ML20199E244

Affidavit of as Masciantonio in Response to Board 860313 Hearing Question Re Guidance on Relationships Between Qualification Test Sample,Production Equipment & Margin Applied During Qualification Tests.Reg Guide 1.89 Encl

ML20199E244
Person / Time
Site:	Vogtle
Issue date:	03/20/1986
From:	Masciantonio A Office of Nuclear Reactor Regulation
To:
Shared Package
ML20199E242	List: ML20199E237 ML20199E244
References
RTR-REGGD-01.089, RTR-REGGD-1.089 NUDOCS 8603250452
Download: ML20199E244 (8)

Text

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k UNITED STATES OF AMERICA -

NUCLEAR REGULATORY COMMISSION BEFORE THE ATOPUC SAFETY AND LICENSING BOARD In the Matter of )

.)

GEORGIA POWER COMPANY ) Docket Nos. 50-424

--et al. ) 50-425

) (OL)

(Vogtle Electric Generating Plant, )

Units 1 and 2) )

i AFFIDAVIT OF ARMANDO S. MASCIANTONIO IN RESPONSE TO LICENSING BOARD HEARING QUESTIONS ON CONTENTION 10.5 (ASCO SOLENOID VALUES)

I, Armando S. Masciantonio, first being duly sworn, state the following:

1. I am presently employed by the U.S. Nuclear Regulatory Commission as a mechanical engineer in the Engineering Branch of PWR-A Division of Licensing, Office of Nuclear Reactor Regulation (NRR).

Before November 1985, I was employed as an equipment qualification engineer in the Equipment Qualification Branch, Division of Engineering, Office of Nuclear Reactor Regulation. I was responsible for the technical j reviews, analyses and evaluations of the adequacy of the environmental qualification of electric equipment important to safety and safety-related mechanical equipment whose failure under postulated environmental conditions could adversely affect the performance of safety systems in nuclear power plants. My professional qualifications are set forth at page l

2 of my written prefiled testimony on Contention 10.5 (ASCO Solenoid l

l l

I 8603250452 B60320

, PDR ADOCM 05000424 L Q PDR

Valves) which was incorporated into the record as if read on March 13, 1986. Masciantonio, ff. Tr. 550.

At the hearing on Contention .10.5 (ASCO Solenoid Valves) held

~

2.

.on March 13, 1986, I was asket by the Licensing Board (Judge Linenberger) if Regulatory Guide 1.89, Revision 1 contained ~ specific guidance on the relationships '.setween the qualification test sample, the production equipment of the same type and the " margin" applied during qualification tests. Tr. 552-3.

3. Since Regulatory Guide 1.89, Revision 1 was not available to me for review at that time, I was unable to answer the above questio:-*. It was agreed that I would provide the answers at a later time. I have since reviewed Regulatory Guide 1.89, Revision 1 and provide the following answers.

4.. Regulatory Guide 1.89, Revision 1, a copy of which is attached, provides an indirect reference addressing the similarity of the test sample to the production unit. Regulatory Guide 1.89, Revision 1 does not directly state that the qualification test unit be the same as the production units of the same type. The specific guidance on test sample similarity to production units is contained in IEEE Standard 323-1974, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations. " This standard defines " sample equipment" as

" Production equipment tested to obtain data that are valid over a range of-ratings and for specific services." Further, Section 4 of this standard states, "It is the primary role of qualification to assure that for each l-type of Class 1E equipment the design and manufacturing processes are such that there is a high degree of confidence that future equipment of

= -. .- - . - ..-. -- -.. ---.. . - - - ..

4 4 the same type will perform as required. The other steps in the quality ,

assurance .. program require strict control to assure that subsequent equipment of the same type matches that which was qualified and is suitably applied, installed, maintained, and periodically tested. Margins used during type testing (qualification testing) provide additional

, assurance that the equipment will perform as required."

5. Regulatory Guide 1.89, Revision 1, Section C, formally endorses the i procedures described in IEEE Standard 323-1974 as acceptable to the NRC Staff for satisfying the Commission's regulations pertaining to the

qualification of electric equipment for service in nuclear power plants.

Therefore, by its endorsement of IEEE Standard 323-1974, Regulatory Guide 1.89 indirectly requires that the qualification test sample be the

same as the production units of the same type. Similarly, margin is

addressed in Regulatory Guide 1.89, Revision 1, Section C.4 by reference j

to and endorsement of the margin values listed in Section 6.3.1.5 of IEEE j Standard 323-1974. These margins serve to account for variations in commercial production and inaccuracies in the test equipment.

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'S

Armando S. Masciantonio Subscribed and Sworn to before j me this gday of March,1986

!: 42 lYf Notary Public I

My Commission Expires: ///pg

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9 Revision 18

[  % U.O. NUCLEAR REEULATORY CEMMISSIZN June 19s4 GU DE I

@\g)* OFFICE OF NUCLEAR REGU

, REGULATORY GUIDE 1.89 (Task EE 042-2) l ENVIRONMENTAL QUALIFICATION OF CERTAIN ELECTRIC EQUIPMENT IMPORTANT TO SAFETY FOR NUCLEAR POWER PLANTS A. INTRODUCTION Section 50.49 does not include requirements for seismic and dynamic qualification, protection of electric The Commission's regulations in 10 CFR Part 50, equipment against other natural phenomena and external

" Domestic Licensing of Production and Utilization events, and equipment located in a mild environrnent.

Facilities," require that structures, systems, and com-ponents important to safety in a nuclest power plant This regulatory guide describes a method acceptable be designed to accommodate the effects of environ- to the NRC staff for complying with { 50.49 of mental conditions (i.e., remain functional under postu- 10 CFR Part 50 with regard to qualification of electric lated accident conditions) and that design control equipment important to safety for service in nuclear measures such as testing be used to check the adequacy power plants to ensure that the equipment can perform of design. These general requirements are contained in its safety function during and after a design basis General Design Criteria 1, 2, 4, and 23 of Appendix accident.

A, " General Design Criteria for Nuclear Power Plants,"

to Part 50; in Criterion 111, " Design Control," Criterion The Advisory Committee on Reactor Safeguards has XI, " Test Control," and Criterion XVII, " Quality been consulted concerning this guide and has con-Assurance Records," of Appendix B, "Qualitv Assurance curred in the regulatory position. Any guidance in Criteria for Nuclear Power Plants and Fuel 6.trocessing this document related to information collection activities

_,_ Plants," to Part 50; and in { 50.55a. has been cleared under OMB Clearance No. sl50-0011.

Specific requirements pertaining to qualification of B. DISCUSSION certair. electric equipment important to safety are contained in Q 50.49, " Environmental Qualification of IEEE Std 3231974, "lEEE Standard for Qualifying Electric Equipment important to Safety for Nuclear Class IE Equipment for Nuclear Power Generating Power Plants," of 10 CFR Part 50. Section 50.49 Stations,"3 published February 28, 1974, was prepared requires that three categories of electric equipment by Subcommittee 2, Equipment Qualification, of the important to safety be qualified for their application Nuclear Power Engineering Committee of the Institute and specified performance and provides requirements of Electrical and Electronics Engineers (IEEE) and was for establishing environmental qualification methods approved by the IEEE Standards Board on Decem-and qualification parameters. These three categories are ber 13,1973. The standard describes basic procedures (1) safety-related electric equipment (Class IE), (2) for qualifying Class IE equipment and interfaces that non-safety-related electric equipment (non-Class IE) are to be used in nuclear power plants, including com-whose failure under postulated environmental conditions ponents or equipment of any interface whose failure could prevent satisfactory accomplishment of safety could adversely affect any Class IE equipment.

functions by safety-related equipment, and (3) certain postaccident monitoring equipment. This regulatory For the purposes of this guide, " qualification" is a guide applies only to these three categories of electric verification of design limited to demonstrating that the equipment important to safety. electric equipment is capable of g'erforming its safety 3

Copies may be obtained from the Institute of Dectrical and

'The substantial number of changes in this revision has made Dectronics Ensineers, Inc., 345 East 47th Street, New York, it impractical to indicate the chanses with lines in the marsin. New York 1o017.

USNRC REGULATORY GUIDES Comments should be sent to the Secretary of the Commission.

Regulatory Guldes are issued to describe and make available to the At noe ocke ing an 6c I!r ch.

I ic par s of No C mi s1o s reg I tions, t det ae h The guides are issued in the following ten broad divisions:

ated acc'ident or t pro Ide u ance t op i t egu atory" 3. Power Reactors 6. Products he [s n r utred ethods an olutions dlNerenNro one e ue s and terials F lt ties cupat on 4ealth

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k j , Power Plants," of 10 CFR Part 50 requires that safety- b. Effects of Sprays and Cheadcals The effects of i related ~ electric equipment (Class _ IE) as defined in containment spray system operation should be considered.

paragraph 50.49(b)(1) be qualified to perform its intended This consideration should include, as appropriate, the

[ safety functions. Typical safety-related equipment and effects of demineralized water spray or chemical spray

\ systems are listed in Appendix A to this guide. Paragraph systems.

50.49(b)(2) requires that non-safety-related electric equip-ment be environmentauy qualified if its failure under c. Radiation t'adh-Inside and Outside Contain.

1 postulated environmental conditions could prevent satis- meat. The radiation ewironment for qualification of i

l factory accomplishment of the safety functions by electric equipment should be based on the radiation safety-related equipment. Typical examples of non-safety- environment normally expected over the instaued life of I related electric equipment are included in Appendix B the equipment plus that associated with the most severe i to this guide. Paragraph 50.49(b)(3) requires that certain design basis accident during or fouowing which the l postaccident monitoring equipment also be environmen- equipment must remain functional The accident-related i . tally qualified. These are specified as " Categories I enWronmental conditions should be assumed to occur at i and 2" in Revision 2 of Regulatory Guide 1.97, "Instru- the end of the installed life of the equipment. Methods l mentation for Light Water-Cooled Nuclear Power Plants acceptable to the NRC staff for establishing radiation j to Assess Plant and Epirons Conditions . During and doses for the qualification of equipment for BWRs and

! Following an Accident." PWRs are provided in Appendix D and the following:

, (1) The source term to be used in determining

2. Paragraph 50.49(d) and Section 6.2 of IEEE Std the radiation environment associated with a design basis 323-1974 require equipment specifications to include LOCA should be taken as an instantaneous release to performance and environmental conditions. For the the containment of 100% of the noble gas activity,50%

requirements called for in item (7) of Section 6.2 of ,

of the halogen activity, and 1% of the remaining fission j IEEE 323-1974 and paragraph 50.49(d)(3), the following product activity The fission product solids should be

should be included

assumed to remain in the primary coolant and to be j carried by the coolant to th- containment sump (s).

j a. Temperature and Pressure Conditions Inside  ;

Containment for LOCA and Main Steam Line Break (2) For all other design basis accidents (e.g.,

j (MSLB). The following methods are acceptable to the non-LOCA highenergy line breaks or rod ejection or NRC staff for calculating and establishing the contain- rod drop accidents), the qualification source term ment pressure and temperature emelopes to which calculations should use the percentage of fuel damage I equipment should be qualified: assumed in the plant-specific analysis (provided in the j Final Safety Analysis Report (FSAR)). The nuclide

. (1) Methods for calculating mass and energy imentory of the breached fuel elements should be release rates for LOCAs and MSLBs are referenced in calculated at the end of core life assuming continuous Appendix C to this guide. The calculations should full-power operation. The imentory of the fuel rod gap

, account for the time dependence and spatial distribution should be assumed to be 10% of the total rod activity l of these variables. For example, superheated steam imentory of iodine and 10% of the total activity imen-

! followed by saturated steam may be a limiting condition tory of noble gases (except for krypton-85, for which a l

and should be considered. release of 30% should be assumed). All the gaseous

{ constituents in the gaps of the breached fuel rods I (2) For pressurized water reactors (PWRs) with should be assumed to be instantaneously released to the

a dry containment, calculate LOCA or MSLB contain- primary system. When substantial fuel damage is postu-ment environment using CONTEMPT LT or equivalent lated,100% of the noble gases, 50% of the halogens, industry codes. and 1% of the remaining fission product solids in the affected fuel rods should be assumed to be instantane-(3) For PWRs with an ice condenser contain- ously released to the primary system.

ment, calculate LOCA or MSLB containment emiron-j ment using LOTIC or equivalent industry codes. (3) For a limited number of accident monitoring instrumentation channels with instrument raries that (4) For boiling water reactors (BWRs) with a extend well beyond the values the selected variables can Mark I, II, or III containment, calculate .LOCA or attain under limiting conditions as specified in Regulatory MSLB emironment using CONTEMPT-LT or equivalent Guide 1.97, Revision 2, the environmental qualification industry codes. should be consistent with Regulatory Positions C.I.3.1.s and C.I.3.2.a of Regulatory Guide 1.97, Revision 2.

Since the test profiles included in Appendix A to IEEE Std 323 1974 are only representative, they should (4) The calculation of the radiation environment not be considered an acceptable alternative to using associated with design basis accidents should take into plant specific containment temperature and pressure account the time dependent transport of released fission design profiles unless plant-specific analysis is provided products within various regions of the containment and ,

to verify the applicability of those profiles. auxiliary structures. I

!  !.89-3 l- - - - - - _ . - . _ . - . - - - _ _ -

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function under synincant emironmental stresses resulting E'ectric equipment to be . qualified in a nuclear ,

from design basis accidents !n order to avoid common- radiation environment should be exposed to radiation cause failures. Paragraph 50.49(e)(5) calls for equipment that simulates the calculated integrated dose (normal qualified by test to be preconditioned by natural or and accident) that the equipment must withstand prior artificial (accelerated) aging to its end-ofinstalled-life to completion of its intended safety function. Regulatory f condition and further specifies that consideration must Position C.2.c proposes the use of source terms that are be given to all significsnt types of degradation that ,can consistent with previous guidance in the original edition

- have an effect on the functional capabdity of the of this guide, NUREG-0588, " Interim Staff Position on equipment. There are considerable uncertainties regarding Emironmental Qualification of Safety-Related Electrical the processes and emironmental factors that could result Equipment,"2 and the DOR Guidelines, " Guidelines for in such degradation. Oxygen diffusion, humidity, and Evaluating Epironmental Qualification of Class IE accumulation of deposits are examples of such effects. Electrical Equipment in Operating Reactors."3 Because of these uncertainties, state-of-the-art precondi-tioning techniques are not capable of simulating all Item (8) of Regulatory Position C.2.c addresses mia nincant types of degradation, and natural pre aging is quahfication of equipment exposed to low-level radiation difficult and costly. As the state of the art advances doses. Numerous studies that have compiled radiation and uncertainties are resolved, preconditioning techniques effects data on all classes of organic compounds show may become more effective. Experience suggests that that compounds with the least radiation resistance have consideration should be given, for example, to a combi- damage thresholds greater than IO* rads and would nation of (1) preconditioning of test samples employing remain functional with exposures somewhat above the the Arrhenius theory and (2) surveillance, testing, and threshold value. Thus, for organic materials, radiation maintenance of selected equipment specifically directed qualification may be readily justified by existing test toward detecting those degradation processes that, based data or operating experience for radiation exposures on experience, are not amenable to preconditioning and below 104rads. However, for electronic components, that could result in common-cause functional failure of studies have shown failures in metal oxide semiconductor the equipment during design basis accidents. devices at somewhat lower doses. Therefore, radia-tion qualification for electronic components may have a It is essential that safety-related electric equipment be lower exposure threshold.

qualified to demonstrate that it can perform its safety function under the environmental service conditions in The regulatory positions delineated in this guide which it will be required to function and for the length of reflect the state of the art. Research programs currently ,

time its function is required and that non-safety-related in progress are investigating such concerns as the effects electric equipment covered by paragraph 50.49(b)(2) of oxygen in a LOCA environment, the validity of x ,

be able to withstand environmental stresses caused sequential versus simultaneous applications of steam and by design basis accidents under which its failure could radiation environments, and fission product releases prevent the antisfactory accomplishment of safety func- following accidents. The staff recognizes that the results tions by safety-related equipment. This concept applies of research programs may lead to revisions of the throughout this guide. The specific environment for regulatory positions.

which individual electric equipment must be qualified will depend on the installed location and the conditions C. REGULATORY POSITION under which it is required to perform its safety function.

The procedures described by IEEE Std 323-1974, The following are examples of considerations to be "lEEE Standard for Qualifying Class IE Equipment for taken into account when determining the environment Nuclear Power Generating Stations,"8 are acceptable to for which the equipment is to be qualified: (1) equip. the NRC staff for satisfying the Commission's regulations ment outside containment would generally see a less ~ pertaining to the qualification of electric equipment for severe environment than equipment inside containment; service in nuclear power plants to ensure that the (2) equipment whose location is shielded from a radia- equipment can perform its safety functions subject to tion source would generally receive a smaller radia- the following:

tion dose than equipment at the same distance from the source but exposed to its direct radiation; (3) equip- 1. Section 50.49, " Environmental Qualification of ment required to initiate protective action would generally Electric Equipment important to Safety for Nuclear be required for a shorter period of time than instrumen-tation required to follow the course of an accident; and (4) analyses taking into account arrangements of equip-ment and radiation sources may be necessary to deter-mine whether equipment needed for mitigatior, of design be obtained from the NRC GPO Sales Program,

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basis accidents other than loss-of-coolant accidents (LOCA) or high-energy line breaks (IIELB) could be 3 Available for inspection or copying at the ILS Nuclear exposed to a more severe environment than the LOCA Q3*g'Q,",'"@,n gug n ge,nt o,goong tjaggtgt a

or HELB environments delineated in this guide. Januarr 14,19ao. ,

1.89-2

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e first ten hours of the event. This equipment shouM a. The item of equipment to be replaced is a l

remain functional in the accident environment for a component of equipment that is routinely replaced as period of at least I hour in excess of the time assumed part oi normal equipment maintenance, e.g., gaskets,

( in the accident analysis unless a time margm of less o-rings, coils; these may be replaced with identical than one hour can be justified. This justification must components. ~

include, for each piece of equipment, (1) consideration of a spectrum of breaks, (2) the potential need for the b. The item to be replaced is a component that is

, equipment later in an' event or during recovery opere- part of an item of equipment quahfied as an assembly; j tionr, (3) a determmation that failure of the equipment these may be replaced . with identical components.

i after performance of its safety function will not be detrimental to plant safety ' or mislead the operator, c. Identical ' equipment to be used as a replacement

and (4) a determination that the margin apphed to the was on hand as a part of the utility's stock prior to i

mimmum operaiality time, when combined with the February 22, 1983.

other test margins, will account for the uncertainties associated with the use of analytical techniques in the d. Replacement equipment quahfied in accordance derivation of environmental parameters, the number of with the provisions of 9 50.49 does not exist.

units tested, production tolerances, and test equipment inaccuracies. For all other equipment (e.g., postaccident e. Replacement equipment qualified in accordance

monitoring, recombiners), the 10% time margin identified with the provisions of g 50.49 is not available to meet in Section 6.3.1.5 of IEEE Std 323-1974 shouM be installation and operation schedules. However, in such used. case, the replacement equipment may be used only mitil upgraded equipment can be obtained and an outage of

5. Section 6.3.3, "Asms," of IEEE Std 323-1974 sufficient duration is available for replacement.

and paragraph 50.49(e)(5) should be supplemented with the following: f. Replacement equipment quahfied in accordance with { 50.49 would require manificant plant modifica-

a. If synergistic effects have been identified prior tions to accommodate its use.

to the initiation of qualification, they should be accounted for in the qualification program. Synergistic effects 3. The use of replacement equipment qualified in known at this time are dose rate effects and effects accordance with $ 50.49 has a .i..ir"t probability of

, resulting from the different sequence of applying radia- creating human factor problems that would negatively l tion and (elevated) temperature, rffect plant safety and performance, for example:

i b. The expected operating temperature of the (1) Knowledge, skills, and ability of existing

. equipment under service conditions should be accounted plant staff would require sagruficant upgradmg to operate for in thermal aging. The Arrhenius methodology is or maintain the specific replacement equipment; considered an acceptable method of addressing accelerated j thermal aging within the limitation of state-of the-art (2) The use of the replacement equipment j technology. Other aging methods will be evaluated on a would create a one of.a-kind application; or

case by-case basis.

(3) Maintenance, surveillance, or calibration activ-I

c. The aging acceleration rate and activation ities would be unnecessarily complex.

i energies used during qualification testing and the basis i upon which the rate and activation energy were estab- 7. In addition to the requirements of paragraph i lished should be defined, justified, and documented. 50.49(j) of 10 CFR Part 50 and Section 8, "Documen-I tation," of IEEE Std 323-1974, documentation should

d. Periodic surnillance and testing programs are address the information identified in Appendix E to this acceptable to account for uncertainties regarding age- guide. A record of the qualification should be maintained l related degradation that could affect the functional in an audi42 file to permit verification that each item capability of equipment. Results of such programs will of electric equipment is qualified to perform its safety

, be acceptable as ongoing qualification to modify desig- function ander its postulated environmental conditions nated life (or qualified life) of equipment and should be throughout its installed life.

incorporated into the maintenance and refurbishment /

j replacement schedules. D. HrLEMENTATION i

6. Replacement electric equipment installed subse- The purpose of this se: tion is to provide information quent to February 22, 1983, must be qualified in accor- to applicants and licensees regarding the NRC staff's dance with the provisions of $ 50.49 unless there are plans for using this regulatory guide.

sound reasons to the contrary. The NRC staff considers the following to be sound reasons for the use of replace- Except in those cases in which the applicant or ment equipment previously qualified in accordance with licensee proposes an acceptable alternative method for the DOR Guidelines or NURECrO588 in lieu of upgrading: complying with specified portions of the Commission's i

! 1.89-5

.. o 1

(5) Electnc equipment that could be exposed a. Electric equipment that could be submerged '

to radiation should be environmentaBy quahfied to a should be identified and qualified by testing in a sub.

radiation does that simulates the calculated radiation merged condition to demonstrate operability for the environment (normal and accident) that the equipment duration required. Analytical extrapolation of results for should withstand prior to completion of its required test periods shorter than the required duration should F safety functions. Such quahficataon should consider that be justified, equipment d==aye is a function of total integrated dose

- and can be influenced by does rate, energy spectrum, b. Electric equipment located in an area where and partacle type. The radiation quahfication should rapid pressure changes are postulated simultaneously factor in doses from all potential radiation sources at with the most adverse relative humidity should be '

the equipment location. Plant-specific analysis should be qualified to demonstrate that the equipment seals and used to justify any reductions in dose or dose rate vapor barriers will prevent moisture from penetrating resulting from component location or ahwlding The into the equipment to the degree necessary to maintain qualift-atian environment at the equipment location equipment functionability.

should be established using an analysis smular in nature

. and scope to that included in Appendix D to this guide c. The parameters to which electric equipment is and incorporating appropriate factors pertinent to the being qualified (e.g., temperature, pressure, radiation) by actual plant desigr. (e.g., reactor type, containroent exposure to a simulated environment in a test chamber desagn). should be measured sufficiently close to the equipment i

to ensure that actual test conditions accurately represent (6) Shielded components need be quahfied only the environment characterized by the test, to the samma radiation environment provided it can be demonstrated that the sensitive portions of the compo- d. Performance characteristics that demonstrate the ment or equipment are not exposed to significant beta operability of equipment should '>e verified before, radiation dose rates or that the effects of beta radiation, after, and periodically during testing throughout its including heating and secondary radiation, have no range of required operability. Variables indicative of deleterious effects on component performance. If, after momentary failure that prevent the equipment from l considering the appropriate shielding factors, the total performing its safety function, e.g., momentary opening t

i beta radiation dose contribution to the equipmerit or of a relay contact, should be monitored continuously to component is calculated to be less than 10% of the ensure that momentary failures (if any) have been total samma radiation dose to which the equipment or accounted for during testing. For long-term testing, ,,

component has been quahfied, the equipment or compo- however, monitoring during periodic intervals may be r

nent is considered quahfied for the beta and samma used if justified. (/

. radiation environment.

e. Chemical spray or demineralized water spray (7) Electric equipment located outside contain- that is representative of service conditions should be ment that is exposed to the radiation from a recirculat- incorporated during simulated event testing at pressure ing fluid should be qualified to withstand the radiation and temperature conditions that would occur when the penetrating the containment plus the radiation from the spray systems actuate.

recirculating fluid.

f. Cobalt-60 or cesium-137 would be acceptable (8) Electric equipment that may be exposed to gamma radiation sources for environmental qualification.

Iow-level radiation doses should not generally be consid-ered exempt from radiation qualification testing. Excep- 4. The suggested values in Section 6.3.1.5, " Margin,"

tions may be based on qualification by analysis supported of IEEE Std 323-1974, except time margins, are accept-j' by test data or operating experience that verifies that able for meeting the requirements of paragraph SQ49(e)(8).

3 the dose and dose rates will not degrade the operability Alternatively, quantified margins should be applied to

, of the equipment below acceptable values. the environmental parameters discussed in Regulatory Position C.2 to ensure that the postulated accident

d. Environmental Conditions for Equipment Outside conditions have been enveloped during testing. These Containment. Electric equipmen'. that is "ubjected to the margins should be applied in addition to any conserva-effects of pipe breaks and is required to mitigste the tism applied during the derivation of local environmental consequences of the breaks or to bring the plant to conditions of the equipment unless these conservatisms safe shutdown should be ' qualified for the expected can be quantified and shown to contain appropriate environmental conditions. The techniques to calculate the margins. The margins should account for variations in environmental conditions should employ a plant-specific commercial production of the equipment and the inac-model curacies in the test equipment.

3. Section 6.3, " Type Test Procedures," of IEEE Std Some electric equipment may be required by the 323 1974 should be supplemented with the following: design to perform its safety function only within the h 1.89-4

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