ML18033B497

Summary of 900710-11 Meetings W/Util at Plant Site Re Proposed Tech Spec Change Concerning Table 3.7-A, Primary Containment Isolation Valves & Containment Isolation Arrangements at Plant.List of Attendees & TVA Handouts Encl

ML18033B497
Person / Time
Site:	Browns Ferry
Issue date:	08/17/1990
From:	Moran D Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
Shared Package
ML18033B496	List: ML18033B495 ML18033B497
References
TASK-2.E.4.2, TASK-TM TAC-R00080, TAC-R00081, TAC-R00082, TAC-R80, TAC-R81, TAC-R82, NUDOCS 9008280187
Download: ML18033B497 (77)

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 406 1v )gg Docket Nos. 50-259, 50-260 and 50-296 LICENSEE:

Tennessee Valley Authority (TVA)

FACILITY:

Browns Ferry Nuclear Plant, Unit 2

SUBJECT:

SUHHARY OF HEETING MITH TVA CONCERNING PRIHARY CONTAINHENT ISOLATION AT BROMNS FERRY NUCLEAR PLANT% UNITS I, 2, AND 3 (TAC NOS. R00080,

R00081, AND R00082)

A working level meeting between TVA (the licensee) and the NRC staff was held on July 10 and 11, 1990 at the Browns Ferry Nuclear (BFN) Plant site.

The purpose of this meeting was to discuss a proposed Technical Specification (TS) change dealing with Table 3.7-A, Primary Containment Isolation Valves, and to obtain additional information about containment isolation arrangements at BFN.

TVA and the staff also went over previous TVA responses regarding THI Item II.E.4.2 and Appendix J testing of Containment Isolation Valves.

The meeting was conducted at the BFN Administration Building.

A list of the attendees for this two day meeting is provided in enclosure I; attendees at the exit meeting of July 11, 1990 are listed separately in enclosure 2.

TVA was very well prepared for the meeting.

All necessary documents to explain the subject matter were provided, and technical personnel conversant with the details were on hand as needed.

TVA presented the staff with a package of documentation that addressed each of the subject areas contained in the staff's agenda published in a prior meeting announcement (see enclosure 3).

At the onset of the meeting, TVA provided a historical narrative of all applicable correspondence in order to summarize the prior licensing basis of the BFN Primary Containment configuration.

Afterwards, and for most of the meeting, TVA and the staff reviewed in detail the BFN containment isolation valve arrangements (including their design basis);

TVA's response and conformance to post THI action item II.E.4.2 and 10 CFR 50, Appendix J; and the relevance of applicable 10 CFR 50, Appendix A, General Design Criteria (GDC).

TVA's handout material encompassing the subject areas of discussion is contained in enclosure 4.

The table of contents (see enclosure

5) for the documentation made available to the staff by TVA established a logical order to consider and discuss the containment penetrations, system by system.

The list of these penetrations (see enclosure

6) provides a table which specifies for each penetration:

applicable GDC; physical location; components; valve type; valve position; actuation signal; back-up isolation capability; other NRC defined bases; and applicable drawing number.

The drawings were annotated by colored highlighting to show the valves and piping runs pertinent to each penetration of each system.

Each penetration was discussed with TVA in detail.

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At the exit meeting, the staff requested that TVA consider the following recommendations:

Manual valves used as primary containment isolation valves should be locked (or sealed) and included within the BFN locked valve program.

Auxiliary Boiler and Demineralized Water Systems utilize in-series check valves for primary isolation.

TVA could diversify the isolation arrangement, by incorporating an associated block valve to establish a more positive means of isolation (i.e.

one check valve and one block valve).

For Residual Heat Removal recirculation and pump test lines, one of the isolation barriers is the suppression pool.

The staff does not consider the suppression pool an adequate barrier, and suggested that an existing test valve in the piping run be designated the isolation valve.

Some systems use two check valves in series as primary containment isolation.

Although this arrangement was part of the original design basis, and as such is acceptable, it would not be acceptable if evaluated to the current GDC.

However, most of these systems already have a downstream manual valve that could be identified in the BFN Emergency Operating Instructions (such valves would not require Appendix J testing) as additional assurance for long term isolation.

With regard to the adequacy of the Reactor Building Closed Cooling Water (RBCCW) System to function as a closed

system, the staff suggested that TVA could take the actions listed below as a method to resolve this issue:

l.

Assess the pipe restraint program for all drywell piping.

2.

Identify those components or sources in the drywell which could become missiles that would endanger RBCCW integrity inside containment.

3.

Establish procedures for manually isolating all coolers upon receipt of a valid isolation signal to minimize loss of RBCCW integrity.

Upon receipt of TVA's response to the aforementioned recommendations, an NRC safety evaluation report wi 11 be forthcoming that addresses their TS amendment application, and BFN conformance with TMI Item II.E.4.2 and Appendix J.

Original signed by David H. Moran, Project Engineer Project Directorate II-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1.

List of Attendees 2.

Exit List of Attendees 3.

Meeting Agenda 4.

TVA Handouts 5.

Table of Contents 6.

List of Penetrations OFC

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Docket Nos. 50-259,80-260

'and 50-296

'NCLOSURE 3

uwiVSa a'ATaS NUCLEAR REGULATORYCDMMlSSIQN WACHINOTON,0< C< 10555 July 2, 1990 M~OSLNOVM FAR:

FROM:

SUBJECT:

Date and Time:

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Purpose:

Frederfck J.

Hebdon, Director Pm)act Directorate II-0 Ofvfsfon of Reactor Pro)acts

'I II Office of Nuclear Reactor Regulation David H. Moran, Progect Kngfneer Profect Directorate II-4 Division of Reactor Prospects I-II Office of Nuclear Reactor Regulation FORTHCOMING MEETING HITH TKNNESSKK VALLEY AUTHORITY RKGARDING BROGANS FERRY NUCLEAR PLANT, UNITS 1, 2 ANO 3 July 10 12, 1990 8:00 a.m. - Close of Business Orownc t'cr".y Huc1alr Pler<t Administrative Building Hafn Conference Room D"catur, Al ba.,a Meeting wing TVA concerning propo d Technical Specificati-onon Change 251 dealfng with Table 3.7-A - Pr fmary Contafnaent Isolation Valves and to obtain additional information concerning contafnment isolation arrangements and response to 911 Item II.K.4,2 and Appendix J Testing of Containmant Isolation Valves.

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NRC J. Kudrfck J. Harold T, Ross 0, H. Moran TVA J. McCarthy K. R. Mulling.,

J. 0. Rolcott J.

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Heetfng Agenda cc w/enclosure:

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MEETING AGEHtjA Meeting Ob)ectfve:

To supplement, axfstfng fnformatfon to allow the staff's completion of a Safety Kvaluatfon Report,

Purpose:

To obtain addft1onal information concerning containment isolation arrangements and response to TNI Item II.E.4.2 and the Appendix J testing of containment fsolatfon valves.

Also, answers the enclosed questions about the ffgures lfsted'fn the supplement to the June 15, 1989 meetfng.

TVA Presentation a.S.

251

-Chronology

-Orfgfnal containment, 1solatfon design bases

• Reference Documents used for orfgfnal review

~ Compare original design basks to other BMR plants of the same vintage

'* Lfst upgrades to the orfgfnal fsolatfon design, ff any

-THI Item II.K.4.2

• Usfng PAID's; for each the penetration that differs from the current GDC provfde

- line sfze

- fsolatfon valve locatfon relative to the contafnmint (1fne length)

- actuation sfgnal(s)

- List other valves that may be consfdered fsolatfon valves

- Identify valves that are Appendix J tested and that are part of the IST program

- For penetratfons that do not meet current GDC requfrettlnts state what wou]d be necessaij to otirut therm

- For fnstrumenz taps and purge vent taps discuss location of the 1solatfon valve.

Also provfde a drawing that fdentfffes the seal1ng

- Dfscuss the plant status on response to. GL 84-09

- Discuss the role of the CAD system for LOCA events per the EOP's

- Discuss which KPG revision has been 1mplemented

-Plant Ma 1k Down

~ Yfew penetrat~ons fdentfffed as devfat1on from the GDC's, 'also photograph and vfdeot,ape the penetratfons

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-Conclud)ng Meeting

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$4ent)fy any future a'ct$ ons needed for complet)on of the Safety Evaluation

-References used by the staff for review ta dates NUREG<<0737 IX.E.4.2 NUREG-0578 NEDC-22253 NEDO 24782 June 15, 1989 meet4na slides TVA'.s July 13, 1989 5'ansmfttal of Brogans Ferry Technical Speclffcatfon Change T.S. 251, Supplement 1

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OF STAFF CDNCERNS RE~ AT!VE To CONTAlmSHT isoi~T~ON AT BROWt<'8 FF>nv, UHTT 2 Figure 24 V~nt to Standby Gas Treatment gues41ons Dfaaram fs not complete.

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- Cheek valve may not be used 'as the automatic isolation valve outside containment.

- Mhat fs the sfze of the lfne2 Ffgure 27, - Rx Sufldfng.Closed Coolfng Mater guest$ an Does FCV-70 47 saMsfy tha rI<uiriaients of CuC 57 one contafi4tent 1:olatfon val,e whfch fs capable of remote manual operatfon2 Ffgure 31 - HPCI Nfnfmum Flee gU85t10ll

- For the RHR Pump Test Lfne,. what type of valve is used for fsolatfon2 Ffgure 34 - HPCI Pump Suctfon Luastfon

- Essentfal system - why auto closure2 Ffgure 35 - RClC Mfnfmum Flow Ouestfon Mhat type of fsolatfon fs used on the lfne from the RHR Pump Test Lfne2

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ENCLOSURE 4 BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGUIMTION APPLICABLE CORRESPONDENCE

BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGUIMTION I.

A Review of the Original Ucenslng and Design Basis for the Containment Isolation Valve Configuration.

'he BFN licensing and design basis for the containment isolation valve configuration was originally established in Final Safety Analysis Report (FSAR) Section 5.2.3.5, Zsolation Valves, Appendix A, Conformance to Proposed AEC General Design Criterion, to the FSAR (Attachment 1), the Technical Specifications (Attachment 2) and in the original Safety Evaluation Report (SER) and its Supplements (Attachment 3).

Appendix A to the FSAR presents the interpretation, discussions, and conclusions on how the design of BFN conformed to the AEC proposed general design criteria at the time of the BFN design.

During the construction permit licensing process, Unit 2 was evaluated against the draft of the 27 General Design Criteria.(GDC) which were issued on November 22, 1965.

The design bases was reevaluated at the time of initial FSAR preparation and license application against the draft of the 70 GDC which were issued on July 10, 1967.

Draft GDC 53, Containment Isolation Valves, states that penetrations that require closure, for the containment function shall be protected by redundant valving and associated apparatus.

TVA concluded that the design of the plant was in conformance with draft GDC 53, since pipes which penetrate the primary containment and which connect to the primary

system, or are open to the drywell, are provided with at least two isolation valves in series.

Technical Specification Bases Section 3.7.D/4.7.D, Primary Containment Isolation Valves, states that double isolation valves are provided on lines penetrating the primary containment and open to the free space of the containment.

Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a LOCA.

Section 5.2.5, Zsolation Valves, of the original Safety Evaluation Report (SER) concludes that the isolation valves and their control systems have been

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reviewed to assure that no single accident or failure can result in a loss of containment integrity.

The sole exception occurs in the case of instrument lines and that was found to be acceptable.

SER Section 14.0, Conformance with General Design Criteria, concluded that there is reasonable assurance that the intent of the GDC for Nuclear Power Plants, published in the Federal, Register on March 21, 1971, in the final design of the station will be met.

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BROGANS FERRY NUCLEARPLAIW CONTAINMENTISOLATIONVALVECONFIGUIMTION (Continued)

II.

A Review of the Original Ucensing and Oeslgn Basis for the Postulation of HELBs Inside Containment.

The BFN licensing basis for postulation of HELBs inside containment was established in FSAR Section 5.2.4.6, Missile and Pipe Whip Prevention (Attachment 4), by receipt of, and in response to, Questions 4.1.4 and 5.19 (Attachment 5) which were issued by the Atomic Energy Commission (AEC) on March 25, 1971, in the original Safety Evaluation and its Supplements (Attachment 6), and in its review of the reconstituted design baseline of BFN Unit 2 (Attachments 7 and 8).

FSAR Section 5.2.4.6 and the responses to the NRC {}uestions provide the major design considerations for missile and pipe whip protections Emphasis is placed upon prevention of the occurrence of pipe breaks through design, procurement, quality control, inservice inspections, and the detection of pipe leaks, Energy absorbing material added to the interior of the drywell, Siding attached to the pressure

vessel, and Physical separation of safety-related components.

The response to Question 4.1.4 references FSAR Section 5.2.4.6 and states that it was. never intended to claim BFN impervious to the consequences of a whipping pipe, and it is because these consequences were not ignored that BFN is designed for the prevention of pipe failures.

Further, since the plant is in final stages of construction, it is necessary to consider installation complications, in addition to the contribution to safety, in the resolution of the pipe whip issue.

{}uestion 5.19 requested BFN describe the measures taken to assure that the damage caused by component failure within primary containment, and resulting in pipe whip and jet forces will not remove from service more than

one, redundant subsection of a vital system.

BFN responded by referring to Question 4.1.4 and by stating that redundant subsections of vital systems are physically separated within the primary containment to minimize the damage probability.

Zt is important to note that protection of non-vital systems was not questioned.

The logical conclusion being that the AEC did not consider the protection of non-vital systems to be of safety significance.

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BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGUIMTION (Continued)

The NRC accepted the adequacy of BFN's designed protection against pipe whip during the initial licensing process as documented in Section 5.2.2 of the Safety Evaluation of the TVA BFN Units 1, 2 and 3, dated June 26, 1972, in the revised section which was contained in Supplement 1, dated December 21,

1972, and in Section 3.0 of Supplement 4, dated September 10, 1973.

The NRC recently reviewed the reconstituted design basis for BFN concerning

)et impingement on miscellaneous steel inside primary containment as part of the Design Baseline Verification Program (DBVP).

A presentation on this topic was given to the NRC in Knoxville on March 6, 1989.

Meeting minutes are included as Attachment 7.

The NRC accepted TVA's position that )et impingement loads do not need to be applied to structural steel as part of the BFN design'asis, as documented in Znspection Report 50-259, 50-260 and 50-296/89-07 (Attachment 8).

The Znspection Report concluded that!

"Primary emphasis for jet impingement protection inside the drywell was directed toward protecting the primary containment.

Zn addition to the recirculation, main steam and reactor feedwater system restraints, further consideration to containment protection was provided by installation of honeycomb panels on the inside surface of the drywell shell and 5et deflectors over the main vent openings to the wetwellg Protection of other equipment in the drywell is inherent in the plant arrangement of equipment.

Redundant systems and devices are located on opposite sides of the drywell to minimize the concerns of dynamic forces associated with a pipe break.... TVA's response to'his item is acceptable and this item is closed."

Page 3 of 12

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1 BROWNS PERRY NUCLEAR PLAIN'ONTAINMENT ISOLATIONVALVECONPIGUID.TION (Continued)

III.

A Review of the Post-TMI Changes to the Design and Ucenslng Basis The initial request for action of BWR licensees was issued as IE Bulletin 79-08, Events Relevant to Boi.ling Water Power Reactors Identifi.ed During Three Mile Island Incident, dated April 13, 1979 (Attachment 9).

Question 2 of the Bulletin requested licensees review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to initiate contai.nment i.solation, whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.'VA responded by letter dated April 24, 1979 (Attachment 10).

TVA stated that the containment isolation design and procedures to ensure that proper isolation of the primary containment occurs upon initi.ation of Emergency Core Cooling Systems.

Reference was made to Part 5 of the Browne Ferry FSAR for further detai.ls regarding the primary containment isolation system design objectives and operation.

The NRC i.ssued a generic review of responses to IE Bulletin 79-08, by letter to all BWR licensees on June 20, 1979 (Attachment 11).

In this letter, the ta def nes c osed coo n

water as a closed s stem w ic e et ates the containment but does not connect to the reactor coolant ressure boundar or t e co ta e t at os here.

The NRC states that system isolati.on i.s determined by the Architect/Engineer and BWR owners should check for implementation.

The staff also states that it. does not'have the resources or the time to resolve these issues with each licensee and requests the formation of a BWR Owners'roup.

The NRC requested additional information by letter dated July 20, 1979 (Attachment 12).

The NRC requested TVA verify that the review considered initiation of containment i.solation of all lines penetrati.ng containment (i.ncluding those designed to transfer potentially radioactive gases and liqui.ds out of contai.nment) upon all automatic initiation of safety injection.

The NRC issued NUREG-0578 by letter to all licensees on July 25, 1979 (Attachment 13).

TVA responded.to the NRC's July 20, 1979 request by letter dated August 6, 1979'(Attachment 14) and stated that the review considered all lines which penetrated containment.

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By letter dated September 13, 1979 (Attachment 15), the NRC requested licensees begin to implement the actions contained in NUREG-0578, as modified by this letter.

Commitments to meet these requirements on the inclosed implementation schedule were requested within 30 days of receipt of the letter.

By letter dated October 13, 1979 (Attachment 16), the NRC requested all operating nuclear plants commit to comply with the requirements of NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations.

NUREG-0578, Recommendation 2.1.4, Containment Isolation Provisions for PWRs and BWRs, states the need for careful reconsideration of the isolation provisions of non-essential systems inside containment.

Reconsideration should include the identification of those systems that can be isolated indefinitely and those systems that should be selectively isolated only after it is established that they are not essential to continued core cooling or performance of engineered safety features.

TVA submitted a response to NUREG-0578, by letter dated October 17, 1979 (Attachment 17).

TVA stated that the BFN design will comply with NRC requirements on the automatic isolation of nonessential systems.

By a separate letter dated October 17, 1979 (Attachment 18), the BWROG informed the staff that Owners Group would coordinate resolution of this item with the, Staff.

However, submittal of these positions by the Owners Group does not constitute a commitment by or for any individual utility.

These commitments will be made in each individual utility implementation letter.

Therefore, while the BWROG submittals, documents, and positions are discussed in the remainder of this section, they do not constitute a part of the BFN licensing or design basis.

Their inclusion is necessary since it provides important information regarding the basis for the resolution of the issue and no TVA/NRC correspondence was located which provides this level of detail.

By letter dated October 30, 1979 (Attachment 19), the staff provided clarification to the short-term requirements.

The NRC stated that non-essential systems should be automatically isolated by containment isolation, signals.

Licensees were requested to submit implementation schedules within fifteen days of receipt of this letter.

By letter dated November 16, 1979 (Attachment 20),

TVA provided schedular information regarding implementation of the NUREG-0578 commitments.

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BROGANS FERRY NUCLEAR PLAKI'ONTAINMENT ISOLATIONVALVECONFIGUIMTION (Continued)

TVA and NRC met on July 12, 1979 to discuss post-TMI requirements and future regulatory guidance.

BFN docketed plant specific system information by letter dated November 27, 1979 (Attachment 21).

TVA provided information such as the RBCCW safety classification, primary containment isolation system data, design requirements for containment isolation barriers,

codes, standards and guides used for the design of the containment isolation system and system componentsg isolation provisions for lines forming a closed loop inside the drywell.

By letter from GE to TVA, dated December 14,

1979, GE recommended no automatic isolation be provided for the RBCCW containment isolation valves (Attachment 22) since the failure of the piping would have to occur after a LOCA to allow any communication between the reactor or the drywell and the area outside of containment.

By letter dated December 19, 1979 (Attachment 23),

TVA committed to be in full compliance with NUREG-0578 short-term requirements by January 1,

1980.

By letter dated December 27, 1979 (Attachment 24), the NRC issued a Safety Evaluation regarding TVA's responded to IE Bulletin 79-08. It concluded that the licensee's review of containment isolation initiation design and procedures satisfied the intent of IE Bulletin 79-08,. Item 2.

The NRC's other December 27, 1979 letter (Attachment 25), stated that the NRC would perform a post-implementation review of TVA's method of implementing each requirement.

The Commission issued a Confirmatory Order on January 2, 1980 (Attachment 26) to implement all "Category A" lessons learned requfi.rements (excluding 2.1.7.a) prior to plant operation after January 31, 1980.

TVA responded to the NRC's October 30, 1979 letter to all operating reactors (in which documentation of the method used for implementation of the NUREG-0578 short-term requirements was requested) on January 17, 1980 (Attachment 27)

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TVA's position of Item 2.1.4, Containment Isolation, was that non-essential systems which were not required for post-accident mitigation are isolated automatically upon receipt of a primary containment isolation signal or are provided with manual valves which are locked closed when containment integrity is required.

Page 6 of 12

BROWNS FERRY NUCLEAR PLANT CONTAII~ENTISOLATIONVALVECONFIGUfMTION (Continued)

By letter dated February ll, 1980 (Attachment 28), the NRC informed TVA that a site vi.sit would be held on February 19 and 20, 1980 to verify implementation of the TMI-2 lessons learned short-term (Category A) items described in NUREG-0578.

One of the principal areas reviewed would be Contai.nment Isolation, Item 2.1.4, and specifically the criteria for classifying essenti.al versus non-essential systems.

The results of the site review are documented in the NRC's Safety Evaluation, dated February 29, 1980 (Attachment 29).

For Item 2.1.4, Containment Isolation, the Safety Evaluation states the li.censee has conducted a system by system analysis of the containment penetrations and has concluded that all non-essential systems receive the proper isolati.on signals.

Zt concludes that the licensee has adequately conformed to the requirements of this item.

In response to the NUREG-0578 item, the BWR Owner's Group (BWROG) evaluated the RBCCW system as documented in NEDO-24782, BWROG Implementation:

Analysis and Positions for Plant Unique Submittals, dated August 1980 (Attachment 30).

NEDO-24782 class'ified the RBCCW as a non-essential system since it is used for normal operation only and is not requi.red for design basi.s accident mitigation.

NEDO-24782 further concludes that the RBCCW is a closed system since it is not open to the reactor or to the drywell atmosphere.

Thereforef no automatic isolation of the containment isolation valves is required.

BFN's plant specific FSAR positions on draft GDC 53 and primary contai.nment isolation valves support the BWROG conclusion.

NUREG-0660 (May 1980),

Item IZ.E.4.2 (Attachment 31), reiterates the requirements on NUREG-0578 and further states that NRR will review li.censee desi.gnations of essential versus nonessential systems that have lines penetrating the containment structure and will develop guidance for industry use and for SD (Undefined NRC Office) use in the preparation of a regulatory guide....

SD will issue Revision 1 to Regulatory Guide 1.141, Containment Isolation Provisions for Fluid Systems, by July 1980.

SD will issue Revision 2 to Regulatory Guide 1.141 to include the designation of essential versus nonessential systems by June 1981.

As previously stated, Revision 0 for Regulatory Guide 1.141 was issued for comment in April 1978 and has never been formally issued.

Page 7 of 12

0

lg BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGMMTION (Continued)

By letter dated September 5,

1980 (Attachment

32) to all licensees and construction permit holders, the NRC stated that all operating plants were required to be in compliance with Item II.E.4.2, Positions 1 through 4 by January 1, 1980.

Additional guidance will be provided by NRR on the classification of essential and nonessential systems by October 1, 1980.

All operati.ng plants will then be required to provide a revised list of their essential and nonessential systems.

On October 31, 1980 (Attachment 33), the Staff issued a Generic Letter to all licensees which transmitted NUREG-0737, Clari.fi.cation of TMZ Action Plan Requirements, under the provisions of 10 CFR 50.54(f).

The provisions of 10 CFR 50.54(f) requi.re licensee to submit information when requested by the Staff ln,order to justify why the applicable license should not be modified, suspended, or revoked.

Regulatory requirements or required modifications can not be imposed under the provisions of 10 CFR 50.54(f).

The NUREG-0737 Action Item potentially germane to this issue is Item ZI.E.4.2, Containment Isolation Dependability, Positions 1 and 3..

Position 1 - Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e., that there by diversity in the parameters sensed for the initiation of containment i.solation).

(NRC Clarification in NUREG-0737:

The reference to SRP 6.2.4 in Position.

1 is only to the diversity requirements set forth in that document.)

Evaluation of Position 1 - NUREG-0737 further states that there were no changes i.n the requirements since NUREG-0660, NRC Action Plan Developed as a

Result of the TMI-2 Accident, was issued.

NUREG-0660, Item ZZ.E.4.2.1, specifically states thht the requirements were to provide containment isolation on diverse signals in conformance wi.th Section 6.2.4 of the Standard Review Plan.

The requirement for diverse signals is not germane to the adequacy of the number or type of RBCCW containment isolation valves.

Also, at the time that NUREG-0737 was evaluated, these RBCCW valves were not considered containment isolation valves and were not included in the BFN 10 CFR 50, Appendix J program.

Page 8 of 12

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BROWNS FERRY NUCLEARPLY CONTAINMENTISOLATIONVALVECONFIGUIMTION (Continued)

Position 3 - All nonessential systems shall be automatically isolated by the containment isolation signal.

(NRC Clarification in NUREG-0737:

Revision 2

to Regulatory Guide 1.141 will contain guidance on the classification of essential versus nonessential systems and is due to be issued by June 1981.

Requirements for operating plants to review their list of essential and nonessential systems will be issued in conjunction with this guide including an appropriate time schedule for completion.)

Evaluation of Position 3 - First, automatic isolation is the specific issue addressed by this item.

Zt is important to note that this item does not address the number of isolation valves per line or the adequacy of check valves in use as containment isolation valves.

Second, Revision 0 for Regulatory Guide 1.141 (Attachment

34) was issued for comment in April 1978 and the Regulatory Guide has never been formally issued.

When NUREG-0737 was issued by Generic Letter on October 31,

1980, Item ZZ.E.4.2 was listed as being completed.

NUREG-0737 states that the requirements of this item were originally issued on September 14, 1979 and clarifications were issued on October 30, 1979.

Licensing submittals were required by January 1,

1980 (Eleven months prior to the issuance of NUREG-0737) and post implementation review was required.

TVA submitted its response to NUREG<<0737 by letter from L. M. Mills to H

R. Denton, on December 23, 1980 (Attachment 35).

Zt stated that Browns Ferry is in compliance with items 1>>4 with the exception of item 4 on Browne Ferry unit 1 regarding ganged resetting of certain isolation valves.

No additional details or clarifications were provided.

At the request of Peach Bottom, the BWROG re-evaluated this issue as documented in NEDC-22253, BWR Owners'roup Evaluation of Containment Isolation Concerns, October 1982 (Attachment 36).

The BWROG held a,

teleconference with the Staff on April 1, 1982 to determine if it was the Staff's intention to backfit the Standard Review Plan, Section 6.2.4, definition of a closed system inside containment via NUREG-0737.

The Staff stated that the RBCCW system configurations and isolation provisions already licensed are acceptable, provided that:

1)

At least one isolation valve is provided on each line as close to the containment as practical, Page 9 of 12

BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGURATION (Continued) 2)

The isolation valves are automatic or capable of remote manual operation, 3)

Power supplies and controls for the isolation valves are safety grade (classified "important to safety"),

and 4)

Piping systems from containment out to, and including, the isolation valves are safety grade.

No documentation was located whi.ch dockets this verbal Staff position either generically or specifically to BFN.

NEDC-22253 further concludes that the BWROG position is thats 1)

That the RBCCW is a closed system since it is not open to the reactor or to the drywell atmosphere.

Therefore, no automatic isolation of the containment isolation valves is required.

2)

That the RBCCW should not be automati.cally isolated since its continued use will tend to mitigate the consequences of an accident even though its operation is not required.

3)

Although NUREG-0737 appears to backfit SRP 6.2.4 requirements to the RBCCW lines@ the NRC clarification is that it was not the Staffs'ntention to backfit the SRP 6.2.4 definition of "closed systems inside containment" to operating plants.

Therefore, the existing designs need not be evaluated against the SRP or Regulatory Guide 1.141.

4)

The RBCCW containment isolation provisions should includec A)

At least one containment isolation valve on each line outside containment and as close to the containment as practical (a simple check valve may not be used),

B)

The isolation valves shall be either automatic, or capable of remote manual operation, and C)

Commensurate with the safety function of containment isolation, power supplies and controls for the isolation valves should be safety grade and the piping system from containment out to and i.ncluding the isolation valves should be safety grade.

Page '10 of 12

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BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGUIMTION (Continued)

BFN's RBCCW piping is not classified as safety class.

The RBCCW was constructed to ANSI B31.1 criteria.

However the verbal NRC position that it should be safety grade has not been located on the BFN docket and therefore, is not a requirement.

BFN uses a check valve as part of the RBCCW containment isolation boundary, which conflicts with the above BWROG position.

However, check valves are allowed to be used as containment isolation valves at BFN as documented in the FSAR.

While BFN's RBCCW containment isolation valve configuration is not in verbatim conformance to Position 3 or to the statement made in the October 17, 1979 or October 31, 1980 submittals regarding the automatic isolation of this nonessential system, it is in conformance with the original licensing and design basis for BFN and with the BWROG generic position'n NUREG-0578.

By letter dated January 2,

1980, Philadelphia Electric (Attachment 37) responded to the short-term lessons learned in NUREG-0578 for Peach Bottom.

They specifically stated that the isolation valves on the Reactor Building Cooling Water and Drywell Chilled Water Systems do not receive automatic isolation signals.

It is their position that these valves should not be automatically isolated since their continued use will tend to,mitigate the consequences of an accident.

In addition, 10 CFR 50, Appendix A, GDC 57 allows the use of a remote-manual valve on lines such as these that are neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere.

In the NRC's Safety Evaluation of Peach Bottom (Attachment 38), dated February 26, 1980, the NRC concluded that this technical justification was adequate.

In conclusion, BFN is in conformance with the original and current licensing and design basis for BFN and is in general agreement with the overall BWROG generic position on TMI>>2 lessons learned.

Page 11 of 12

BROWNS FERRY NUCLEAR PLANT CONTAINMENTISOLATIONVALVECONFIGUIMTION (Continued)

The primary HELB design ob)ective and licensing basis was to ensure that pipe breaks inside containment do not damage the primary containment vessel and the addition of energy absorbing material would minimize the potential for a breach the primary containment.

No consequential failure of safety related systems or the breach of other system's integrity by a HELB was required to be postulated.

NRC Regional Office Instruction No. 2222, Plant-Specific Backfit Procedures, states that once the SER is issued signifying staff acceptance of the program contained in the FSAR, the licensee should be able to conclude that his commitments in the FSAR satisfy the NRC requirements for a particular area.

If the staff were to subsequently require that the licensee commit to additional action other than that specified in the FSAR for the particular

area, such action would constitute a backfit. If there were tacit acceptance by the staff, by being silent on the issue for an extended period of time, then staff action to force a change would be a backfit.

Page 12 of 12

ENCLOSURE 5

CONTENTS 1.

Penetration Listing 2.

Simplified Flow Diagrams 3.

Main Steam System 4.

Demineralized Water System 5.

Reactor Feedwater System 6.

Auxilliary Boiler System 7.. Control Air System 8.

Service Air System 9.

Sampling and Water Quality System 10.

Standby Liquid Control System 11.

'Containment Ventilation System 12.

Recirculation System 13.

Reactor Water Cleanup System 14.

Reactor Building Closed Cooling Water System 15.

Reactor Core Isolation Cooling System 16.

High Pressure Coolant Injection System 17.

Residual Heat Removal System 18.

Core Spray System

CON'IENTS 19.

Containment Inerting System 20.

Radwaste System 21.

Containment Atmosphere Dilution System 22.

Radiation Monitoring System 23.

Traversing Incore Probe System 24.

Appendix A:Interim GDC 53 25.

Appendix B: GE NEDC-222253 26.

Appendix C: GDC 55 27.

Appendix D: NUREG-0800 28.

Appendix E: FSAR 5.2.3.5/5.2.3.6 29.

Appendix F: FSAR 7.3.2 30.

Appendix G: FSAR 7.3.4.1.2 31.

Appendix H:NRC Information Notice 89-55 32.

Appendix I: G.E. System Design Specification 22A1110 33.

Appendix J: G.E. System Design Specification 22A1169 34, Appendix K:NUREG-0737 35.

Appendix L: Appendix A/J Timeline 36.

Appendix M: Safety Guide 1 1

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TORUS WATERTEMPERAll%K NIA NIA NIA N/A NIA NIA TW.64 161C NIA TW-64-161D NIA TW.64-161E NIA TW.64-161F NIA TW.64.161G 2.47E817-1 2-47E817-1 2-47E817-1 2.47E817-1 2-47E817-1 TORUS WA R

MPERA IA IA W.

X.235A X-235B TCRUS WAIFR lEMPERATl%%

TCRI/S WATERllMPERAH%%

NIA NIA NIA W.64 ~ 162A NIA TW.64-162B 2.47E817-1 2-47E817-1 Page 21

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IWOUT BACK-UP150 QK0 DEFNEDBASfS APPLfCABLE TORUS WATERTEMPERATURE TCRfJS WATER lEMPERATURE TCRUS WATER TEMPERATURE NIA N/A IA NIA NIA IA TW.64-162C TW-64-162D

~64-6 2-47E817-1 2.47E817.1 TCRUS WAlERTEMPERATURE TORUS WATER TEMPERATURE TCRUS WATER lEMFERATURE PSC TIE.INTO RHR PSC TIE-INTO CS DffVWEIL 1 tEAD SHEAR LUG ACCESS COVERS 1.8 NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA 53 NIA NIA NIA NIA NIA a/r TW-64-162F TW-64-162G TW-64-162H 74-792 74 ~802 74 ~803 74-804 75.606 7

~607 75.609 75-610 74-220A PS-74.94 PT-74 ~ 94 OIP 0/P OIP OIP OI OIP OIP NIA NIA NIA NIA NIA IA NIA NIA 74-804 74-803 74-802 74-92 75-607 5-606 75 ~ 610 75-609 NTERIMGDC 53 tfTER1MGDC 53 NTERIMGDC 53 9/TERIMGDC 53 fN RIMGDC 53 NTERIMGDC 53 RIM GDC 53 SAFETYGUGE 11 2-47E817 1

2-47E817-1 2-47E817-1 2-47E811-1 2-47E81 4-1 719E532 GE 719E532 GE 2-47E811-1 Page 22

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lleeti~r Surrtt'arg~ Meroranda for ~Tri Reports or Site Visits Docket Fi.lc 3 NRC PDR Local PDR F. Miraglia J; Partlow S. Varga G. Lainas F. Hebdon S. Black M. Krebs T. Ross D. Moran R. Borchardt M. S. Little OGC

f. Jordan J. Kudrick J. Harold ACRS (10)

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