cn unkt admini ve Senycl on sejeded End ficns,.

b.

All og paent tches are c os n.

n...

Bei ocr4 e an I41

'.:.L: air lock ':

g " n cl ^

kan :c:leses excej+ 2.-

--, ': :-^

c' S::' ; ;

3, for erdry and exit.

I The containment leakage rates are within "We.elr rt(Ulr4K

' O.:.'. :, ;..:

~

t: ;f '-

'f'.;^.'.7 h h'

.M _ Rhd_dcwn _ furge _VMyf(S, af=A Clo}iil *J,__,,,_

a,:-- 2 a::,:__ r,: ;;;. w ; -

- t

.,.,.....~..,

(

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 within the reactor pressure vessel with the vessel head remov the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

DO5E E0V! VALENT I 131 which alone would produce the same thyroid dose as the q aisture of I-131, I-132,1-ul,1-D4, and 1-135 actually present.

dose conversion factors used for this calculation shall be these listed inThe thyroid

Table !!! of TID-14444, " Calculation of Distance Factors for power and Test October 1177].deactor $ttas" or Table E-7 of MRC Regulatory Guide 1.10g, Revisio I-AVERAGEDis!NTEGRATIONENERGY 1.11 I shall be the average (weighted in proportion to the concentration of each radionuclides in the saeple) of the sua of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

COMANCHE PEAK - UNIT 1 12 1

\\

l

R e.v. B 3/4.6 CONTAINMENT SYSTEMS

\\

3/4. C.1 44WN CONTAINMENT

(

CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION O

3.6.1.1 * '

CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Without ; '

y CONTA! MENT INTEGRITY, restore CONTAINMENT INTEGRITY w 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

4. 6.1.1

8 : ;' CONTAINMENT INTEGRITY shall be demonstrated:

At least once per 31 days by verifying that all penetrations

not a.

capable of being closed by CPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flan secured in their positions, =ges. or deactivated automatic valves r;^. n p r vid:: '- Td': 3.0 0 ;f

!; rifir tin 0.0.0.;,

5.

S; x-' *y' ; 'h-t t r' et-te'- :-t e'- '-9 ': '

---1't

'it @

"- -- :'rr rt: ? 5;r'"rr f.5.1.?; rf

- r:t -?r'r; c' r$ ;=:ttt' 2.',rt t: T;;; ! trt't;,

=r;; it: r t:f rrt :'- teh, :;;rd ":1?r' ; : ?;;; * - !

' r t, 5; ' n' r it te t' ; : t r?

'tt ;r et :

th=.",r, l50 ;;'y:,:rf cr'*;;f r- '"t er '" e;rn::n et !=:

rrf

't" :- - *-

'r

r:!:.

ff:f ^2 at:.r '::: n't it:-'r! ;r:=n'. 'a

';rifint'n ".'.1. f. 'n -" :thr T;;; 5 rf 0 ::=tr:--

. u _ _m e..

. u. _.

4...

.m..

. en,

Except valves, blind flanges, and deactivated evtomatic valves which are located inside the containment and are locked, sealed or otherwise secured

/2 in the closed position.

W each COLD $HUTDOWN except that such verification need not be p often than once per 92 days.

%MERT @

6 COMANCHE PEAK - UNIT 1 3/4 6-1A o

$cv 0 i

i i

INSERT:

i SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY i

Verify Containment Structural Integrity As specified f

in accordance with the unit containment by the unit structural integrity test program.

containment structural integrity test program.

l l

Cycle each weight or spring loaded check

[92 days) valve testable during plant operation, through one complete cycle of full-travel and verify that each check valve remains closed when the differential pressure in the direction of flow is s [

] psid and l

opens when the differential pressure in l

the direction of flow is 1 [

] psid and < [

] psid.

.C Cycle each weight or spring loaded check

[18 months]

valve not testable during plant operation, through one complete cycle of full travel and verify that each check valve remains closed when the differential pressure in the direction of flow is s [

] psid and opens when the differential pressure in the direction of flow is 1 [

] psid and < [

] psid.

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COMANCHE PEAK - UNIT 1 3/4 6-4A f

ktM O CONTAINMENT SYSTEMS

[(

CONTAINMENT AIR LOCKS

(

LIMITING CONDITION FOR OPERATION

3. 6.1. 3 Each containment air lock shall be OPERABLE with:

ha.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed,.and

@b.

An overall air lock leakage rate of less than or equal to 0.05 L, at P,,[50)psig.

APPLICABILITY: ICDES 1, 2, 3, and 4.

ACTION:

a.

With one containment air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed; O

g 2.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days; 3.

Otherwise, be in at least NOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l

t and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and 4.

The provisions of Specificatib'~n 3.0.4 are not applicable.

b.

With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HDT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O e

COMANCHE PEAK - UNIT 1 3/4 6-5A l

Rn.B CONTAINMENT SYSTEMS

@fO s*

SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is a.

being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying seal leakage is less than [0.01) L as detersined by precision flow measurements when measured for at least,(30] seconds with the volume between the seals at a constant pressure of [50 psig];

b.

b By conducting overall air lock leakage tests at not less than P,,

[$0 psig), and verifying the overall air lock leakage rate is within its Itait:

1)

At least once per 6 months,* and 2)

Prior to establishing CONTAlmiENT INTEGRITY when maintenance has been performed on the air lock that could affect the air Iock sealing capability."*

At least once per 6 months by verifying that only one door in each c.

air lock can be opened at a time.

5 s

1 "The provisions of Specification 4.0.2 are not applicable.

- This represents an exemption to Appendix J, paragraph 111.0.2 of 10 CFR Part 50.

[ Applicant must request this exemption.)

I

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i Ro.8 CONTAINMENT SYSTEMS 3/4.6.4 CONTAINMENT ISOLATION VALVES k

j LIMITING CONDITION FOR OPEAATION D

3.6.4 The containment isolation valves

,---^d *- te'- ? *-1 shall be OPERABLE

't' 'r'"':- t'r: : e t - '- 'e'- ? '- L APPLICABILITY: MODES 1, 2, 2, and 4.

f.CTION:

"With one or more of the isolation valve (s) :; r f ' ' d ': L 5-1 inoperable, maintain at least one isolation valve CPERA8LE in each affected penetration that is open and:

Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , a.

ar b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of 4t least c.

one closed manual valva or blind flange, or d.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS O

' f.#.2 'M T--ht'"2 CIC: - M ^f

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^f Mt M:rd' ;'y.

COMANCHE PEAK - UNIT 1 3/4 6-27A t

Rsv. 9 CONTAINMENT SYSTEMS O

SURVEILLANCE REQUIREMENTS (Continued)

\\

D 4.6.4.2 Each isolation valve ::::f'f f '- ?d': ?.5-! shall be demonstrated OPEUSLE during the COLD SHUTDOWN or REFUELING MODE at least once per is sonths by:

Verifying that on a Phase "A" Isolation test signal, each a.

o Phase "A" isolation valve actuatas to its isolation position; ft/

b.

Verifying that on a Phase 'S' Isolation test signal, each Phase

'8" isolation valve actuates to its isolation position; and c.

Verifying that en a Containment Purge and Exhaust Isolation test signal, each purge and exhaust valve actuates to its isolation position.

4.6.4.3 The isolation time of each power-operated or autematic valve +8 Td': ? 5-1 shall be determined to be within its Ifnit when tested pursuant tt Specification 4.0.5.

l 0

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UNIT 1 3/4 g.2SA 0

I

Purge and Exhaust Isolaticn Radiation Instrumentation 3.5.2 i

,x 3.5 CONTAINMENT SYSTEMS 3.5.2 Purce and Exhaust 1 solation' Radiation Instrumentation (Atmospheric, Dual, Ice Condenser)

LCO 3.5.2 Purge and exhaust isolation radiation instrumentation shall be OPERABLE consisting of the following:

a.

[

] particulate radiation channels with trip setpoints s[

] yCi/cc,

b..[

) gaseous radiation channels with trip setpoints 5[

] pC1/cc and c.

[

] containment atmosphere radioactivity-high channels with trip setpoints 1 [ ] #C1/cc.

APPLICABILITY:

MODES I, 2, 3, and 4.

......................------N0TE-----------------------------

LCO 3.0.3 is not applicable.

t ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required A.I Adjust setpoints to

NOTE-----

purge and exhaust within limits.

Completion isolation radiation Times are on channel setpoints a per channel l

outside limits.

basis 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued) a I

\\

l i

Unit Name 3.5-8 Amendment WOG-MERITS Rev. B 1

]

j 1

Purge and Exhaust Isolation Radiation Instrumentation 3.5.2 ACTIONS (continuad)

{N CONDITION REQUIRED ACTION COMPLETION TIME B.

One or more required B.1 Isolate each purge and I hour purge and exhaust exhaust penetration with isolation radiation a locked closed manual channels inoperable for valve, a deactivated reasons other than automatic valve secured Condition A.

in the closed position, or a blind flange.

E i

Required Action of Condition A not met l

within required l

Completion Time.

l SURVEILLANCE REQUIREMENTS l

l SURVEILLANCE FREQUENCY SR 3.5.2.1 Perform CHANNEL CHECK of each purge and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

exhaust isolation radiation channel.

SR 3.5.2.2 Perform ANALOG CHANNEL OPERATIONAL TEST 31 days for each purge and exhaust isolation radiation channel.

SR 3.S.2.3 Perform CHANHEL CALIBRATION of each purge 18 months and exhaust isolation radiation channel.

l l

\\ [

\\

Unit Name 3.5-9 Amendment WOG-MERITS Rev. B 9

Purge and Exhaust Isolation Radiation Instrumentation 3.5.2 CROSS-REFERENCES TITLE NUMBER Engineered Safety Features Actettion System 3.2.2 Instrumentation - Function 3.C. Containment Purge Isolation

[RCS Leakage Detection Instrumentation)

[3.3.16]

Containment Integrity 3.5.1 1

1 i

i I

i i

\\

l Unit Name 3.5-10 Amendment WOG-MERITS Rev. B

1 1

1 Purge and Exhaust Isolaticn Radiati n Instrumentation 3.5.2 JUSTIFICATION FOR CHANGES TO W-STS REY. 5 The following justifications.are provided to explain the reasons for the differences between the. " Desk Reference" W-STS, Rev. 5, the base Standard Technical Specification used for MERITS, and this MERITS Specification. The i

number for each written justification corresponds to a number appearing on the attached, marked up Specification 3.3.3.1, " Radiation Monitoring for Plant Operations", from " Desk Reference" W STS Rev. 5.

1.

The title was changed to address only the purge and exhaust isolation radiation monitoring instrumentation.

.2.

The purge and exhaust isolation radiation monitoring instrumentation was j

identified per the title.

3.

The MODE applicability was changed'to be consistent with LCO 3.5.1, Containment Integrity. CONTAINMENT INTEGRITY must be maintained in MODES 1, 2, 3 and 4, when a DBA could cause a substantial increase in fission product radioactivity in containment.

In MODES 5 or 6, CONTAINMENT INTEGRITY is not required'because of the pressure and temperature limitations in these MODES. The purge and exhaust isolation i

radiation instrumentation requirements for MODE 6, during CORE j

ALTERATIONS or movement of irradiated fuel within containment, are l

(

provided in LCO 3.8.4, Containment Building Penetrations.

8 l

O

(

4.

The crossed out portions are not applicable to the purge and exhaust

\\

isolation radiation monitoring instrumentation.

5.

The letter designations were changed to their respective times for surveillance intervals.

6.

The reference to former TS 3/4.11.2.1, Gaseous Effluents: Dose Rate, is not applicable. The programmatic controls have been relocated to Administrative Controls, and the procedural details are relocated to the Offsite Dose Calculation Manual.

1 WOG-MERITS Page 1 of 1 Rev. B t

1

AC Sources - Operating 3.7.1

(")/

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the

Desk Reference" W-STS Rev. 5, the base Standard Technical Specification used for MERITS. The number for each written justification corresponds to a number appearing on the attached, marked-up Specification 3.8.1.1, AC Sources - Operating.

1.

All Limiting Condition for Operation requirements are the minimum acceptable requirements. This is understood and need not be stated explicitly.

2.

The Diesel Generators are " independent" and therefore separate.

3.

This requirement has been moved to the Surveillance Requirements (SR) section and is now SR 3.7.1.2.

4.

This requirement has been moved to the SR section and is now SR 3.7.1.3.

5.

The fuel transfer pump requirements have been moved to the SR section and

-)

are now SR 3.7.1.4.

6.

These requirements have been deleted as the availability of the proper i

f level of lubricating oil in the diesel generators, and an on-hand inventory I

(

of replacement lubricating oil with an available means of refilling or topping off the dies'el generator lubricating oil level are items that are part of an effective diesel generator maintenance program. These items are generic to diesel generators and are not dependent on plant configuration or diesel generator manufacturer. The lubricating oil is not rapidly consumed during diesel generator operation, therefore, the replacement lubricating oil inventory is required only for refill purposes and not for normal diesel generator start operations.

Presently licensed plants, such as Farley, Summer, Shearon Harris, MC Guire, Catawba, Callaway, Wolf Creek, North Anna, and Vogtle, do not include lubricating oil storage or transfer considerations as LCO statements.

B 7.

With an inoperable offsite circuit or diesel generator it is prudent to demonstrate diesel generator operability by verifying the ability of the i

diesel generators to start and achieve the required voltage and frequency if not recently verified by the required surveillance. However, it is not desirable to synchronize and parallel load the diesel generators to the buses, since a failure of the offsite circuit powering the buses can lead to a complete loss of AC power as addressed in Information Notice 84-69.

This Information Notice warns against operating diesel generators tied to offsite power when~the units AC power sources are abnormally degraded or threatened.

g 1

WOG-MERITS page 1 of 5 Rev. B L-

i AC Sources - Operating 3.7.1 l

I j'

JUSTIFICATION FOR (HANGES TO W-STS REV. 5 l

(continued) j 8.

A new action has been added to address the condition where one Diesel Generator and one offsite circuit are_ inoperable at the same time. This condition requires SR 4.8.1.1.1.a to be performed within I hour and once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. The same requirements identified for the loss of one Diesel Generuor are also included. The inoperable offsite source or the inoperable Diesel Generator must be returned to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This complies with Regulatory Guide 1.93, Rev. O, December, 1974.

l l

9, Staggered testing was initially instituted to determine reliability of component / system being tested but it has not been conclusively demonstrated that staggered testing has had an impact on reliability. The d

I staggered testing approach is inconsistent with the requirements of Table 4.8-1 as applied to' individual diesel generators or auxiliary equipment.

Increased testing required in accordance with the table for an individual item interrupts the staggered test approach and places the i

involved item on a shorter test frequency.

In addition, the test frequencies of the table are sufficiently short to accomplish the objective of ensuring individual item OPERABILITY and for obtaining reliability data.

J f

10. The Surveillance Requirement is expressed as voltage and frequency for

\\

verification of acceptable surveillance completion. Stating of the requirement in this manner is consistent with the MERITS program objectives and writers guide to express requirements in terms of units and parameters available to the operator. Con)pliance with the frequency requirement correlates directly to achievement of the presently specified rpm. This approach is generic to all plants since the parameters are not specific to either plant configuration or to diesel generator manufacturer.

g

11. The requirement to specify the method by which the Diesel Generator is l

started has been deleted. This is an item for inclusion in individual i

plant operating procedures.

12. A new SR has been added to verify the pressure in the air start receivers at the frequency in Table 3.7.1-1.

This is necessary to ensure that the diesels are capable of starting when required. See SR 3.7.1.8.

13. The specific procedure and standards requirements for fuel oil sample testing are unit specific and not common to all units. These requirements have been moved to the Bases section and are to be contained in unit operating procedures under administrative control. Requirements to verify i

new fuel oil and storage tank fuel oil properties are covered by I

SR 3.7.1.10, SR 3.7.1.11, and SR 3.7.1.13.

i j

WOG MERITS page 2 of 5 Rev. B l

i

AC Sources - Operating 3.7.1

(

e JUSTIFICATION FOR CHANGES TO W-STS REV. 5 (continued)

14. Maximum diesel generator loading for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test must be based upon manufacturers recommendations. The inserted information,110% of continuous rating, is consistent with the present Callaway Unit requirements. The presently bracketed information, 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months rating, is inappropriate since this is the load rating at which the diesel generator would require overhaul after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of operation. Diesel generator run time for rating at load is cumulative and if not properly applied could require premature diesel generator overhaul.

15. The surveillance has been removed from technical specifications since the ability of the diesel generator to start without being inhibited by an invalid diesel generator protection feature, such as turning gear engaged or emergency stop, is demonstrated at least every 31 days in accordance with the frequency prescribed by Table 4.8.1.1.2-1 (MERITS Table 3.7.1-1).

The diesel turning gear engaged and emergency stop features are for equipment protection purposes and are properly items for check during diesel generator maintenance actions. This is analagous to the overcurrent devices on pump motor supply breakers. These items are not technical specification items or requirements. The frequency of diesel generator verification and demonstration of the ability of the diesel generator to start without being inhibited by invalid protective feature action is i

generic to all plants.

B

16. Removed from tech specs since it is a duplication of ASME Code requirements.

17. Reports relocated to Administrative Controls, section 5.9.6.

18. Surveillance requirements for Loss of Voltage and Degraded Voltage Trip Devices were moved from the Instrumentation Chapter and incorporated into AC Sources - Operating. These devices and their operation directly relate to the operational capabilities of the' AC Sources. The Loss of Voltage and Degraded Voltage surveillance are: TRIP ACTUATING DEVICE OPERATIONAL TEST (SR 3.7.1.15 and SR 3.7.1.16); CHANNEL CALIBRATION (SR 3.7.1.23 and SR 3.7.1.24): and ENGINEERED SAFETY FEATURE RESPONSE TIME test (SR 3.7.1.35 and SR 3.7.1.36), respectively. The required frequencies have been established consistent with the WCAP 10272 Supplement 2, February,1986, (TOPS) study.

19. Surveillance requirements for the Load Shedding and Sequencing Timer were l

moved from the Instrumentation Chapter and incorporated into AC Sources -

Operating. This device directly relates to the operational capability of the AC Sources. The incorporated surveillance are: ACTUATION LOGIC TEST (SR 3.7.1.12) and ENGINEERED SAFETY FEATURE RESPONSE TIME test (SR 3.7.1.34), respectively. The required frequencies have been established consistent with WCAP 10272 Supplement 2, February 1986, (TOPS) study.

WOG-MERITS page 3 of 5 Rev. B i

AC Sources - Operating 3.7.1 JUSTIFICATION FOR CHANGES TO W-STS REY. 5 (continued)

20. The test frequency as defined in MERITS LCO 3.7.1, Table 3.7.1-1, is on a per diesel generator basis, incorporating both the number of failures in the last 20 as well as the last 100 starts. As such the number of failures in the last 100 starts has been adjusted to 5 5 and 16 to be at 5% of 100 consistent with the 11 and 12 as stated for the number of failures in the last 20 starts. The frequency of diesel generator starts is determined by the number of failures in the last 20 or 100 tests and is neither plant nor diesel generator manufacturer specific and as such the stated requirements in the MERITS LCO are generic to all plants.

In addition, the Diesel Generator Test Schedule," Number of Failures in Last 100 Hours", was changed to 5 5 and 16 for the Diablo Canyon Units 1 and 2 and approved as Licensing Amendments 14 and 15, Docket Numbers 50-323 and 50-275, based on the same basic justification as stated above.

3

21. A sentence has been added to clearly state that the failure of the fuel oil, auxiliary equipment, or Engineered Safety Features Actuation System requirements does not constitute a valid failure of the diesel generator.

This is consistent with the intent of the technical specification and is added only to provide clarity.

O

22. The requirements to verify the fuel oil level in the storage tank and that the fuel transfer pump starts and transfers fuel oil to the day and engine-mounted tank has been established as 31 days. This frequency is considered adequate to ensure the availability of the necessary volume of fuel oil in the storage tanks. When the diesel generators are operated on the maximum 7 day test frequency, with a run time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , fuel usage would be a small percentage (aboJt 2%) of the volume available in the fuel oil storage tanks. Common plant practice is to maintain the fuel oil storage tanks well above the minimum required volume, therefore during any 31 day cycle more than an adequate volume of fuel oil would be available for an extended diesel generator run time. The ability of the fuel oil transfer pump to start and to transfer fuel oil from the storage system to the day tank is verifiet' by maintenance of the minimum required volume of fuel oil in the day tank. Day tank fuel oil volume is verified by surveillance, SR 3.7.1.2, which is performed on the same frequency as the I

diesel generator starts.

The fuel oil storage tank volume is sized for each plant consistent with the diesel generator maximum run time requirements and the diesel generator fuel oil consumption rates. The fuel oil transfer pumps have the identical functional requirements for every plant. Therefore, these surveillance requirements are not plant specific and are generic to all plants.

s O

WOG-MERITS page 4 of 5 Rev. B

.AC Sources - Operating 3.7.1 r

JUSTIFICATION FOR CHANGES TO W-STS REV. 5 (continued)

23. The requirement for verifying that the fuel oil transfer pump transfers fuel oil from the fuel oil storage. tank to the day tank of each diesel generator is performed every 31 days as specified in SR 3.7.1.4.

The fuel oil cross-connection lines are not part of the primary success path for providing fuel oil to the diesel. generator day tanks as are the direct lines between the fuel oil storage tanks and the associated diesel generator day tanks. This is generic to all plants since the minimum required fuel cil storage tank volume is established to support the diesel generator extended run requirements.

p O

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WOG MERITS page 5 of 5 Rev. B

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AC Sources -. Shutdown 3.7.2 JUSTIFICATION FOR CHANGES TO'W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the " Desk Reference" W-STS Rev. 5, the base Standard Technical Specification used for MERITSc The number for each written justification corresponds to a number appearing on the attached, marked-up Specification 3.8.1.2, AC Sources - Shutdown..

{

1.

All Limiting Conditions for Operation (LCO) requirements are the minimum,

~

acceptable requirements. This is understood and need not be stated explicitly.

2.

This requirement has been moved to the Surveillance Requirements-(SR) '

'l section and is now SR 3.7.1.2.

3. -This requirement-has been moved to the SR section and is now SR'3.7.1.3.

^

The fuel transfer pump requirements have been moved to the SR section and 4.

are now SR 3.7.1.4.

5.

These requirements have been deleted as the availability of.the proper.

level of lubricating oil in the diesel generators, and an on-hand inventory of replacement 1ubricating oil with an available means of refilling or O

j topping off the diesel generator lubricating oil level are items that are part. of an effective diesel generator maintenance program. These items are generic to diesel generators and are not dependent on plant configuration or diesel generator manufacturer. The lubricating oil is not rapidly i

consumed during diesel generator operation, therefore, the replacement

)

lubricating oil inventory is required only for refill purposes and not for normal diesel generator start operations. Presently licensed plants, such as Farley, Summer, Shearon Harris, MC Guire, Catawba.Callaway, Wolf Creek, North Anna, and Vogtle, do not include lubricating oil storage or transfer 3

i considerations as LCO statements.-

6.

Requirement added to extend the APPLICABILITY to include, during movement of irradiated fuel with no fuel assemblies in the reactor vessel and during crane operation over irradiated fuel assemblies with no fuel assemblies in-the reactor vessel. This addition covers operations when the unit is not-in a defined MODE., irradiated fuel not in the reactor vessel, and for l

operations over the spent fuel storage pool. Under.these operational-conditions, with irradiated fuel present, a potential for an accident involving fuel handling exists and proper precautions taken.

l 1

WOG MERITS page 1 of 2 Rev. B

AC Sources --Shutdown 3.7.2

,n JUSTIFICATION FOR CHANGES TO W-STS REV. 5 (continued) 7.

An additional requirement has been added to, within 15 minutes, suspend any operation which may result in draining the reactor vessel. With no safety-grade power available, this action is warranted to prevent inadvertent uncovering the fuel assemblies.

8.

The requirement for depressurizing and venting the reactor coolant. system was previously covered by LCO 3.4.9.3, Overpressure Protection Systems, and is now covered in MERITS by LCO 3.3.18, Cold Overpressure Prevention. The MERITS LCO provides requirements for venting, including a [ ] square inch vent requirement. The requirement was deleted to eliminate redundant requirements.

9.

The requirement to restore the required AC power as soon as possible has been expanded to include all of MODES 5 and 6, not just with low water level in MODE 6 or with the Reactor Coolant System loops not filled in MODE

5..The requirement to restore required sources has been rephrased to initiate, within 15 minutes, and continue action until'the required AC l

electrical power sources are restored.

10. The surveillance for simultaneous start of the Diesel Generators every 10 years was deleted since this surveillance is adequately covered by LC0 g'~'g 3.7.1, AC Sources - Operating and must be performed, if applicable, prior i

(

j to entry into MODE 4.

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l WOG-MERITS page 2 of 2 Rev. B 1

C ntainment Building Penetrations

-3.8.4 r%

(

3.8 REFUELING OPERATIONS 3.8.4 Containment Buildina Penetrations LCO 3.8.4 The containment building penetrations shall be in the following status:

a.

The equipment hatch closed and held in place by four

bolts, b.

One door in each airlock closed, and Each penetration providing direct access from the c.

containment atmosphere to the outside atmosphere either:

1.

Closed by an isolation valve, blind flange, manual valve, or equivalent, or 2.

Capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System with:

a.

[n] particulate radiation channels with trip setpoints s [

] #Ci/cc, and

/^

b.

[n] gaseous radiation channels with trip setpoints s [

] #Ci/cc.

(

c.

[n] containment atmosphere radioactivity-high channels with trip setpoints s [

] pCi/cc.

APPLICABILITY:

MODE 6 during CORE ALTERATIONS or movement of irradiated fuel within containment.

ACTIONS CONDITION REQUIRED ACTION-COMPLETION TIME A.

One or more required A.1 Adjust setpoints to

NOTE-----

purge and exhaust within limits.

Completion isolation radiation Time is on a channel setpoints per channel outside limits, basis 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (continued) iO Unit Name 3.8-5 Amendment WOG-MERITS key. B

1 Containment Building Penetrations 3.8.4

)

ACTIONS (continued)

CONDITION REQUIRED ACTION.

COMPLETION TIME B.

One or more containment B.1 Place affected 15 minutes building penetrations containment i

not in required status.

penetrations in required status.

i E

E I

R One or more required purge and exhaust B.2.1 Suspend CORE 15 minutes isolation radiation ALTERATIONS.

channels inoperable for reasons other than 3D Condition A.

B.2.2

NOTE----------

E Suspension of irradiated fuel Required Action of movement shall not Condition A not met preclude completion of within required movement to a safe Completion Time.

conservative position.

Suspend movement of 15 minutes irradiated fuel within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Perform CHANNEL CHECK of each purge and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months exhaust isolation radiation channel.

SR 3.8.4.2 Verify each containment building 7 days 1

l penetration in required status.

j l

(continued) l

/~

Unit Name 3.8-6 Amendment WOG MERITS Rev. B i

I Containment Building Penetrations 3.8.4

.b SURVEILLANCE FREQUENCY SR 3.8.4.3 Verify containment purge and exhaust 7 days isolation occurs on:

a.

Manual initiation, and b.

High radiation test signal from each purge and exhaust isolation radiation channel.

SR 3.8.4.4 Perform ANALOG CHANNEL OPERATIONAL TEST 31 days for each purge and exhaust isolation radiation channel.

i SR 3.B.4.5 Perform CHANNEL CALIBRATION of each purge 18 months and exhaust isolation radiation channel.

O CROSS-REFERENCES - None.

I O

Unit Name 3.8 7 Amendment WOG MERITS Rev. B f

]

N'e v, B REFUELING OPERATIONS 3 /a. 9. a CONTAINMENT BUILDING PENETRATIONS l

LIMITING CONDITION FOR OPERATION i

3.9.4 The containment building penetrations shall be in the following status:

l w

i The equipment eow closed and held in place by e ='-4-"-

ef four

{

a.

bolts.

N b.

:! 'r r Of one door in each airlock is closed, and Each penetration providing direct access from the containment c.

atmosphere to the outside atmosphere shall be either:

oc e4vWde.d. {

1)

Closed by an isolation valve, blind flange, e manual valve,gor 2)

Se capable of being closed by an OPERABLE e*6 emet 4e gentainment 3

purge and gzhaust [ solation we4*e.Sykm.

McDE (o APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within g

t+.e containment.

ACTION:

C 1

k:kn 65 m n.

i With the requirements of the above specification not satisfied, ' :df:*-?(, 45 y

^

\\

suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. p3 Pau, nc q,ctrJ pc *dian in ik ugared 'Jidus in IC msnVt2S.

5 SURVEILLANCE REQUIREMENTS 7

a.9.4 Each of the above recuired containment building penetrations shall be determined to be either in its closed / isolated condition'or capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve tith'- 100 S: r; prier t: th: ;t:rt ;f r.: : ';;;; once per 7 days during 3

CORE ALTERATIONS or movement of irradiaterj fuel in the containment building by:

Verifying the penetrations are in their closed / isolated condition, a.

er b.

Testing the containment purge and exhaust isolation valves per the i

applicable portions of Specification 4.6.4.2.

O CDMANCHE PEAK - UNIT 1 3/4 9-4 l

l

i AFV, B aEruELING OPERATIONS I

3/a.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION $YSTEM l

)

LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust Isolation System shall be OPERABLE.

40E 6 APPLICABILITY: 4 uring CORE ALTERATIONS or movement of irradiated fuel within 3

0 4

we containment.

l

)

ACTION:

l With the Containment Purge and Exhaust Isolation System inoperable, a.

close each cf the purge and exhaust penetrations providing direct,

7 access from the containment atmosphere to the outsiae atmospheres wi%

5 mi no D.

: ;-. f i:.: ef 5 ::f' fret'e : 3.0.? :-

2.0.' : : et ---!!: e ?:.s g

6

%e CME ALf4DetencMS uct woe.med of irrn24adel het w:%)n testitin Q wi%.n (5 minales 5

i SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Purge and Exhaust Isolation System shall be demonstrated OPERABLE 9 t'3-200 ' u

-':r a n e ster -' r f :t ?:::t once per 7 days 9

during CORE ALTERATIONS by verifying that* containment purge and exhaust isolation occurs on manual initiation and on a High Radiation test signal from each of the containment radiation sonitoring instrumentation channels.

COMNCHE PEAK - UNIT 1 3/4 9-10 u________------_---

h a, 8 JN$TRUMENTATION 3/a.3.3 mon 170RINO INSTRUMENTATION RADIATION #cNITORINO FOR PLANT OPERATIONS LIMITINO CONDITION FOR OPERAT10N 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be DPERABLE with their Alarm / Trip 5etpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

With a radiation monitoring channel Alarm / Trip 5etpoint for plant a.

operations exceeding the value shown in Table 3.3 6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable, b.

With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.

g

't; ; ;_ ;';n; ;' I;;;!;;^..;.; 2.0.0.;;

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O SURVE!LLANCE REQUIREMENTS 4.3.3.3 Each radiation monitoring instrumentation channel for plant operations shall be comonstrated OPERABLE by the performance of the CMANNEL CHECK, CNANNEL CAlltRATION and ANALOG CHANNEL OPERATIONAL TEST for the ICDES and frequencies shown in Table 4.3-3.

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_ TABLE 3.3 6 (Continued)

_ TABLE NOTATION $

Must satisfy

.ation 3.11.2.1 re ts.

With feradiated f the fuel stor

.I areas.

i th fuel in the fuel storage pool areas.

ACTION $TATEMEMTS AC710N 26 -

With less than the Minfeum Channels CPERABLE requirement exhaust valves are saintained closed. operation may continue provided the co CTION 27

Channels OPERABLE reoufrement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate theW h

t of the Control Room Emergency ventilation System in tContro stion recirculation node.

ACTION 28 -

th less than the Minimum Channels OPERABLE re tio may continue for up to 30 days provided irement, opera-portad continuous monitor with the same n appropriate providea the fuel storage pool area.

are Setpoint is monitors to 'ERABLE status within 3 estore th'e inoperable operations iny wing fuel moves >ent sys or suspend all areas.

n the fuel storage pool ACTJON 29 -

With less than the Min t

comply with the ACTION r annels OPERABLE requirement,

.f raments of specification 3.4.6.1.

ACTION 30 -

With the number of ERABLE.cha. els less than required by the Minimum Channels PERA8LE (Ircuire 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , d of monitoring the nts, initiate the preplannet alternate set opriate parameter (s), withir 1)

. her restore the inoperable Channel (s status within 7 cays of the event, or e OPERAELE j

2) prepare and submit a special Report to the Cossi pursuant to Specification 6.9.2 within 14 days foi fon the event outlining the action taken, the cause of t ing the system to OPERABLE status,inoperability and the plans and schedul I

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RHR - High Water Level 3.8.5 JUSTIFICATION FOR CHANGES TO W-STS REV. 5 The following justifications are provided to explain the reasons for the differences between the Desk Reference W-STS Rev. 5, the base Standard Technical Specification, and this MERITS Specification. The number for each written justification corresponds to a number appearing on the attached, marked-up Specification 3.9.8.1, Residual Heat Removal And Coolant Circulation

- High Water Level.

1.

The note is now a part of the LCO statement. The reference to ' CORE ALTERATIONS in the vicinity of the reactor vessel hot legs" has been removed because of the change in the definition of CORE ALTERATIONS to cover only certain operations.

The time period for removal of the required RHR from operation was changed to *1 hour per 2-hour period" consistent with the note in the Shearon Harris Plant Technical Specifications. This change is consistent with the availability of adequate heat sink and cooling with 2 23 feet of water above the reactor vessel flange. The change is not plant specific. All Westinghouse PWR plants have a large volume of water available as a heat sink for natural convection cooling when the refueling cavity is filled to 2 23 feet above the reactor vessel flange.

This heat sink can provide sufficient decay h2at removal to ensure the O

temperature limit of MODE 6 is not exceeded during the 1-hour period that the RHR system is not in operation, provided the RCS temperature at the start of the period is low enough to accommodate the expected temperature increase while RHR is :ecured.

Initially, after shutdown, the temperature limit of MODE 6 may restrict the removal of RHR from service. After the decay heat load is reduced sufficiently, operation of the required RHR for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 2-hour period with 2 23 feet of water above the reactor vessel flange will provide adequate cooling for RCS temperature control and adequate mixing to prevent boron and thermal stratification.

B 2.

The APPLICABILITY statement was reworded consistent with the MERITS format. No change was made to the APPLICABILITY.

3.

The Action Statement now addresses no RRR loop OPERABLE or in operation.

The required actions are the same, except that the action for suspending operation involving an increase in reactor decay heat load or a reduction in boron concentration of the RCS and to initiate corrective action to return one loop to OPERABLE and operating status are to be initiated within 15 minutes. The term "immediately* is not a defined term and the time requirement to suspend operations was taken to mean immediately and were defined as 15 minutes. The requirement to have one RHR loop OPERABLE and in operation has been clearly stated by replacing the words "the required" with the word 'one'.

4.

An additional action has been added to continue actions to restore the RHR loop to OPERABLE status. This Action meets the intent of this LCO, i.e.,

O

'to complete the actions started in Required Action A.3.1.

The requirement to ' complete" the action is now specified.

WOG MERITS page 1 of 2 Rev. B j

RHR - High Water Level 3.8.5 JUSTIFICATION FOR CriANGES TO W-STS REV. 5 5.

The requirement to close containment penetrations within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months was deleted. The requirements of NRC Generic letter 88-17, October 17, 1988, identifies requirements to be in-place for closure of containment penetrations in administrative procedures and controls. These requirements were specified for the more severe concerns of low water level. The same procedures should include the high water level condition concerns to assure and maintain consistency in operating actions.

i 6.

The identification of a specific RHR loop flow rate in the surveillance was deleted. This maintains consistency between the RHR MODES 5 and 6 technical specifications. Flow rate requirements are expected to be placed in plant operating procedures in order to adequately and safely address RHR flow rate. The proper flow rate depends upon several factors, such as, time since shutdown for decay heat, reactor coolant temperature rise through the core, prevention of thermal and boron stratification, and protection of the RHR pumps from conditions, at low water levels, which could cause vortexing conditions and pump cavitation.

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A WOG-MERITS page 2 of 2 Rev. B r

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