ML20245B280

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Summary of CRGR Meeting 113 on 870407 Re Proposed Reg Guide EE-006-5, Qualification of Safety-Related Lead Storage Batteries for Nuclear Power Plants & Proposed Generic Ltr on Containment Integrated Leakage Tests.Encls Withheld
ML20245B280
Person / Time
Issue date: 05/21/1987
From: Jordan E
Committee To Review Generic Requirements
To: Stello V
NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
Shared Package
ML20244D710 List:
References
TASK-EE-006-5, TASK-RE NUDOCS 8706030443
Download: ML20245B280 (30)
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Text

_ _ _ . . . .

Victor Stello, Jr.

May 21, 1987 /

MEMORANDUM FOR:

Executive Director for Operations AOappt FROM: Edward L. Jordan, Chairman Connittee to Review Generic Requirements 6G1Au-

SUBJECT:

MINUTES Ot CRGR FEETING NUMBER 113 The Committee to Review Generic Requirements (CRGR) met ton riday, April 7, 1987, from 9 12 a.m. A list of attendees for this meeting enclosed (Enclosure 1). The following items were addressed at the meeting:

1. W. Morris and S. Aggarwal, RES, presented a proposed Regulatory Guide EE-006-5, " Qualification nf Safety _Related Lead Ctorage Batteries for Nuclear Power Plants." The CRGR recommended that the proposed Regulatory Guide should be issued for public conunent, af ter modifications to reflect CRGR comments. This matter is discussed in Enclosure 2.
2. L. Shao, A. Thadani and J. Pulsipher, NRR, presented a proposed generic letter intended to advise licensees / applicants of the current regulations concerning containment integrated leakage tests. The CRGR recommended that NRR prepare a new proposal that would result in an immediately effective rule change to modify the current regulations. The new proposal would be reviewed by the CRGR. This matter is discussed in Enclosure 3.

Enclosures 2 end 3 contain predecisional information and. tterefore, will not be released to the Public Document Room until the NRC has considered (in a public forum) or decided the matter addressed by the information.

In accordance with the ED0's July 18, 1983 directive concerning " Feedback and Closure on CRGR Reviews," a written response is required from the cognizant office to report agreement or disagreement with CRGR recommendations in these minutes. The response, which is required within five working days af ter receipt of these meeting minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with the CRGR recornnendations, to the EDO for decisionnaking.

Questions concerning these meeting minutes should be referred to Tom Cox (492 9855).

Onginal Sigr:ed by:

Edward L. Jordan, Chairman Committee to Review Generic Requirements

Enclosures:

Distribution: w/o encl.

As stated Central File J. Zerbe PDR (NRC/CRGR) CRGR Ctaff (w/ enc.)

cc: See next page J. Clifford CRGR CF (w/ enc.)

5. Treby CRGR Meeting File

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SECY Office Directors Regional Administrators l CRGR Members l W. Parler W. Morris L. Shao l

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Enclosure 1 LIST OF ATTENDEES CRGR MEETING NO. 113 April 17,1987 . _ .

CRGR MEMBERS E. Jordan W. Morris (for D. Ross)

T. Martin R. Bernero J. Sniezek J. Scinto OTHERS R. Bosnak S. Aggarwal M. El-Zeftawy D. F. Sullivan J. Zerbe  ;

T. Cox J. Conran J. Clifford L. Shao J. Pulsipher A. Thadani J. Kudrick J. Craig M. Taylor E. Jakel G. Arndt C. Mullins i

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Enclosure 2 to the Minutes of CRGR Meeting No. 113 Qualification of Lead Storage Batteries (Proposed Reg. Guide EE 006-5)

G. Arlotto (RES) and S. Aggarwal (RES) presented for CRGR review a proposal to issue for public comment a Reg. Guide that would endorse without exceptions an existing industry standard (ANSI /IEEE Std. 535-1986) which provides guidance on environmental qualification and accelerated aging (prerequisite to seismic testing) of safety-related lead storage batteries. Copies of the briefing slides urri be the staff to guide their presentation and discussions with the Committee u this meeting are attached to this Enclosure.

The package submitted by RES for CRGR review in this matter was transmitted by memorandum dated December 9, 1986, E. S. Beckjord to J. H. Sniezek; that package included the following items:

1. Proposed Reg. Guide EE 006-5, Draft 1, dated October 14, 1986,

" Qualification of Safety-Related Lead Storage Batteries for Nuclear Power Plants"

2. ANSI /IEEE Std. 535-1985, "IEEE Standard for Qualification of Class IE Lead Storage Batteries for Nuclear. Power Generating Stations."
3. " Summary of Proposed Reg. Guide for CRGR Review", dated October 10, 1986 (as required by CRGR procedures).
4. NRC Staff Comments on Proposed Reg. Guide:
a. NRR Comments (Memo dated August 12, 1986, Rosa to Aggarwal)
b. IE Comments (Memo dated August 27, 1986, Grimes to Shao)
5. RES. Responses to Staff Comments: j
a. Memo dated September 10, 1986, Aggarwal to Rosa {
b. Memo dated October 10, 1986, Aggarwal to Heishman/Baer l
6. Supplemental supporting material for the proposed Reg. Guide (specifically a separate, enhanced Regulatory Analysis document) was transmitted to CRGR by memorandum dated March 18, 1987, G. A. Arlotto to J. E. Zerbe; that supplemental package is attached to this Enclosure.

Major points of discussion at this meeting regarding the proposed Reg. Guide were as follows:

1. The Committee questioned why the proposed Reg. Guide applied only to safety-related batteries, not to all station batteries. The staff responded that the same qualification procedures could have been spe-cified for all batteries. In endorsing industry Standards in Reg. Guides, however, the general intent is te endorse the existing standard without exceptions where possible. As written, this Standard is applicable to safety-related battery qualification only. So endorsing t5e Standard for applicability in the qualification of all station batteries would have required taking an exception to the existing Standard where no need or justification was seen for doing so.
2. The Committee questioned the adequacy and/or prudence of the guidance given in Section 8.2 of the Standard proposed for endorsement. Speci-fically, they expressed concern regarding the reference to use of high temperatures in the accelerated aging process that could result in the need to support the walls of the cell being tested to prever,t bulging.

There was indication in the documentation submitted in connection with this package that the use of such high temperatures in battery agi.ng/ test procedures could cause premature failure of the artificially aged cells in service due to cracking of cell walls and flaking of battery plates.

(See Background Items 1.d.2 and 1.e.2.) The Committee also wondered whether such high temperatures might not lead to overstressing of internal metal parts of batteries undergoing qualification testing, due to excessive expansion of battery plates against their supports. The staff defended the practices endorsed in the Standard because (a) it is I

necessary to have some feel for how batteries near end of life would behave in response to earthquake induced stresses, and (b) and the aging / test methods in question appear to represent a consenste of industry expert opinion regarding best available state-of-th -art procedures for obtaining such information.

The Committee also questioned why naturally aged batteries removed from operating plants for replacement had not been used in such testing. The staff stated that some testing had been done by Sandia Laboratories; and the results obtained were generally consistent with test results on artificially aged test batteries (enough so to give some confidence in the artificial aging / test methods endorsed by the Standard). However, the batteries in question generally require replacement on about a ten year cycle; and battery design has been changing so rapidly that test

! results on the naturally aged batteries available from the earlier i " cycles" cannot be taken as representative of the new replacement batteries. It is r.ecessary, therefore, to continue to employ artificial aging methods for qualification of new battery designs that are used as replacements.

After considerable discussion on these points the Committee recommended, and the staff agreed, th t public comment should be solicited on the acceptability of using artificial aging methods that produce extreme effects (e.g., high temperatures) in the cells being tested that are significantly outside the ranges typical of anticipated service conditions (particularly where those effects are severe enough to possibly lead to premature failure of batteries in service). Also, because of the fact that in general the batteries in question are not i good for the life of the plants in which they are installed, the

! Committee further recommended that consideration be given to including explicitly within the scope of the proposed Reg. Guide the qualification of replacement batteries in operating plants. As proposed, the Reg.

Guide is applicable only to new plants, and operating plants that voluntarily commit to its provisions. In view of present circumstances within the industry, however, and the current dim outlook for new plant orders, the guidance provided by the proposed Reg. Guide would seem to be more useful in the operating plant context. (This would, of course, introduce backfitting considerations not evaluated in the review package submitted.) The staff agreed with these comments, and with the Committee's recommendation in this regard, and will inc-lude request for comment on this point in the final package that is issued.

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RECOMMENDATION TO THE ED0

'As a result of their review of this matter and the discussions with the staff at this meeting, the Committee recommended that the proposed Reg. Guide be issued for comment, subject to modifications that reflect CRGR's comments and recommendations noted in the preceding, specifically (a) regarding the ade-quacy of accelerated aging methods endorsed in the subject Standard, and (b) regarding applicability of the proposed Reg. Guide to replacement batteries in operating plants. Modifications made by the staff are to be reviewed by the CRGR staff prior to issuance of the revised package for comment.

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1 REGULATORY GUIDE EE 006-5 SAFETY RELATED BATTERIES PRESENTATION BY SATISH K. AGGARWAL 1

CRGR MEETING 1

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s Batteries Must Be Able to Provide Needed Power .for the Time Required to Recover from a Station Blackout

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  • Control for critical functions

- Start of diesel generators

- Operating circuit breakers

- Feedwater control (PWR)

- Reactor core isolation cooling system (BWR)

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Lead-Acid Storage Batteries Materials of Construction Cell Components Materials of Construction Grids (all) Lead-calcium alloy Active material Lead dioxide (positive plate)

Lead with expander added (negative plate)

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Rubber / glass mat Electrolyte Sulfuric acid and water witn density of 1.200 to 1.220 g/cc Vent Fused alumina tunnel Top conouctor Lead-calcium alloy Terminals Lead-calcium alloy or Copper inserts in lead-calcium posts Contamer and cover Polycarbonate (LEX AN),

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Battery- Failure Events Reported in. LERs Number . Percent age Failure Cause

67 27 Low specific gravity' Personnel (operation, maintenance, testing) 52 21 ..

27 11 Insulticient charge 22 9 Defective /seak cells 14 6 Low electrolyte solution level

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Detective procedures 11 4

Charger malfunction 9 1

8 3 Design lactication, construction 8 3 High electrolyte solution level 5 2 Unknown causes 4 .

2 Corres>on 4 2 Short circutt '1 3

Normal wear / natural end of life 1

<1 Extreme environment 248 100 u..nu Battery Failure Events Reported in NPRDS Failure Cause Number Percentage Wearout 28 36 Unknown 24 31 Manuf acturing defect 6 8 Engineering / design 5 6 Incorrect procedure 5 6 Installation error 4 5 Maintenance / testing 4 5 Other davices 1 3 78 100 n,w ,

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Failures Reported in NRC information Notices

  • Cracked and leaking cases
  • Flaking of plates and book area of plates
  • 0xidation of plates causing separation or reduced plate to bus bar connection
  • Improper float voltages
  • Not following established procedures

- Incorrect discharge rate during tests ,

- Failure to correct specific gravity for temperature

- Failure to compare intercell resistances with various valu.es

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  • Shedding of active material
  • Reduced capacity electrolyte
  • Corrosion of positive' grid
  • Reduced capacity
  • Hycrolysis
  • Reduced capacity Ditt and moisture
  • Low resistance to ground
  • Ground f aults/ snorts on cases
  • 0xidation/ corrosion of
  • Battery f ailure or terminals reduced capacity Gases
  • 0xidation/ corrosion of
  • Reduced capacity terminals l Seismic event
  • Broken positive plates.
  • Battery f ailure straps, and/or terminals
  • Cracked cases
  • Leaking of electrolyte
  • Broken posts / connectors
  • Battery f ailure a.xm

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BACKGROUND --

QUALITY A'SSURANCE' PROGRAM DESIGN PRODUCTION QUALIlY CONTROL QUALIFICATION

  • : INSTALLATION MAINTENANCE-PERIODIC TESTING NO: SPECIFIC GUIDANCE PUBLISHED BY NRC FOR QUALIFICATION:

IEEE STD 535-1979: REFERENCED IN.SRP - NOT ENDORSED IEEE STD 535-1986: MEETS STAFF'S: APPROVAL

- REGULATORY BASIS:

GDC 1, 2,11 & 23 R.G 1.100 SEISMIC SAFETY-RELATED (CLASS lE) EQUIPMENT IN MILD ENVIRONMENT SECTION 3.11 (PAGE 5) S.R.P.

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R.G.1.NO-b SEISMIC REVIEWi SOME KIND OF PRE-AGING. PRIOR

.T0 TESTING NEW STAFF POSITION ON NEW PLANTS -

CODIFIES! EXISTING GOOD PRACTICES BY

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CONSISTENT WITH THE MORE' GENERAL GUIDANCE E PROV'0ED BY THE SRP.

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[ SAFETY RELATED BATTERIES LOCATED OUTSIDE CONTAltlMENT (MILD ENVIRONMENT) e:

REGULATORY POSITION: ENDORSES IEEE 535-1986 WITHOUT ANY EXCEPTIONS

. QUALIFICATION: - PRE-AGING-

- SEISMIC IMPLEMENTATION '

(1) CP ISSUED AFTER ISSUANCE OF THIS-RG (2)- OL DOCKETED 6 MONTHS OR MORE (3) VOLUNTARILY-STAFF COMMENTS NRR-L (1) ANALYSIS: NOT ACCEPTABLE (2) RADIATION AGING JE (1) SURVEILLANCE: BATTERY - RACK DETERIORATION -

SEISMIC EVENT (2) COMPONENTS FAILURE DURING AGING L

(3) 50/50 FAILURE / SUCCESS RATE LL:_ -__. -- ___- __ - _ _ _ - ___ _ ______ ___- - - __ ________- _ _ _ _ _ _ _ _ _ _ : _ Q

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4 TECHNICAL _ SECTIONS '0F' IEEE STD 535-1986 1,

5.0 -QUALIFICATION METHODS _.

5.Li- QUALIFIED LIFE 6.0 TEST PLAN 7.0 REQUIREMENTS FOR PROTOTYPE TESTli16-8.0 PRESCRIPTIVE REQUIREMENTS FOR TYPE TESTING

' AND. ANALYSIS PROCEDURES 8.3 SEISMIC TESTING

- 9.0 DOCUMENTATION.

IMPACT l EX1DE C8D G0ULD  ;

QUALIFIFIED T0 IEEE STD 535-1979 $30,000 - $180,000 NO SIGNIFICANT DIFFERENCE WITH 1979 & 1986 VERSIONS REGARD TO PRE-AGING i

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'M With ' respect to . Paragraph 1.2, "Appbcation," ' of to the design, reliability, qualifiertion. and testabiht) of

. !EEE Std ~603 1980,- in the context of NRC classifica-the power.' instrumentation, and control portions of tion,- the . fire protection system is not classified - as a safety systems as modified and supplemented by the l safety related system. Additionally, in applying the following:

I- criteria of the standard, it is helpful to understand that the - following are considered to be synonymous: (1) . l. The term " safety syst'e m" used throughout IEEE electric portions of the safety system, (2) Class IE Std 6031980 should be understood to be synonymous equipment, and (3) safety-related electric equipment as with " safety-related system," in which the term " safety-defined in i 50.49 of 10 CFR Part 50. It should also ; related" has the meaning stated in 5 50.49(b)(1) of 'l0 be noted that the' scope of the standard is broader thani CFR Part 50. -.

d(l), (2), and (3) above since, for example, pr.eumatic!

instruments may be part of the safety system. 2. For displays for manually controlled actions covered in Section 5.8.1 of IEEE Std 603-1980, the

The following is a brief discussion of the basis of provisions for Type A instruments in Regulato:y Guide each regulatory position: 1.97, " Instrumentation for" Light Water-Cooled Nuclear Tower Plants To Assess Plant and Environs Conditions

' l. The terms " safety system" and " safety-related sys- During and Following an Accident," should be followed tem" have evolved separately, and it is essential in apply- instead IEEE Std 497-1977.

ing IEEE Std 603 1980 that the relationship of the terms

-(as stated in Regulatory Position .1) be understood. 3. Instead of the first sentence in Section 6.3.l(!) of IEEE Std 603 1980, the first sentence in Section 4.7.4.1

2. IEEE Std - 497 1977 is referenced in IEEE Std of IEEE Std 279 should be used: " Alternate channels,

' 6031980, but some of the requirements of IEEE Std not subject to failure resulting from the same single 497 1977 are at variance with current regulatory prac- event, shall be provided to limit the consequences of tice. Hence, in Regulatory Position 2, reference to IEEE this event to a value specified by the design bases."

Std 4971977 in IEEE Std 603 1980 is being replaced

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with reference to Regulatory Guide 1.97, ." Instruments- 4. Section 6.3.1 of IEEE Std 603 1980 references tion for Light Water Cooled Nuclear Power Plants To Figure 7, a decision chart for applying the requirements Assess Plant and Environs Conditions During and Fol- of the section. Figure 1 of this guide should be used lowing an Accident," which provides specific recommen- in lieu of Figure 7 of IEEE Std 603 1980.

g dations for the identification, design, installation, and maintenance of certain instrumentation considered Type 5. Section 3 of IEEE Std 603 1980 lists additional c-w A in Regulatory Guide 1.97. standards that are referenced in the standard. Those referenced standards not endorsed by a regulatory guide

3. In Regulatory Position 3, Section 6.3.l(1) of IEEE or incorporated into the regulations contain valuable Std 603-1980 has been changed to currect a printing information; if used, they should M used in a manner ,

error in the standard. consistent with current regulatory practice.

4< Figure 7 of IEEE St i 603 1980, which provides an D. IMPLEMENTATION interpretation of Section ,6.3.1, is confusing and could be misleading in that the upper .left " diamond" cannot The purpose of this section is to provide information accommodate an event that, by itself, results in a to applicants and licensees regarding the NRC staff's condition requiring a safety function while simultaneous- plans for using this regulatory guide.

ly causing action by a non-safety system. A modified chart is included as Figure I of this guide as stated in Except in those cases in which an applicant or Regulatory Position 4 Licensee proposes an acceptable alternative method for complying with specified portions of the Commission's

' f Additional IEEE standards that are referenced in regulations, the method described in this guide will be j other sections of the standard are listed in Section 3 of used by the NRC staff in its evaJuation of the design, i IEEE Std 603-1980. Because the NRC staff may not reliability, qualification, and testability of the power, have endorsed these other standards, a caution regs: ding instrumentation, and control portions of safety related their use is provided in Regulatory Position 5. tystems for construction permit applications docketed after November 1985. All other applications will be l C. REGULATORY POSITION evaluated against the provisions of this guide only to 1 the extent that the licensee elects to use the guide as a l The requirements contained in IEEE Std 603-1980 basis for system modifications requiring staff approval.

provide a method acceptable to the NRC staff for *The term power includes electric, pneumatic, and hydraunc complying with the Commission's regulations with regard power.

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Regulatory Analysis for Proposed Guide on Qualification of Safety-Related 1.ead-Storage Ratteries

Background:

The Commission's regulations (10CFR Part 50) require that safety-related systems and components in nuclear power plants be designed to accommodate the effects of environmental conditions (i.e., remain functional durina and after ,

postulated seismic and accident conditions) and that design control measures,  :

such as testing, be used to verify the adequacy of design. Section 50.44 to )

10CFR Part 50 and Regulatory Guide 1.89 provide requirements and acceptable  ;

methods for the environmental qualification of electric equipment.

The environmental qualification of electric equipment in mild environments is not included within the scope of Section 50.49 However, qualification for the mild environment (and the harsh environment) is included in the scope of IEEE Std. 323-1974 which is endorsed by R.G. 1.80, but component specific quidance is not provided. During the development of Section 50.49, the Comission concluded that the quality and surveillance requirements applicable to electric equipment (10 CFR Part 50, Appendix B; R.G.1.33 Revision 2) are generally sufficient to ensure adequate performance of safety related equipment located in mild environments.

During this rulemaking, the staff was directed to develop regulatory guides for specific electric equipment located in mild environments where additional guidance is needed. '

Discussion:

Safety-related batteries are located in mild environments. Section 3.11 (Page 5) '

of the Standard Review Plan states the followino regarding qualification for electric equipment in mild environment:

"The environmental qualification of all electrical and mechanical equipmer.t located in the mild environment is acceptable if the following procedure is followed:

The documentation required to demonstate qualification of equipment in a mild environment are the " Design / Purchase" specifications. The specifications shall contain a description of the functional requirements for its specific environmental zone during nomal and abnomal environmental conditions. A well supported maintenance / surveillance program in conjunction with a good preventive maintenance program will suffice to assure that equipment that meets the design / purchase specifica-tions is qualified for the designed life.

Furthermore the maintenance / surveillance program data and records shall be reviewed periodically (not more than 18 monthsl to ensure that the design qualified life has not suffered thermal and cyclic degradation resulting from the accumulated stresses triggered by the i

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abnormal environmental conditions and the normal wear due to its  ;

service condition. Engineering judgment shall be used to modify the replacement program arm, eplace the equipment as deemed neces sa ry. "

Seismic qualification of safety-related batteries is covered by Regulatory Guide 1.100 There is no regulatory guide for regulatory document) that describes methods acceptable to the NRC staff for " pre-aging" prior to seismic qualifica- <

tion of safety-related batteries. However, Section 3.11 (page a) of the Standard Review plan, dated July 1981, states: "In addition, IEEE Standards 381, 535...  !

can be used for guidance purposes even though NRC has not formally endorsed these standards through the issuance of a Regulatory Guide." i The staff has accepted voluntary compliance with IEEE Std 535 (all versions) for .

meeting the Commission's regulations. '

present Licensing Status:

Seismic qualification of safety-related equipment, including batteries, has been evaluated by NRR against Regulatory Guide 1.100 for the past several years. During seistic qualification reviews, the NRR staff has noted that some kind of pre-aging was done prior to seismic testing of batteries.

Staff now plans to endorse IEEE Std 535-1986 by a regulatory guide., No backfit is involved; only future plants (or operating plant licensees voluntarily committing to the RG) are affected. The IEEE Standard represents a national consensus on qualification methods to assure the reliability, availability and functionality of batteries used in nuclear power plants.

Although, strictly speaking, the RG does impose a new staff position on new plants, the position imposed essentially codifies existing good practices by the industry and is consistent with the more general guidance provided by the

' Standard Resiew Plan.

Analysis of Technical Sections of IEEE Std 535-1986 4

1. Section 5.0 of IEEE Std 535-1986 specified three methods of qualification:

5.1 Type testing 5.2 Operating experience  :

5.3 Analysis supported by test data, operating experience or physical laws of nature.

These methods, except for (5.3) which permits operating experience or laws of nature as supporting bases, are essentially identical to Section 50.49(f) for equipment located in harsh environments. They are currently acceptable methods of qualification and, as such, do not impose any new requirements.

2. Section 5.4 of IEEE Std 535-1986 describes several methods of extending

" qualified life." These methods are consistent with IEEE Std 323-1974, which is endorsed by RG 1.89.

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3. Section 5 of IEEE Std 535-1986 specifies the information which should be included in the test plan for qualification of the battery, e.g.,

installation details, preventive maintenance schedule, electrical data, and qualified life. This information is readily available from the licensee (s).

4 Section 7 of IEEE Std 535-1986 specifies requirements for the prototype testing of batteries. These are: (1) Test' Plan, (2) Test Sequence and (3)

Acceptance Criteria. These are the basic requirements for any successful test program for any equipment.

5. Section 8 of IEEE Std 535-1986 specifies prescriptive requirements for type tests and analysis procedures. These include accelerated aging procedure, capacity test and discharge test. These tests are described in j IEEE Std 450, endorsed by R.G.1.129, and are routinely done in the field (although not for qualification purposes).
6. Section 8.3 of IEEE Std 535-1986 deals with seismic testing which is covered by R.G. 1.100. These are prescriptive requirements but do not go beyond the general requirements imposed by R.G.1.100.
7. Section 9 of IEEE Std 535-1086 deals with documentation. These requirements are consistent with R.G.1.89 and Appendix R of 10 CFR Part 50, and, although somewhat more prescriptive, impose no requirements beyond those implicit in the Commission's regulations.

Impact:

Since IEEE Std 535 was first published in 1979, the pre-aging prior to seismic testing was not done systematically and strictly in accordance with IEEE Std 535-1970 All three U.S. manufacturers, Exide, CAD and Gould, have since qualified their batteries in accordance with IEEE Std 535-1979 for use in nuclear power plants. The cost to each company of testing and pre-aging has ranged from $30,000 to $180,000 - depending uoon the number of prototypes tested, when they were tested, and whether the batteries were aged naturally or artificially. There is no significant difference between the 1970 and 1986 versions of IEEE Std 535 in respect to the " pre-aging" part of the qualification testing.

Conclusion:

Based upon our in-depth review of IEEE Std 535-1986, we conclude that the qualification requirements described in the IEEE standard are based on current industry practices. Further, issuance of this Guide will not result in any immediate additional cost, since all U.S. manufacturers have already cualified their current battery designs.

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LWe further conclude the requirements of IEEE 535-1986 will satisfy the

. Commission's regulations with respect to the qualification of safety-related L batteries located in mild environments, swa Satish K. Aggarwal sign Program Manager Engineering' Aranch, RES J//&f6'l F. Rosa Branch Chief-PAEI, NRR-l

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safety-rela ted system. Additionally, in tipplying the criteria of the standard, it is helpful to understand that following:

the following are considered to be synonymous: (1) I 1. The term " safety system" used throughout IEEE electric portions of the safety system, (2) Class IE ' Std 6031980 should be understood 'to be synonymous equipment, and (3) safety-related electric equipment as with " safety related system," in which the term " safety-

. defined in 5 50.49 of 10 CFR Part 50. It should also related" has the meaning stated in i 50.49(b)(1) of 10 be noted that the scope of the standard is broader than ; CFR Part 50. --

-(l), (2), and (3) above since, for example, pneumatic

instruments may be part of the safety system. 2. For displays for manually controlled actions covered in Section 5.8.1 of IEEE Std 6031980, the The fouowing is a brief discussion of the basis of provisions for Type A instruments in Regulatory Guide each regulatory position
1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and . Environs Conditions
1. The terms " safety system" and " safety-related sys- During and Following an Accident," should be followed tem" have evolved separately, and it is essential in apply- instead of IEEE Std 497-1977.

ing IEEE Std 603-1980 that the relationship of the terms (as stated in Regulatory Position 1) be understood. 3. Instead of the first sentence in Section 6.3.!(1) of IEEE Std 603-1980, the first sentence in Section 4.7.4.1

2. IEEE Std 497-1977 is referer.ced in IEEE Std of IEEE Std 279 should be used: " Alternate channels, 603-1980, but some of the requirements of IEEE Std not subject to failure resulting from the same single 497-1977 are at variance with current regulatory prac- event, shau be provided to limit the consequences of tice. Hence, in Regulatory Position 2, reference to IEEE th% event to a value specified by the design bases."

Std 4971977 in IEEE Std 603-1980 is being replaced with reference to Regulatory Guide 1.97, " Instruments- 4. Section 6.3.1 of IEEE Std 603-1980 references tion for Light Water Cooled Nuclear Power Plants To Figure 7, a decision chart for applying the requirements Assess Plant and Environs Conditions During and Fol- of the section. Figure 1 of this guide should be used lowing an Accident," which provides specific recommen- in lieu of Figure 7 of IEEE Std 603-1980.

g dations for the identification, design, instauation, and maintenance of certain instrumentation considered Type 5. Section 3 of IEEE Std 603 1980 lists additional

-w A in Regulatory Guide 1.97. standards that are referenced i'i the stac lard. Those referenced standards not endorsed by a regulatory guide

3. In Regulatory Position 3, Section 6.3.!(1) of IEEE or incorporated into the regulations contain valuable Std 6031980 has been changed to correct a printing information; if used, they should be used in a manner error in the standard. consistent with current regulatory practice.
4. Figure 7 of IEEE Std 603-1980, which provides an D. IMPLEMENTATION interpretation of Section 6.3,1, is confusing and could be misleading in that the upper left " diamond" cannot The purpose of this section is to provide information accommodate an event that, by itself, results in a to applicants and licensees regarding the NRC staff's condition requiring a safety function while simultaneous- plans for using this regulatory guide, ly castsing action by a non-safety system. A modified chart is included as Figure 1 of this guide as stated in Except in those cases in w hich an applicant or Regulatory Position 4. Licensee proposes an acceptable alternative method for complying with specified portions of the Commission's
5. Additional IEEE standards that are referenced in regulations, the method described in this guide will be other sections of the standard are listed in Section 3 of used by the NRC staff in its evaJuation of the design, I IEEE Std 603-1980. Because the NRC staff may not reliability, qualification, and testability of the power, have endorsed these other standards, a caution regarding instrumentation, and control portions of safety-related their use is provided in Regulatory Position 5. systems for construction permit applications docketed ,

after November 1985. AU other applications will be C. REGULATORY POSITION evaluated against the provisions of this guide only to the extent that the licmee elects to use the guide as a The requirements contained in IEEE Std 603-1980 basis for system modifications requiring staff approval.

provide a method acceptable to the NRC staff for .The term power includes electric pneumanc, and hydraulic complying with the Commission's regulations with regard power.

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Enclosure 3 to the Minutes of CRGk Meeting No. 113 CRGR Review of a Proposed Generic Letter Concerning Containment Integrated Leak Rate Tests

,, L. Shao (NRR) and J. Pulsipher (NRR) presented to the CRGR a proposed generic

! letter for the purpose of informing the regulated industry of the current requirements regarding containment integrated leak rate testing. The current requirements are embodied in Appendix J to 10 CFR 50, which requires that containment leak rate testing shall be done in accordance with the provisions of ANSI N45.4-1972, " Leakage-Rate Testing of Containment Structures for Nuclear Reactors." Section 5.1 of ANSI N45.4 permits two methods of performing con-tainment integrated leak rate tests (CILRTs), the absolute method and the reference-vessel method. Further, the standard specifies two methods for leakage rate computations permitted for the 24-hour period absolute method CILRTs. The permitted methods are termed the point-to point and the total time calculations. .Hence, compliance with the regulatory requirements of Appendix J for absolute method CILRTs can only be achieved through the use of the point-to point or total time. calculations, corrected for instrument error.

Applicants and licensees have for several years used a newer test computation method, the mass point calculation method described in ANSI /ANS-56.8-1981,

" Containment System Leakage Testing Requirements." The NRC staff believes that the mass point method described in the newer ANSI standard has technical merit.

A staff proposed change to Appendix J, which was issued for public comment (51 FR 39538,10-29-86), endorses the mass point method, among other proposed changes. However, until such a rule change.would be formally approved and issued, compliance with the existing Appendix J requires a CILRT leakage rate calculated by total time or the point-to point method. CILRTs are done by licensees approximately every three years, and in several instances NRC staff has approved licensee test results and conclusions that were accomplished using the mass point method. While staff believes that such test results and conclusions were, in technical substance, generally more accurate and less costly than tests done using the total time or point-to point methods, past decisions to accept mass point calculations were legally in error, unless formal exemptions to the regulations were granted. Accordingly, staff's proposed generic letter was intended to rectify this situation.

The proposed generic letter would advise recipients of the approved inter-pretation of the current Appendix J and the necessity to conform to it in '

future CILRTs, and would also state that tests previously accepted by the staff would not need to be redone or results recalculated.

The staff presentation to the CRGR included a handout (Attachment to this enclosure) which covers the historical development of the situation, experience reported by the regions in dealing with this matter on individual cases, region views on what should be done, a review of instances where NRC approved use of the mass point method, and consideration of some options other than the pro-posed generic letter. In two cases formal exemptions to the regulations were granted, and in three cases changes to plant Technical Specifications were approved. As the discussion proceeded, CRGR members quickly came to the con-clusion that the purpose of the generic letter was correct and warranted, but

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i r that the letter was an undesirable vehicle with which to effect the needed

change in staff requirements. The proposed' generic letter would simply continue the current regulatory status which requires that a licensee either request a plant-specific exemption or decline to use a method' acknowledged by both NRC and industry to be a technically superior method.

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RECOMMENDATION.TO THE EDO The CRGR observed that the fundamental problem was a legal one and that a prompt legal remedy should be sought. J. Scinto suggested that the 14 years of staff practice supporting the mass point method should make possible an ED0 decision to clarify a long-held de facto staff position and to make that.

decision a matter of regulation. The CRGR agreed that'NRR should, in con-junction with OGC, prepare a document that'would result in an immediately effective rule change to permit the mass point method. That document, along-with an analysis addressing the effects of the change on safety and cost,'and other matters as normally required for CRGR review, should be presented to the >

.. 'CRGR for review.

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I OBJECTIVE OF THIS MEETING l

'- REVIEW PROPOSED GENEPIC LETTER AND APPPOVE FDP ISSUANCE HISTORY

-' OGC MEMORAraxe, JULY 16, 39P6, SAID MASS PnINT MEm0D NOT ALLnWED BY REGULATION SOMF PLANTS WERE USING MASS POINT MEm0D TECHNICAL STAFF HAD APPROVED MASS Pn!NT FOR 8 YEARS OR MORE

- SCHEDULE FOR APPROVAL OF PROPOSED APPENDIX J PEVISinN IS UNCERTAIN; TARGET IS SPRIT!G 1988

- REGIONS HAVE BEEN IMPLEMENTING OGC RULING SINCE ISSUANCE, WITH FEW PROBLEMS GENERIC LETTER INFORMS UTILITIES OF THE REGULATIOP REQUIREMENTS

- STATES f%SS POINT OPTION WILL NOT COMPLY WITH REQUIREMENTS; OTHERWISE MAKES NO CHANGES

- CLAP.IFICATION TO PUBLIC NEEDED; UNCERTAINTY PRESENTLY EXISTS

- PAST TESTS, ALREADY COMPLETED, WILL NOT BE REDONE OR RECALCULATED WILL NO LONGER BE NEEDED WHEN (AND IF) REVISED RULE BEC0fES FINAL

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t' REGIONAL FXPERIENCE SINCF OGC MEMn i 5 ALL PLANTS NOW AWARE OF RULING 1

- ALL PLAN'S ARE COMPLYING, EXCEPT:

FITZPATRICK CONSIDERS MASS POINT TO BE THEIR "0FFICIAL" METHOD (ALLOWED BY WEIR T.S.), BUT DID PASS USING TOTAL TIME MEm0D AT LAST TEST (APPIL 1987),

- OYSTER CREEK USED Mass POINT FOP OCTOBER 1986 TEST; IDENTIFIED AS UNRESOLVED' INSPECTION ITEM' YANKEE (ROWE) USED MASS POINT IN 1984; MAY TRY TO USE IT FOR NEXT TEST-OCONEE RECENTLY WAS GRANTED AN EXEMPTION TO ALLOW USE OF MASS POINT

- H. B, ROBINSON TEST APPIL 1987 FAILED BY TOTAL TIME, PASSED BY MASS POINT; WILL REQUEST QUICK EXEMPTION TO ENABLE MASS POINT; FIRST FAILURE LNDER OGC RULING c

i_ _ _________ _ _ _ _ _ _

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I- REGIONAL VIFws

- REGIONS WOULD PREFER A GENERIC LETTER OR INFORMATION NOTICE; WOULD BE "CLEANEP," MORE DEFINITE AGENCY POSITION; HCHEVER, NOT URGENTLY NEEDED

- ' INSPECTION & BlFORCEMENT MANUAL, PART M , PEVISFD DECEMBER.19P,6,._.

l ALREADY CONTAlt1S GUIDANCE TO INSPECTORS

- REGIONS DO NOT INTEND TO BACKFIT POSITION TO OLDEP, COMPLETED TESTS

. (BEFOPE OGC MEMO)

REGION l'HAS ENCOURAGED LICENSEES TO. SEEK EXEMPTIONS TO ALLOW USE OF MASS P0ltn e

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' EXPLICIT NRC ACCEPTANCE OF MASS POINT METHOD

- PLANTS KNOWN TO HAVE TECH SPECS ALLOWING MASS Poltrr-ftGUIRE 1 #1D 2 CATAWBA 3 FITZPATRICK PLANT KNOWN TO HAVE SER ACCEPTING Mass POINT COMANCHE PFAK (SSER l?)l PLANTS WITH FORMAL EXEMPTIONS TO ALLOW USE OF MASS Poltrr

- OCONEE 1, 2, 3 (GRANTED EARLY 1987)

- H. B. ROBINSON ?. (REQUESTED APRll 1987; REVIEW PENDING) t

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I OPTIONS OTHFP THAN GENERIC LETTER

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,7 NRR MFNO TO REGIONS, WOULD SOLIDIFY ENFORCEMENT POSITION SOMEWHAT

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WOULD NOT GO DIRECTLY TO PUBLIC m . - - LIMITED RULE CHANGE WOULD FULLY RESOLVE "0LD TESTS" ISSUE AND TECHNICAL N'D LEGAL CONCERNS ADMINISTRATIVE CONFLICTS WIDi PENDING GENERAL REVISION OF APPENDIX J WOULD MAKE GENERAL REVISION HARDFR TO GET THPOUGH TO FINAL ISSUNCE

. DO NOTHING 7 REGIONS HAVE ISSUE WELL IN HMO

- HOWEVER, SO E CONFUSION mfd UNCERTAINTY WOULD REMAlf! FOP PUBLIC M O REGIONS we f

_ _ _ . _ _ . _ _ _ . . _ _ . . - _ _ . _ _ _ . _ . _ - _ _ _ m_ _ _ _ _ __________m____ ____m. _

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. PROPOSED ADDITIONAL WORDS FOR GFNERIC LETTER, DISPOSITION OF "0LD TESTS" THE BASES FOR CONTINUING TO ACCEPT PAST TESTS THAT HAD PASSED USING THE MASS POINT METHOD ARE THAT, ALTHOUGH THESE WERE NOT STPICTLY IN COMPLIANCE:

1.. A CEPTAIN MEASURE OF NRC REGIONAL DISCRETIONARY ENFORCEMENT IS ALLOVED AND PPOPER WHERE TFCHNICAL ADEQUACY IS NOT REDUCED.

2. THE MASS POINT METHOD IS GENEPALLY RECOGNIZED AS BEING TECHNICALLY EQUIVALENT TO OR MORE ACCUPATE THAN THE CALCULATIONAL MFTHODS PEFERENCED IN ANSI N45.4-1972.
3. SUCH PLANTS WOULD BE IN COMPLIANCE WITH THE CURRENT LEGAL POSITION BY THE NEXT TEST.
4. PUBLIC SAFETY IS NOT MATERIALLY AFFECTED BY WHETHER A PLANT PASSED ITS LEAKAGE PATE TEST BASED ON PAST OR CURRENT TECHNICALLY ACCEPTABLE CALCULATIONAL E THODS. ,

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