ML19269F000

AEC Rept on Environ Qualification of Safety-Related Electrical Equipment

ML19269F000
Person / Time
Issue date:	02/28/1978
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0413, NUREG-413, NUDOCS 7911140425
Download: ML19269F000 (75)

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NU REG-0413 STAFF REPORT ON THE ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT 2190 271

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llN, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7911140y)

Available from flational Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $6.00, Microfiche $3.00 The price of this document for requesters outside of the fiorth American Continent can be obtained from the National Technical Information Service.

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NU R EG-0413 STAFF REPORT ON THE ENVIRONMENTAL O.UALIFICATION OF SAFETY-RELATED ELECTRICAL EO.UIPMENT 2190 273 Manuscript Comple+ed: December 1977 Date Published: February 1978 Division of Operating Reactors Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

TABLE OF CONTENTS Page Number 1.0 Introduction 1

2.0 Equipment Qualification - Older Plants 3

2.1 Technical Safety Issue 3

2.2 Operating Experience 5

2.2.1 Boiling Water Reactors 6

2.2.2 Pressurized Water Reactors 7

2.3 Qualification of Equipment in Older Plants 9

9 2.3.1 Background 7\\q y0[

2.3.2 Previous Backfitting 10 2.3.3 Connectors and Penetrations 14 2.3.4 Cabling 15 2.4 Other Considerations 18 2.4.1 Timing Considerations 18 2.4.2 Equipt 'nt Response 20 21 2.4.3 IE Inspection Program 2.4.4 Routine Experience Review 22 3.0 Adequacy of Qualification Procram 23 4.0 NRC Confimatory Research Program 25 4.1 Synergistic Testing 25 26 4.2 Aging Effects 4.3 Source Term Equivalences 28 2190 274

TABLr. OF CONTENTS (CONTINUED)

Page Number 5.0 SEP Program 31 5.1 Scope of Technical Review 31 5.2 Extent of Present Program 32 6.0 Conclusions Appendix A Appendix B 2190 275 pp

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ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT 1.0 Introduction The current NRC safety review process for nuclear power plants includes criteria related to the qualification of certain electrical equipment.

These criteria require that electrical equipment important to strety must be qualified to function in the environment that might result from various accident conditions. Although such criteria have been applied since the early days of commercial nuclear power, the details of these criteria have been changed over the years, The evolution of environmental qualification of safety-related electrical equipment is described in Appendix A.

Changes to these criteria have raised some questions as to:

(1) the degree to which electrical equipment used in older plant designs (those operating) is capable of withstanding the environmental conditions (pressure, temperature, humidity, steam, chemicals, vibration, and radiation) of various accident condi-tions under which it must function (i.e., the " qualification of equipment" in these older plants), and (2) the adequacy of tests or analyses conducted for equipment used in newer plants to " qualify" such equipment as capable of with-standing the conditions of the environment created by various accidents dur,ing which the equipment must function (i.e., the U,l\\

" adequacy" of qualification tests).

2190 276 As noted in Appendix A, both of these items are the subjects of ongoing generic programs of the NRR staff. The first subject, the " qualification of equipment review", and the safety of operating plants is the principal subject of this status report.

This subject will be discussed in Section 2.0.

The second item, the " adequacy" of qualification tests, is the subject of two NRR Category A Technical Activities.

An allied concern, whether tests for certain parameters (temperature, prescure, humidity, steam and chemicals) and radiation which are conducted sequentially accurately reflect electrical component performance under accident conditions in which all conditions would be imposed simultaneously, is the subject of an ongoing NRC confirmatory research program. These subjects will be discussed in Sections 3.0 and 4.0.

For both the " qualification" and the " adequacy" issues, it is our conclusion that there is reasonable assurance that public health and safety is adequately protected during the period of time necessary to acnieve their systematic, complete resolution and that no additional immediate action is warranted at this time.

The bases for this conclusion, supporting information, and our established programs for the resolution of these issues are described in later sections of this report.

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3 2.0 Eouipment Oualification - Older Plants 2.1 Technical Safety Concern Despite the conservative design, construction and operating practices and quality assurance measures required for nuclear power piants, safety systems are installed at nuclear olants to mitigate the consequences of postulated accidents.

It is the equipment associated with these safety systems that i's of princi-pal concern with respect to the environmental qualification issue.

Equipment needed for normal operation and equipment needed to respond to plant transients is not of primary concern. This is because past operating experience, especially that from frequent testing requirements, has shown that safety-related equipment can perform in a normal operating envircnment. When problems are encountered during normal operation they are easily identified and corrected. Additionally, the principal safety systems required to function during anticipated transients (those events that are likely to occur during a plant's lifetime) would not be subjected to environmental conditions as severe as those of Dostulated accidents before performing their safety function.

The postulated accidents that have been identified as creating severe environmental conditions inside of containment are breaks of high energy pipes.

The most limiting of these accidents are Y

Y 0?[Q 2190 278 the loss of coolant accident (LOCA) and main steam line break (MSLB).

In each of these cases, hot pressurized water and steam can create a high temperature environment (250 to 400'F) at high humidity (including steam) and pressure (as high as about 50 psig).

For some applications, chemicals are included in sprays that are used to reduce the pressure in the containment. Additionally, some electrical equipment is predicted to be submerged following the postulated accident.

Therefore, for these applications, chemicals and submergence are included in the environmental qualification in addition to the temperature, radiation, humidity and pressure conditions.

While such accidents are judged to have a low like'lihood of occur-rence, less than once per thousand reactor years for smaller pipe breaks, and much less for the large design basis breaks, the NRC has required that safety systems, principally the ECCS and containment isolation and cleanup systems, be environmentally qualified to mitigate these accidents.

For PWRs, the ECCS is often supplemented with additional instrumentation and controls to isolate a steam generator with broken pipes to mitigate the consequences of a postulated MSLB. To assure that these systems would perform their required function, the NRC has required not only redundancy in this equipment, but that it be designed with the capability to perform in the environment associated with such an accident.

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The electrical equipment of concern during postulated accident conditions includes (1) the instrumentation needed to initiate the safety systems and provide diagnostic information to the plant operators (e.g., electrical penetrations into containment, any electrical connectors to cabling which transmits signals, and the instruments themselves), and (2) control power to motor operators for certain valves (e.g., ECCS and containment isolation valves located inside containment). Because of the Comission's requirements for redundancy of electrical equipment and " active" components (e.g., those valves that must change position), nuclear power plants have backup instruments, penetrations, connectors, cables, control devices and valves in addition to the primary safety system components. Therefore, to prevent a safety system from adequately performing its function, a substantial amount of equip-ment, both the primary and backup, must fail' due to environmental conditions.

In some cases the redundant equipment is outside containment and not subject to the same hostile environmental conditions as the primary equipment (e.g. centainment isolation valves).

2.2 Operating Exoerience Several events have occurred at operating nuclear power plants

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which provided a level of independent verification of the environmental qualification of safety related equioment.

Each 2190 280 of these events involved a release of steam to the containment.

Although the radiological releases were negligible, sufficient steam was released to create adverse enviornmental conditions within containment. The most severe event of this type involved an inadvertent discharge of steam through primary system safety valves directly into containment at Dresden Nuclear Power Station.

2.2.1 Boilina Water Reactors On June 5, 1970 an unexpected reactor scram at Dresden Unit 2 was followed by a main steam line safety valve discharge to the containment. About 250,000 pounds of primary coolant were released to containment causing an increase in the primary containment (drywell) pressure to an estimated maximum value of 20 psig at 320*F. Although some equipment damage did occur during the event, the ECCS and containment isolation systems functioned properly.'

On December 8,1971, a similar event occurred at Dresden Unit 3.

In this case the containment reached a maximum pressure of 20 psig and a maximum temperature of 295 F.

Some equipment damage occurred, although much less than at the earlier Dresden linit 2 event. Again, all systems necessary for safety remained operable.

2190 281 un 00iS Steam discharge events of less severe consequences have occurred at several other BWR facilities.

In these occurrences the containment was only slightly pressurized and temperatures remained below 200*F. No damage to any electrical systems was reported for any of these events.

2.2.2 Pressurized Water Reactors Several events have also occurred at PWR facilities which resulted in steam release to containment.

On September 24, 1977, an event occurred at Davis-Besse Unit 1 which resulted in a partial depressurization of the primary system and the release of about 11,000 gallons of water in the form of steam into containment through the pressurizer quench tank rupture disk. The steam release began about 4.5 minutes after the reactor was tripped and was tenninated about 20 minutes after the trip. The only equipment damage occurred within the vicinity of steam discharge.

No equipment damage occurred in safety systems needed to mitigate the event.

On Novemoer 13, 1973, following a reactor trip at Indian Point Unit 2, a break occurred in the feedwater line to a

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steam generator. The break was attributed to water-hamer.

During the event, containment temperature reached a maximum of 110*F and relative humidity reached a maximum of about 50%. No containment pressurization was reported. One cable tray was partially submerged in water. However, the cable within this tray was designed and tested for submersion.

No adverse effects on this cable or on any other electrical equipment in containment were found.

On May 1,1975, while H. B. Robinson Unit _ was in hot shutdown condition, a failure of a reactor coolant pump seal resulted in discharge of about 130,000 gallons of reactor coolant to the containment floor.

The containment was pressurized to a maximum of 3 psig with the maximum temperature estimated to be about 140 to 150 F.

High humidity conditions within containment remained for several hours. No adverse effects on electrical equipment were reported.

Other releases of high or moderate energy coolant into containment have occurred at other PWRs.

However, no pressure rise or equipment damage resulting from an adverse environ-ment were reported.

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2.3 Qualification of Equipment in Older plants 2.3.1 Sackground For older nuclear power plants (those licensed for opera-tion prior to about 1967) specific nuclear environmental qualification requirements had not been established in the industry or by the Commission. Licensing conclusions at the time were based upen overall knowledge of the nature of systems and components used and on the types of accidents of interest, and upon the awareness that nuclear components, including electrica.1 system compon-ents were, and are, of high industrial quality.

At the time, design and purchase specifications for electrical equipment adhered to applicable industry standards, such as the National Electrical Manufacturers Association Standards and existing IEEE Standards (e.g., IEEE Std. 98-1957).

In addition, applicants referenced various testing programs, such as those conducted at the Franklin Institute Research Laboratories and those conducted in the AEC's Naval Reactors program in support of their plant designs.

While these standards are not uniformly comparable to the more specific criteria currently used for nuclear facilities, they nonetheless are of high quality.

2190 284 2.3.2 Previous Backfitting The staff has continued to review the degree to which licensed operating reactors comply with NRC regulations as significant new safety information becomes available or new regulations are established. This effort, generally termed upgrading or "backfitting", has had the effect of increasing the degree to which older operating reactors have improved the documentation of environmental qualification of safety systems. Generally, this is accomplished by staff discussions convincing licensees of the desirability of plant upgrading. However, the NRC has directed licensees to upgrade systems in the past.

In one case, that of Dresden Unit 1, on June 23, 1976, the staff ordered that the reactor protection system be upgraded to meet IEEE Std. 279-1968 and that other safety related equipment not previously environmentally qualified be qualified or replaced. Dresden Unit 1 is the oldest U. S.

operating reactor (licensed in 1959) and for that reason documentation of conformance to current licensing requirements including 10 CFR Part 50, Appendix A (GDC 4) and Appendix B which were issued in 1971 is lacking.

A second, more extensive effort, involved the review of operating reactors to the new ECCS requirements issued An UV!E 2190 285 in 1974 (10 CFR Part 50.46 and Appendix K).

In conducting these reviews the staff audited the degree to which certain operating reactors met requirements for environ-mental qualification of ECCS equipment.

For those plants reviewed in detail, certain areas were discovered as not meeting licensing criteria for equipment qualification.

Most of these defici;:ncies involved those plants licensed prior to the introduction of the General Design Criteria (Appendix A to 10 CFR Part 50). These plants were licensed before systems to mitigate the consequences of a LOCA were requi' J.

As would be expected, some of the safety-related equipment in these older facilities was found not be qualified for LOCA environmental conditions.

Environment 91 qualification can be established by either suitable analyses or tests on equioment to substantiate that the equipment can withstand the postulated environ-mental conditions. Licensees were required to upgrade existing systems nd components in their ECCS in order to ensure reliable performance if subjected to adverse LOCA environmental conditions.

Several examoles of these staff reviews are discussed below.

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\\8\\ u?[s A case in point is the lacrosse boiling water reactor which was licensed in 1967.

Extensive modifications were required to satisfy the ECCS requirements. An integral part of the staff review was the environmental qualification of safety related equipment. Solenoid operated valves in the ECCS were provided with uninterruptible power sources, the solenoids were housed in water free junction boxes connected by sealed mineral insulated cables.

Environmentally qualified valves were added in series to existing valves where qualification had not been established. Analyses were performed and tests were conducted to verify that electrical equipment inside contain-ment would operate in an accident environment.

Containment electrical penetrations were reviewed and found to be caoable of withstanding temperature, pressure, radiation, chemical attack, and submergence in excess of that which is postulated for the Design Basis Event.

In May 1976, Consumers Power Company was granted, by Order from the NRC, an exemption to a portion of the FCCS criteria set forth in 10 CFR 50.46 Appendix K for Big Rock Point (licensed in 1962). The order set forth several conditions which had to be satisfied prior to the Big Rock Point facility's return to operation.

Condition 2.(iii) stated tnat prior to further operation: of Big Rock Point, Consumers Power Company shall:

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2190 287 Protect the controls, indication and annunciation circuitry associated with the ECC. including the core : pray valves, against the conseqt ances of flooding following a LOCA which affects the ability of the ECCS or plant operator to take corrective action during the course of a LOCA.

During the course of the exemption review the staff and licensee identified other areas where environmental qualifica-tion including submergence was suspect.

Corrective measures (relocation of equipment, augmented procedures, etc.) pro-posed for envi: onmental qualification were found acceptable.

San Onofre (licensed in 1967) installed a new onsite power system in order to comply with the ECCS requirements. This modifica-tion included diesel generators, circuitry, and associated switch gear. All of which were designed and implemented in accordance with the IEEE-279 standard and are required to be environmentally qualified.

In addition to these major modifications, in 1974 San Onofre proposed an environmental qualification program which has been implemented and approved by the NRC staff.

ECCS reviews presently ongoing at San Onofre involve again the environmental qualifications of equipment related to safety.

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.. g 2.3.3 Connectors and Penetrations As a result of the recent staff surveys of environmental qualifications for electrical connectors and penetration assemblies, the staff has concluded that operating plants are safe but questions related to documentation remain.

In the

" case of electrical connectors, as part of the staff's followup of experimental results, a staff survey showed that there were several cases in which a licensee asserted that compon-ents were qualified but adequate documentation of qualifica-tion was unavailable.

Inonlyonecase1I was there an absence of both documentation and other evidence of capability to perform in the LOCA environmental conditions.

As a result of certain operating experience with electrical pene-trations, the staff conducted a survey of all licensees since all plants have these components.

In this case, all operating facilities could provide some assurance that their penetrations had the capability to perform in the LOCA environment, although several did not have adequate documentation of qualifica-tion. Our experience in following through with these instances of questionable qualification indicates that ll In the one facility (D. C. Cook Unit 1) where connectors were used without adequate documentation of qualification, there was a tamparary period during which the licensee could not assure itself and the NRC that there was reasonable basis for concluding that the syste:nsgcould nevertheless safely function. Because of this (incertainEy, and after discussions with the NRC, the licensee agreed to shut down the facility while connectors could be reolaced with qualified splices.

2190 289 there are sufficient alternatives to assure safety while the qualifications of the components are established.

It is noteworthy that, although these surveys addressed only two types of components that need to function in accident environments, the survey results do not indicate that unqualified components are in widespread use in operating reactors.

2.3.4 Cabling Wire and cable manufacturers have been aware for many years of the potentially harsh and extreme environments to which electri-cal circuits may be exposed.

Cable has been developed and classified not only on the basis of its electrical capabilities but also for its mechanical integrity, i.e., physical strength and ability to survive extreme temperature, moisture, steam, harsh chemicals, and to varying degrees, fires. Of course the adequacy of cabling for a particular environment is predicated upon the selection of cable with the properties required to survive that environment. When cable is manufactured, not all

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2190 290 electrical and mechanical attributes can be practically incorporated into one cable design.

For example, the very best in electrical properties may not allow the ultimate in mechanical strength or fire resistant properties to be attained.

Compromises are made and designs include the important design features and properties to ensure a high reliability electrical insulation resistance system for a given installation.

IEEE Std. 98-1957 outlines the preparation of test procedures for detennining experimentally the probable life of insulating materials and systems. Test procedures for specific materials and systems incorporating insulating materials are given in a number of other older and updated versions of IEEE publications.

Insulation life test procedures are a part of ASTM (American Society for Testing Materials) and NEMA (National Electrical Manufac-turers Association) standards and proposed procedures and constitute a frequent and continuing part of current technical literature.

2190 291 ot Military and commercial nuclear applications of cabling have required the additional consideration of the effects of exposure to radiation as well as other environmental conditions These environmental effects were studied on insulating materials, as well as insulating systems which are a combina-tion of insulation materials used in the manufacturing of cable, during the early portion of the research and develop-ment phase of nuclear energy applications, both mil.itary and commerci al.

Several professional societier, some of which are mentioned above, participated in the studies and developed industry standards for guidance on radiation effects.

IEEE Std. 279-1968, which was begun in 1964, contained ind,stry thinking that type-test data or reasonable engineerin extrapolation based on type test data should be available to verify that safety related components or equip-ment would perform their function during an accident condi-tion in nuclear power plant designs. These early standards provided documentation of what was existing industry practice and provide reasonable assurance that the cable used in the older nucle ir power plants was carefully specified and pro-cured for che service conditions to which it would be exposed.

One purpo's'e of the Systematic Evaluation Program, discussed in Section 5, is to determine whether cabling built to these V

earlier standards provides sufficient safety margin or whether i

upgrading, discussed in Section 2.3.2, is needed.

2190 292 2.4 Other Considerations 2.4.1 Timing Considerations In considering the potential effects of severe LOCA conditions on performance of electrical systems, it is important to consider the equipment in two separate classes - that which must function essentially instan-tanrously to transmit a protective signal in the event of an ac.cident and tnat which must function for an extended time.

For those systems and components whose primary purpose is to promptly transmit a signal in the event of an accident, - there is a high likelihood that they will successfully perform their function well before the onset of cr.lironmental conditions (temperature, pressure, etc.) which could cause deterioration of the equipment and, thereby, interfere with their function. Typically such signals are transmitted in less than a few seconds, often in fractions of a second after the event (e.g., scram signals which actuate control rods and permit their inser-tion).

It is unlikely that exposure to accidental environmental conditions would cause equipment deteriora-tion in such a limited period of time.

- These include such items as the reactor control rods, sensors, that indicate reactor parameters such as flow rate and neutron flux, associated cables, connectors and penetrations.

2190 293 sn U?K In any event, once a signal is generated (even if only by the environmental conditions themselves), the initiation of the remaining safety-related equipment is accomplished and the signal typically is " sealed in" and maintained even if the equipment providing the actuation signal were to sub-sequently fail.

Once the safety systems are initiated, few components located inside the containment are required to function.

Such components include:

valves that change position, usually within the first minute, to align the flow paths of the safety systems; isolation valves that quickly close to seal the containment from leakage; and the few instrument signals and valves that would normally be required to diagnose the type of accident and institute long tenn actions (usually required several hours after an accident).

Given the requirements for redundance for all " active" components in safety-related systems and their quick acting nature (fractions of a second to several tens of seconds), we would expect that the adverse environmental conditions would not prevent the essential functioning of the safety-related equipment.

However, because this a potential common cause for

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Il (; f equipme'nt'fa'ilbre,('the staff requires environmental qualifica-tion of safety equipment.

In olace of equipment that 2190 294 would need to function as long as several hours after an accident, other equipment can often be ust.d to perform many of the same functions.

2.4.2 Equipment Resoonse Analyses of postulated design basis accidents inside the reactor containment have identified transient conditions in the containment atmosphere that may be severe but are often of short duration.

Generally, the licensing approach of the NRC requires that a conservative detennination of containment atmosphere parameters (e.g., temperature) be calculated and that these conservative parameters then be used as the basis for qualification of electrical equipment located inside containment. Historically, equipment was qualified to the conditions predicted to exist during and subsequent to a large LOCA since they were beliaved to be limiting. As discussed in Appendix A, the staff has detennined that a postulated main steam line break (MSLB) in PWR type plants with dry containments could result in predicted temperatures higher than that of a LOCA, but only for a short 2190 295 i r

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period of time (i.e., 60 to 100 seconds).

In order to better understand the themal response of selected typical electrical equipment inside containment, the staff has performed "best estimate" evaluations of typical components to determine the surface temperature of such components for a particular calculated temperature in the containment that might result from a MSLB. This evaluation 1/ ndicates that because of the short duration of the i

predicted MSLB environmental conditions that exceed typical predicted LOCA conditions, the themal response of the type of equipment in question would generally not exceed the conditions under which the equipment was originally qualified.

2.4.3 IE Insoection Program Another program in the NRC that provides additional confidence in the environmental qualification of electrical equipment, is the inspection program of the Office of Inspection and Enforcement. The Office of Inspection and Enforcement in its routine inspection program has emphasized review of environmental qualification test results for engineered safety systems. The emphasis has been placed primarily on the larger components, such as motors, switchgear, breakers, controls, transmitters, cables, and emergency diesels.

These activities hate been carried out by review of documentation at the licensee's facilities and by inspectors

, accompany;ing the licensee at inspection of vendor facilities or 1 Semorandum to R. Boyd, V. Stello and R. Mattson from R. Tedesco dated December 2, 1977.

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testir.g laboratories. These IE practices, add to the with engineered safety systems will perform their required functions in the accident environments for which they were designed.

2.4.4 Routine Experience Review As the number of nuclear power facilities licensed to operate continues to grow (67 at present), the amount of operating experience at these facilities also increases.

The NRC has a thorough mechanism for the reporting of operating experience. Over the past year there have been about 3000 Licensee Event Reports (LERs) submitted to the NRC. These LERs are routinely reviewed by IE and the Division of Operating Reactors. A complete file is maintained by the Office of Managen ant Information and Prograr. Control (0MIPC).

The NRC's review of LERs he,lps to identify, first, whether any electrical equipment is degrading under normal operation and secondly, whether operational transients and occurrences have degraded performance of electrical equipment under these conditions.

2190 297 on Ovis 3.0 Adequacy of Qualification Tests The second issue identified in Section 1.0 relates to the " adequacy" of present environmental qualification testing. The question is basically one of whether qualification tests that have been and are being performed are adequate to demonstrate that safety-related electrical equipment will perform in accident environments. One aspect of this question is whether successive exposure to certaia parameters (temperature, humidity, pressure, etc.) and radiation is adequate to reflect performance of electrical components under accident conditions in which the exposure to these conditions is concurrent. The present industry standard, IEEE Std 323-1974 and its predecessors are based on the assumption that sequential testing is adequate. At the present time, the staff believes that the successful qualification of electrical equipment for these conditions, albeit sequentially, nonetheless provides a reasonableassurance;Eji that,such equipment will be able to perform i;

e successfully under combined accident conditions. However, there is an absence of rigorous testing under concurrent exposure conditions, and while it is likely that sequentially tested components will successfully perform under accident conditions, the staff believes it prudent to confirm this iudgment. The confirmatory research efforts in this regard are discussed in Section 4.0 2190 298 NRR has a Technical Activities Program involving generic technical activities judged by the staff to warrant priority attention in terms of resources to attain early resolution. These are designated as Category A Technical Activities.

The Technical Activities Program was developed to provide a basic framework of policy, organizational structure, priority, and procedures for the effective management of the major technical activities within NRR.

Two activities in this program are directly related to the environ-mental qualification of safety-related mechanical and electrical equipment. These activities, designated A-21 and A-24, are related to main steam line break considerations inside containment and to qualification of safety-related equipment, respectively.

These efforts are described in Appendix A.

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4.0 NRC Confimatory Research Pr3 grams The current NRC Qualificati;n Testing Evaluation Program is directed toward providing a confirmatory assessment of current environmental qualification testing procedures for LOCA conditions and includes the following specific program elements:

1.

An assessment to determine if sequential (as opposed to simultaneous) environmental qualification testing. is conservative, i.e., an investigation of synergistic effects.

2.

Confirmation that accelerated aging methodology can be utilized.

for qualification testing of safety-related equipment.

3.

Definition of the nuclear radiation source based on the Regulatory Guide 1.89 accident assumptions and evaluation of the adequacy of radiation simulators.

4.1 Synergistic Testing The tests were to confirm that the sequential. test sequence recommended in IEEE Std 323-1974 conservatively simulates the combined radiation and steam environment to which safety-related equipment would be exposed in the unlikely event of a LOCA. A research program to investigate potential synergistic effects was initiated at Sandia Laboratories in FY 1975. LOCA qualification tests are being conducted using the same experimental test chamber and identical test samples (1) sequentially

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4 2190 200 as recomended in IEEE-Std-323-7", with radiation exposure preceding exposure to steam and chemical environments and (2) simultaneously with radiation, steam, and chemical environments imposed together.

A qualitative comparison of the performance of the test specimens in both tests will be made, using on-line measurements and post-test evaluation.

Preliminary evaluation of the Sandia tests completed to date does not indicate a significant functional synergism for electrical cables; however, with respect to connectors, it was not possible to determine whether synergism exists because of the failures that occurred.

4.2 Aging Ef fects Considerations of aging in environmental qualification test programs are important cecause of the potential to create a weakened condition in a safety-related component through some aging mechanism that may not be detected through routine periodic testing.

A research program to develop a methodology that can be utilized for simulation of the natural aging process 2190 301

!:0 09ti of safety-related mate' ials on an accelerated basis was initiated in FY 1976 and is continuing. The current effort includes the following elements:

a.

Single environment aging tests on polymeric electric cable materials are being conducted to obtain data from the separate effects of radiation, temperature and humidity.

From these tests, single environment acceleration functiors of damage versus time will be obtained at relatively 'aw stress levels using a test cycle of about one year.

Cable elongation is being used as a relative damage indicator to verify the aging methodology and for conparison with naturally aged cable sampies.

b.

Combined environment aging tests will be conducted to obtain data on the synergistic aging effect of temperature and radiation.

Synergisms with other aging parameters are planned later in the

program, c.

Tests to determine rate effects are underway.

Of particular interest are the rate effects associated with oxygen diffusion and radiation.

d.

A study of alternate damage indicators is undenvay that could be utilized in addition to the material elongation criterion which is cc 2ntly being used as the reference damage indicator for aging damage to electrical cable.

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2190 302 e.

Naturally aged samples are being collected so that the aging methodology being developed can be checked with naturally aged material.

4.3 Source Term Equivalences One of the exceptions taken to IEEE Std-323-1974 by the NRC is the required radiation environment to be utilized for environmental qualification testing.

LOCA radiation releases are defined in Regulatory Guide 1.89 which is to be used by an applicant in establishing the nuclear radiation environment for type testi19 The accident conditions are defined in terms of the percentage of halogens and solid fission products contained in the coolant, the percentage of noble gases and halogens released to the containment atmosphere and the percentage of halogens plated out on surfaces inside containme't.

The adequacy of currently used radiation simulators to duplicate the accident radiation environment requires additional experimental evaluation. A research program to assist in this evaluation was initiated in FY 1976 and is continuing.

Progress to date has consisted of analysis to determine the time relationship following a LOCA of dose, dose rate, energy spectra and particle type. These data show that current industry practice with regard to radiation 2190 303 J

simulation testing may be significantly different in terms of dose rate, spectrum and particle type than that described in Regulatory Guide 1.89.

The ongoing work in this area is aimed at determining the importance of these differences in terms of damage to safety related equipment.

The current effort consists of the following three tasks:

a.

Additional source term calculations will be made,' based on Regulatory Guide 1.89 assumptions, taking into account new codes and test data develcrgd in other programs. Also, calculations based on proposed regulatory guide modifications will be performed. These calculations will be based on the proposed revision to Regulatory Guide 1.89 which allows for reduced release assumptions for certain classes of safety-related equipment.

In addition, source term calculations will be made with best estimate LOCA release assumptions as required.

b.

An evaluation is being made of the adequacy of currently utilized radiation simulators to duplicate the hypothetical environment following the radioactive release postulated in Regulatory Guide 1.89. An initial assessment will be made comparing dose rates and energy spectra resulting from the conservati~ve accident assumptions in Regulatory Guide 1.89 to the dose rates and energy spectra obtainable with practical simula ters.

2190 304 c.

Studies will be conducted to determine the damage to safety-related equipment materials as a function of the gamma and beta dose-rates and to determine how close the dose-rate prc, files resulting from the Regulatory Guide 1.89 assumptions must be simulated during qualification testing. Materials studies will be conducted utilizing radiation damage data available, and additional experimental data will be obtained as needed.

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5.0 SEP Program 5.1 Scope of Technical Review Very recently NRR embarked on a program to re-evaluate selected safety considerations for eleven older operating facilities.

That program called the Systematic Evaluation Program is described in a recently issued NRC report.*

One of the topics included in this program, which is directed at a determination and documentation of the degree to which these older facilities meet current licensing requirements, concerns the environmental qualification of safety-related equipment.

The object.ive of the SEP's review of this topic is to evaluate the degree to which the mechanical and Class IE electrical equipment of safety-related systems has been qualified for the environments associated with design basis events.

As such, the SEP will be directed toicard the determination of existing safety margins and the evaluation of the adequacy of such safety margins te determine if any backfitting or facility upgrading is necessary.

Because of recent operational occurrences at the Milestone plant, and in view of the results of the recent surveys regarding connec-tors and penetrations, this review topic will be completed

• Report Iod the " Systematic Evaluation of Operating Reactors", dated November 25, 1977.

2190 j06 as the first topic of the SEP. While the overall Systematic Evaluation Program is scheduled to be completed in about three years, the review of this topic will be accelerated.

It is expected that within about 90 days the review effort will be sufficient to assess any safety implications in sufficient detail to decide whether or not additional review of facilities other than those included in the SEP is required'. The adequacy of the environmental qualification of mechanical equipment will follow that of the electrical equipment.

The review plan Tor this effort for the eleven SEP facilities is set forth in Appendix B.

5.2 Extent of Present Program The present Systematic Evaluation Program includes the review of eleven of the older operating nuclear reactors.

These eleven include piants licensed before 1969 and those facilities which require a review for conversion of a Provisional Operating License (POL) to a Full-Term Operating License (FTOL).

Following the 90-day review of the environmental qualification of electrical equipment for these eleven older facilities, the staff will determine whether any plant modifications or followup actions are required for those facilities and will also consider 2 M0 507 o,

whether the environmental qualification review should be extended to include the remainder of the operating facili ties.

We have concluded that the eleven older facilities can be used as a basis to make such a decision because as noted in Appendix A, they represent a grouping of plants that would likely have a lesser degree of environmental qualification for the safety-related electrical equipment than more recent plant designs.

2190.508

(

n v, r e.

6.0 Conclusions The NRC stcff has concluded that no immediate action Commission is needed on the question of environmental qualification of safety-related electri-cal equipment in operating reactors is warranted.

Beyond the question of immediate action, the staff has considered whether the recently comoleced preliminary surveys regarding electrical connectors and containment electrical penetrations in operating plants should be expanded to consider, on a longer term basis, the safety adequacy and environmental. qualification of other electrical equipment in these plants. The Systematic Evaluation Program for Operating Reactors recently approved by the Comission provides a suitable framework for such an expanded effort since it already includes

" Environmental Qualfication of Safety-Related Equipment" as one of the topics considered. The SEP further provides that topics considered to be of special safety significance may be considM ed on a case-by-cace basis in advance of completing the overall program.

The staff has determined that it is appropriate to complete the review of this subject as the first topic of the Systematic Evaluation Prcgram.

It is expected that within about 90 days the review effort wi11 be sufficient to assess any safety implication 3 in sufficient detail to decide whether or not additional review of facilities otherthan those included in the SEP is required.

2190 $M The results of the detailed staff review of these topics for these facilities, the eleven of the older reactors, will indicate whether further action is needed on the other operating reacters.

In reaching the judgment that no innediate action is required on operating reactors, the staff, as discussed elsewhere in this repcrt, considered the following:

1.

Nuclear power plants include provisions, such as redundancy and diversity, to cope with equipment failures without af fecting the public health and safety.

2.

Operating experience indicates that electrical equipment has performed adequately under both normal operating environmental conditions and on the few occasions where severe environmental conditions have existed.

3.

Even the older operating reactors used conservative design and construction practices and many improvements have been made in the area of environmental qualification.

4.

A preliminary audit of the environmental qualification of electrical connectors and penetrations in operating reactcrs has indicated that there is reasonable assurance that this equipment would perform its safety function under accident conditions even thot.;h conolete documentation is not readily avdilable in all cases.

It is the staff's belief that these findirfgs would be' essentially the same for other safety-related equipment.

2190 a10 The likelihood that emntial safety-related equipment or other non-safety equipmer.t woul1 not perform the necessary safety function prior to failure due to environmental reasonsO coupled with the likelihood of a major accident requiring the performance of this equipment is very low.

6.

The regulations have included requirements for environmental qualification and a comprehensive quality assurance program since 1971. The requirement for environmental cualification was included in initial versions of these regulations in the mid 1960s. The NRC compliance effort by the Office of Inspection and Enforcement has emphasized review of environmental quali-fication test results for safety systems in its routine inspection program.

4/ It should be noted that even in the Sandia tests, under the conditions of Reg. Guide 1.89, whica envelope DBA conditions and are thus conservative, particularly as to radiation and steam temperature conditions, a number of unqualified connectors, survived for periods in excess of several hour::.

2190 311

,t, no,p, s >-

Appendix A REPORT ON THE HISTORICAL EVOLUTION OF ENVIRONMENTAL QUALIFICATION REQUIREMENTS FOR SAFETY-RELATED ELECTRICAL EQUIPMENT 2190

'>l2

TABLE OF CONTENTS PAGE ABSTRACT A-iii

1.0 INTRODUCTION

A-1 2.0 EVOLUTION OF OVERALL NRC LICENSING A-1 CRITERIA 3.3 EVOLUTION OF NRC ENVIRONMENTAL QUALIFICA-A-5 TION REQUIREMENTS 3.1 EVOLUTION OF APPLICABLE INDUSTRY A-5 STANDARDS 3.2 EVOLUTION OF NRC (ONRR) LICENSING A-10 REQUIREMENTS 3.3 EVOLUTICJ OF OI&E INSPECTION PRACTICES A-18 4.0 CURRENT ACTIVITIES A-20 4.1 SYSTEMATIC EVALUATION PROGRAM A-21 4.2 NRR CATEGORY A TECHNICAL ACTIVITIES A-22 4.3 CONFIRMATORY RESEARCH PROGRAMS A-24 4.4 STANDARDS DEVELOPMENT PROGRAMS A-27

5.0 REFERENCES

A-28 LISTING OF TABLES A-29 Table 1 - Development of Standards for Class IE Equipment Table 2 - Facility Groupings by the Evolutionary Stage of NRC Licensing Requirements for Environmental Qualification of Safety-Related Electrical Equipment Table 3 - Development of AEC/NRC Requirements ana/or Guidance for the Enviromental Qualification of Safety-Related Electrical Equipment 2190 313 A-i

A-iii ABSTRACT Since the early days of the commercial nuclear power industry, the NRC (formerly AEC) criteria for the licensing of nuclear facilities have undergone an evolutionary process. This document traces the develop-ment of Commission and industry requirements for environmental qualifications of safety-related electrical equipment from the late 1950s to the present time.

It also addresses the expanding role of related Inspection and Enforcement activities over this ceriod of time.

2190 314

A-1

1.0 INTRODUCTION

The purpose of this report is to describe the evolution of NRC licensing requirements for the environmental qualification of safety-related electrical equipment. The scope of tha report has been 1imited to criteria which relate to normal, abnormal, accident, and post-accident environmental conditions, i.e.,

temperature, pressure, relative humidity, steam, radiation, che.r.icals, and vibration.

As has been the case with virtually all NRC licensing criteria, the licensing criteria for the environmental qualification of safety-related electrical equipment have evolved over the years as the design of reactoc systems has changed and as regulatory and operating experience have accumulated.

The evolution of NRC licensing criteria for environne qtal qualification of safety-related electrical equipment has occurred simultaneously and at essentially the same pace as the evolution of the overall NRC licensing criteria, as summ rized belcw.

2.0 EVOLUTION OF OVERALL NRC LICENSING CRITERIA In the early days of the civilian nuclear power industry, the Commission's licensing review of the acceptability of proposed nuclear plant designs was based on much less documented design information than is presently required by the NRC licensing process.

In addition, as early reactor designs evolved, the 2190 315

A-2 Comission's " standards of acceptability" were established on an ad hoc basis unique to each new licensing review. Many of these

" standards of acceptability" were not formalized; rather they evolved during plant specific licensing reviews, thereby estab-lishing precedents for subsequent reviews. As the number of applications for Construction Permits (cps) submitted to the Commission began to grow, it became evident that more. uniform and consistent guidelines (" standards for acceptability") were necessary.

In 1966, the Commission issued a " Guide to the Organiza-tion and Contents of Safety Analysis Reports."l/ Historically, from t:ds time on, the amount of documentation required by the Commission's licensing process began to significantly increase.

That guide identified the areas of NRC staff safety concern and identified the degree of detailed information and analyses required from applicants to permit the staff to complete its review.

In a further effort to provide guidance to the industry and to increase staff review efficiency and effectiveness, the Commis-sion began issuing Safety Guides in 1970.

These guides present methods acceptable to the Commission for implementing specific parts of the Regulations, including the General Design Criteria of 10 CFR Part 50 Appendix A.

In 1971, The Commission beoan issuing Information ruides to list needed infonnation which was frequently omitted from applications.

In 1972 the Safety and Information Guides were replaced by the broader based NRC Regulatory Guide program which h

U ? !'i 2190 316

A-3 continucs icd,y. Regulatory Guides are not substitutes for Regulations and compliance with them is not in itself a legal requirement. Methods and solutions different from those set forth in the Guides may be, and have been found to be, acceptable by the NRC staff.

The next major improvement in guidance to applicants was provided in a document entitled " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants",2f initially issued in 1972 and subsequently revised in 1975. These documents reflect the increased scope and detail of infomation required to sunport license applications.

In a similar way, the licensing criteria and requirements used by the NRC staff to determine acceptability have evolved ov!r the years.

In 1965, the Commission published, for public comment, 27 proposed General Design Criteria (GDC) for nuclear power pl an ts. Those criteria established minimum requirements for the principal design criteria for commerical nuclear plants.

These GDC were refined and formally adopted in the Regulations, as Appendix A to 10 CFR Part 50 in 1971 in the forn of 55 General Design Criteria. These GDC have been used by the Commission as guidance in reviewing plant applications since they were originally drafted.

The GDC were specifically w.. tten in general terms such that a variety of alternative design techniques may be utilized to satisfy them.

As the knowledge of reactor designs increased and operating exper-ience accumulated, additional licensing reouirements were also developed.

In an effort to document such requirements and thus E i l, M 15, 2190 317

A-4 increase regulatory effectiveness, efficiency and predictability, the Commission developed and published the Standard Review Plan (SRP) in 1975.3_/ That document provides guidance for staff re-viewers to improve the quality and uniformity of staff reviews.

It also improves communication and understanding of the staff review pro-cess with interested members of the public and the nuclear power industry and helps to standardize the licensing process.

In general, the detailed acceptance criteria published in the SRP did not represent new licensing requirements; rather, they reflected current staff review practices and standards of acceptability which had evolved during previous licensing reviews. However, in many cases, these criteria had not previously been published in any regulatory document. The SRP will be periodically revised to incorporate new or modified requirements as they are developed and approved.

Because of the continual evolution of reactor designs and asso-ciated licensing requirements, operating nuclear power plants that were reviewed and approved in the past have a broad spectrun of design characteristics. Each of these reactors was fou'i to be in confonnance with the licensing " requirements" in effect at the time of licensing. As noted above, however, these requirements have become more detailed over the years. Conse-quently, if some older licensed facilities were reevaluated using current licensing procedures, they would likely be at variance in some respects; the older the plant the more at variance it is likely to be.

Although such variances may not necessarily 2190 318

A-5 represent significant safety deficiencies, the current Syste-matic Evaluation Program (SEP), which is described in Section 4.1 of this report, is directed toward the determination of existing safety margins at older coerating reactors and the evaluation of such safety margins to detennine if any backfitting or facility upgrading for safety is necessary.

3.0 EVOLUTION OF NRC ENVIRONMENTAL QUALIFICATION RE0VIREMENTS 3.1 Evolution of Acolicable National Standards Over the past ten years, many national Standards have been prepared to describe the methods commonly used to demonstrate the environmental qualification of safety-related electrical equipmca.t utilized in nuclear power generating stations.

These Standards reflect the elements of good engineering practice which have evolved in the development of reactor system designs and regulatory licensing requirements.

The NRC staff has, for many years, participated with representatives of industry in the development of these Standards and, after independent review, has often incorporated these Standards into its Regula-tions and Regulatory Guides with appropriate supplemental material.

Each of the pertinent Standards is described in the following sections of this report.

A graphical presentatie i of the

^-~

sequence of the development of these Standards is provided in Table 1.

As c.oted previously, hcwever, these Standards,as endorsed by NRC degulatory e,

2190 319

A-6 Guides, are not substitutes for Regulations. Consequently, compliance with these Standards (with the exception of IEEE Standard 279 which is incorporated by reference in 10 CFR Part 50, 550.551) is not a legal requirement.

During the course of the licensing review for a particular facility, methods or solutions different from those set forth in these Standards may be f]und to be acceptable by the staff, but usually on some basis of comparability with the provisions of the Standards.

3.1.1 Section 4.4 of IEEE Standarc 279-1968, " Proposed IEEE Criteria for Nuclear Power Plant Protection Systems,"

and its revision dated 1971, require that either type test data or reasonable extrapolations based on test data be available to demonstrate the environmental qualification of protection systen equicment at nuclear power plants. This Standard has been incorporated as part of the Commission's regulations by reference in 10 CFR Part 50.55a, " Codes and Standards".

More detailed qualification guidance for electrical equipment has been developed in several later IEEE Standards and, where appropriate, these Standards have been endorsed by NRC Regulatory Guides (sonetimes with supplementary material) as acceptable methods for qualifying electric equipment in general or specific kinds of electric equipment.

vit U?iS 2190 320

A-7 3.1.2 IEEE Standard 323-1971

" General Guide for Qualifying Class I Electrical EoJipment for Nuclear Power Generating Stations," and its revision dated 1974, describe the basic require-ments for qualifying Class IE equipment and interfaces that are to be used in nuclear power generating stations, in partial support of the requirements of GDC 4 and 21 (Appendix A to 10 CFR Part 50) and Sections 4.4 and 4.5 of IEEr Standard 279-1971. The 1974 Revision of this Standard included criteria which establish requirements for Qualification procedures, methods,and documentation and incorporated new or improved guidelines related to aging, testing margins, and the sequence for testing of different environmental parameters. This Standard recognizes that environnental qualification of safety-related electrical equipment can be accomolished by several different methods (e.g., type testing, operating experience, analysis) utilized separately or in combination.

Furthennore, this Standard rer',qnizes that, while fulfillment of its requirements does not rzeessarily fully establish the adequacy of the environ-mental gh:.lification of electrical equipment, omission of any of its requirements will, in most instances, be an indication of inadequate qualification.

2190 c21 nr

A-8 IEEE 323-1974, has been endorsed by NRC Regulatory Guide 1.89.

The Standard and Regulatory Guide 1.89 have been written such that they can be followed by successive ancillary Standards and Guides which reference the parent documents for common qualification techniaues and specify the additional requirements pertinent to a specific component.

For example, IEEE 334-1974 (motor qualification) is ancillary to IEEE 323-1974 Respective endorsement guides have the same relationship. A listing of applicable qualifica-tion Standards, including ancillary Standards, follows:

a.

IEEE Standard 383-1974, " Type Test of Class IE Elec-tric Cables, Field Splices, and Connections for Nuclear Power Generating Stations," provides direction for establishing type tests which may be used in qualifying Class IE electric cables, field splices, and connections for service in nuclear power generating nations in conjunction with of the general guidelines for qualification which are given in IEEE Std. 323-1974 and in GDC 3.

Regulatory Guide 1.131, endorsing this ancillary IEEE Standard, was issued for public comment in August 1977.

b.

IEEE Standard 317-1971, " Standard for Electrical Penetra-1

{J p l C

' tion Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations," and its revisions dated 2190.527

A-9 1972 and 1976, provide guidance for qualifying electrical penet ations and include testing requirements. The 1976 version of this Standard includes additional design and testing requirements and is ancillary to the 1974 version of IEEE Standard 323.

Revision 1 to NRC Regulatory Guide 1.63, in turn, endorses this Standard.

c.

IEEE Standard 334-1971, " Type Tests of Continuous Duty Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations," and its revision dated 1974, specify acceptable methods for qualifying electric motors.

The 1974 version of the St idard re-flects the 1974 update of IEEE Standard 323 and will be endorsed by Revision 1 to NRC Regulatory Guide 1.40.

Regulatory 'suide 1.40 endorsed IEEE Standard 334-1971.

d.

IEEE Standard 382-1972, " Type Tests of Class I Electric Valve Operators for Nuclear Power Generating Stations" saeci-fies acceptable methods for qualifying electric valve operators. NRC Regulatory Guide 1.73, in turn, endorses this Standard. (This is not an ancillary Standard).

e.

IEEE St ndard 381-1977, " Type Tests of Class IE Modules Used in Nuclear Power Generating Stations," is an ancillary Standard which specifies acceptable methods for qualifying electric modules.

An NRC Regulatory Guide endorsing this Standard is being considered for development.

W 005 2190

.323

A-10 3.2 Evolution of NRC (ONRR) Licensing Recuirements The NRC's licensing requirements for the environmental qualification of safety-related electrical equipment have evolved over the years as the ( tign of reactor systems has changed, as operating and regulatory experience have been accum-ulated, and as testing facility capabilities and testing techniques have expanded and improved.

In a very general sense, the evolution of these NRC licensing requirements can be characterized by three stages:

the evaluation of facilities licensed prior to 1967, facilities licensed af ter 1967 up to facilities with Construction Permit (CP) applications tendered prior to July 1974, and facilities with CP applications tendered after July 1974. A listing of the facilities which fall within each of these three groupings is provided in Table 2.

A graphical presentation of the sequence of development of licensing require-ments and/or guidance for the environmental qualification of safety-related electrical equipment is provided in Table 3.

For all plants in the first two of the above-mentioned three groupings, the licensing review included an evaluation of the environmental qualification of safety-related electrical equipment located inside containment. As noted, the scope and depth of such reviews have increased with time.

In addition, equipment environ-mental qualification has been considered by the staff for subsequent plant design modifications proposed by licensees of these facilities and for modifications required by changes in the Regulations; e.g.,

modifications associated with the demonstration of compliance with

, Appendix K of 10 CFR Part 50.

2190 324

A-11 The staff review of plants falling within each of the three above-mentioned groupings is characterized in the following subsections.

3.2.1 PRIOR TO 1967 For facilities licensed prior to 1967, the information initially submitted to the regulatory staff for review of the environmental qualification of safety-related electrical equipment included, as a minimum, design specifications and demonstration of the adherence of such design specifications to appropriate industry standards, such as the National Electrical Manufacturers Association Standards, existing IEEE Standards which were not specifically developed for application by the nuclear industry (e.g., IEEE Standard 117-1956, "IEEE Standard Test Procedures for Evalua-tion of Systems of Insulating Materials for Random-Wound AC Electrical Machinery"), and other IEEE ;tandards under development.

In addition, acLicants referenced various environmental testing programs such as those conducted at the Franklin Institute Research Laboratories and those conducted in tne Naval Reactors Program.

3.2.2 1967 - 1974 The staff's review of the second grouping of facilities utilized the initial criteria of IEEE Standard 279-1968 and IEEE Standard 323-1971.

These criteria, and others that became available subseauenti have been utilized, as they became available, for subsequent e /alua-tions.

It should be noted that several of the test programs to dem-onstrate the qualification of safety-related electrical equip-ment [were initiated during the development of the criteria 2190 325

A-12 established in IEEE Standard 323-1971 prior to its issuance.

In some cases during the course of its reviews, the staff determined that the test plan procedures or other available documentation for particular facilities did not meet all of the IEEE 323-1971 guidelines.

In such cases, the staff re-quired confimatory programs to be implemented, additional analyses to be performed, and/or additional documentation related to previous environmental testing programs to be supplied. Consideration of the plant specific design and site conditions were included in the staff's evaluation of the adecuacy of equipment to perform its safety-related functions during nomal, abnomal, accident, and post accident environmental conditions.

These evaluations were perfomed on a selected component basis, i.e., the review treated com-ponents judged by the staff to be representative of all safety-related components in the facility.

As indicated in Section 3.1, the IEEE Star;dards continued to evolve during this time frame and, consequently, the licensing reviews of the plants that preceded the implemen-tation of IEEE Standard 323-1971 and other Standards that because availability subsequently had varying degrees of qualification, testing, analysis, and associated docunentation. The evolution of the NRC and AEC Regulatory process and the Standards deveicoment process have markedly increased the anount of documentation and the scoce and depth of the roviews cerfomed by the staff since 1971.

2190 326 he 0?ll

A-13 Appendix X to the " Reactor Safety Study"4/ provides a detailed summary of the environmental qualification of safety-related equipment installed in four of the facili-ties within this grouping, i.e., Surry Units 1 and 2 and Peach Bottom Units 2 and 3.

3.2.3 cps Tendered After July 1974 The staff's. reviews of the environmental qualification of safety-related electrical equipment for plants tendering cps after July 1974 reflect the more comprehensive guidelines specified in IEEE Standard 323-1974 and the successive ancillary Standards. As discussed in Section 4.2.2 of this report, the staff is currently reviewing the concepts, methods, test procedures, and acceptance criteria proposed by the major NSSS vendors and Balance-of-Plant equipment suppliers to meet the guidelines for environ-mental qualification of safety-related electrical eauf p-ment provided in IEEE Standard 323-1974 and its ancillary Standards.

The staff has not completed its review of these programs.

The methods and procedures which are reviewed for facilities within this grouping are documented in Section 3.11 of the Standard Review Plan (SRP) which was issued in September 1975.

The Standard Review Plan is written so as to cover a variety of site conditions and plant designs.

For any given apoli-dS G b!S 2190 327

A-14 cation, the staff selects and emphasizes particular aspects of the Standard Review Plan as is appropriate for that application.

In some cases, a plant feature may be suffi-ciently similar to that of a previous plant so that a detailed re-review is not needed. Therefore, the staff has not and does not expect to perform in detail all of the review steps listed in the Standard Review Plan for the review of each application. This approach is typical for most other review areas.

The SRP-type environmental qualification review includes an information audit review to detemine if the following informa-tion is included in the application:

a.

All safety-related mechanical and electrical ecuipment must be identified.

The equipment tabulations provided should be checked for completeness against the descrip-tions of safety-related systems. Definitions of the three categories of safety related systems are contained in Section 7.1 of the SRP.

b.

The location of each item of safety-related eauipment, both inside and outside the containment, must be iden-ti fi ed.

Location of the equipment is required in order to establish accurate definitions of both the normal, abnomal, and accident environments.

2190 328 ix

A-15 c.

Both the nomal, abnomal, and accident environmental con-ditions must be explicitly defined for each item of equip-ment.

These definitions must include the following parameters:

temperature, pressure, relative humidity, steam, radia-tion, chemicals, and vibration. For the nomal environ-ment, specific values should be provided. For the abnomal and accident environments these parameters should be presented as functions of time for the particular cause of the postulated environment, (e.g., pipe break, or other).

d.

The length of time that each item of equipment is required to operate in an abnormal or accident environment must be provided.

e.

The qualification report should contain a complete description of the design bases and environmental qualification tests and/or analyses that have been perfomed on each item of safety-related equipment.

This should include qualification for the accident environments, qualification for extreme normal opera-ting environments, and qualification to assure that loss of environmental control systems that are not classified as safety-related will not adversely affect the operability of safety-related equipment, particularly electrical equipment located in the control room or other roons housing control equipment.

2190.:29

A-16 The staff review involves an evaluation of the completeness and adequacy of the information presented to assure that an adequate demonstration of the required environ-mental capabilities of safety-related equipment has been provided. This phase of the review is performed after it has been established (by means of the information audit phase of the review previously described) that the infoma-tion content requirements for Section 3.11 of the Standard Review Plan have been satisfied. An essential part of this evaluation is the formulation of cuestions by the licen-sing reviewer and responses by apr

ants to document their qualificatici programs. Typical of these requests for additional information are:

a.

A statement that the staff recuires that the fol-lowing qualification test program infomation be provided for a specified list of Class IE equipment:

(1) equipment design specification requirements, (2) test plan, (3) test set up, (4) test procedures, (5) acceptability goals and requirements, and (6) test resul ts.

b.

A requirement that the applicant damonstrate that the sequence of the environmentSl conditions which were imposed during the qualification testing is at least as severe as the actual environment sequence during a postulated accident.

6 u?ls 2190 a30

A-17 c.

A requirement that the applicant provide supporting analyses, operating experience, or other information that demonstrates the adequacy of specific equipment to perform its required safety function in nomal, abnormal, accident, and post-accident environments.

The Regulations, current IEEE Standards, and the NRC Regulatory Guides which are identified in Section 3.11 in the Standard Review Plan are used as guidelines and as the acceptance criteria for the staff reviews to provide assurance that the equipment can perform its safety-related function. The staff reviews the proposed concepts, methods, and test procedures that will be utilized to demonstrate compliance with the current criteria and evaluates the results of analyses, tests, experience, or other methods (including combinations of the above) for acceptability when they are submitted at the FSAR stage of the licensing process.

3.2.4 Other Considerations Further evidence of the evolutionary nature of the licensing considerations for environmental qualification of safety-related electrical equipment has been the recent recoonition W MK 2190 33l

A-18 that the environment associated with a postulated main stean line break (MSLB) accident in PWR facilities may, in some respects, be more deiaanding on electrical equipment installed inside containment than the environmental conditions associated with the design basis loss of coolant accident (LOCA).

Prior to 1976, the accident environment against which safety-related electrical equipment located inside containment was qualified was bounded by the environment produced by the loss of coolant accident. However, in 1976, information became available that indicated that the calculated temperature inside the containment associated with the MSLB accident could be as much as 100 - 150 F higher, for a short time duration (i.e.,60-100 seconds), than that associated with a LOCA. As indicated in Section 4.2.1 of this report, further efforts are presently underway to establish environmental envelope requirements for MSLB accidents inside containment.

3.3 Evolution of OI AE Inspection Practices The NRC Office of Inspection & Enforcement's involvement in the environmental qualification testing of safety-related electrical equipment has evolved in step with the NRC licensing (0NRR) re-6 2190 332

A-19 quirements. During the late 1960's and early 1970's, the Office of Inspection & Enforcement's inspectors perit ":: ally visited equic, ment vendors for the purpose of auditing vendor practices with respect to qualification testing.

Inadequacies identified during such inspections were brought to the attention of appro-priate NRC licensing personnel.

In a number of instances, Generic problems and/or design deficiencies were identified and corrected.

Since the early 1970's the involvement of the Office of Inspection and Enforcement's inspectors in this area has increased substan-tially.

In the 1972-1973 time frame, the current Office of Inspec-tion and Enforcement's licensee contractor and vendor inspection program was initiated.

Included in that program are provisions for the inspection, on a sampling basis, of suppliers of all types of equipment (including electrical and instrumentation conpcnents).

While that program focuses on the review of vendor cuality assurance programs and fabrication activities, witnessing of actual cualifica-tion tests is occasionally performed.

At the reactor construction sites, inspectors review selected qualifi-cation test results for safety-related components and systems.

These reviews are based on specified requirements and/or test conditions.

During the early 1970's inspectors reviewed testing and cuality 2190 333

A-20 control requirements, material certification, and test resul ts.

Since about 1975, the inspection requirements have been expanded and have been further clarified in areas re-lating to equipment qualification tests. For example, existing Office of Inspection and Enforcement inspection procedures provide for (1) reviews of selected vendor supplied documents which identify the environmental qualification testing performed for safety-related electrical equipment, and (2) inspections to assess the adequacy of the licensee's or applicant's cuality assurance programs which are designed to assure that safety-related electrical equipment installed at the facility have been qualified in accordance with the vendor's testing programs.

Additionally, increased emphasis has been placed on the ins 7ec-tor's involvement in the pra-operational testing phase of facility systems.

The inspectors witness an increased number of testing activities and perform a more thorough review of s4fety-related test results.

4.0 CURRENT ACTIVITIES The nuclear industry and the NRC have various procrams currently in progress related to the environmental qualification of safety-related electrical equipment.

Activities of each of these procrans are discussed in the following sections.

2190.334 ec wIc

A 21 4.1 Systematic Evaluation Program, (NRC/0NRR)

The Division of Operating Reactors, ONRR, presently has undervaY Phase II of the Systematic Evaluation Program (SEP).

This Pro-gram consists of the systematic review of eleven older nuclear power facilities (plants licensed for operation before 1969 and those which require a review for conversion of a Provisional Operating License to a Full-Tem Operating License) to determine and document the degree to which they meet Turrent licensing requirements for new plants. Phase II of the SEP was approved by the Commission on November 9,1977.

The results of this systematic evaluation, which is scheduled for completion within the next *hree years, will be considered by the Commission in deciding whether the program should be extended to include other operating facilities.

One of the specific review topics included in this program is titled " Environmental Qualification of Safety-Related Ecuip-ment," (SEP Topic List No. III-12).

The objective of the SEP review of this topic is to evaluate the degree to which the mechanical and Class IE electrical equipment of safety-related systems have been qualified for the most severe environment (i..e., temperature, pressure, humidity, steam, chemistry, and radiation) of design basis accidents. As such, the SEP will be directed toward the determination of existing safety margins and the evaluation of the adequacy of such safety margins to deterTnine if any backfittino or facility upgradi>g is necessary.

!; D ( C n

2190 335

A-22 4.2 NRR Category A Technical Activities As part of NRR's Technical Activities Program, which was devel-oped to provide a basic framework of policy, organizational structure, priority, and procedures for the effective.nanagement of the major technical activities within NRR, those generic technical activities judged by the staff to warrant priority attention in terms of resources to attain early resolution were designated as Category A Priority Activities. Of those so designated, two are directly related to the environmental qualification of safety-related mechanical and electrical equipment.

A brief description of the applicable portions of each of these generic technical activities is provided below.

4.2.1 Cateoory A Technical Activity No. A-21, " Main Steam Line Break Inside Containment" One of the subtasks of this technical activity is to perform an evaluation of the procedures and content of analyses performed to establish environmental enve-lope requirements for main steam line break accidents inside containment. These environmental envelope requirements will then be utilized to assess the adeauacy of the environmental qualification of safety-rela 1.4 2190 336 i

A-23 equipment inside containment.

In addition, criteria for the methodology of environmental simulation will be evaluated to determine if the important environmental parameters have been properly simulated during *,esting.

The target date for completion of this task is December 1978.

4.2.2 Cateoory A Technical Activity No., A-24, " Qualification of Class IE Safety Related Equicment" The objective of this technical activity is to perform a generic review and evaluation of methods developed by industry to qualify safety-related equipment to the requirements established in IEEE Standard 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations." Certain concepts and methods proposed by industry in addressing equipment qualification, such as testing margins, aging effects on materials and equipment, and adequacy of testing simulators (which sim-ulate che worst case environment for the testing of equipment) have yet to be reviewed and accepted by the staff.

In order to expedite the review of the adeouacy of the proposed qualification methods, a generic review 2190 337

,4

A-24 of the qualification methodology and associated acceptance criteria used by the major NSSS vendors and Balance-of-Plant equipment suppliers will be conducted.

The Target Date for completion of this task is early 1979.

4.3 Confirmatory Research Procrams (NRC/RES)

The current NRC Qualification Testing Evaluation Program is directed towards providing a confirmatory assessment of current environmental qualification testing procedures for LOCA conditions and includes the following specific program elements:

1.

Assessment to determine if sequential (as opposed to simultaneous) environmental qualification testing is conservative, i.e., an investigation of synergistic effects.

2.

Confirmation that accelerated aging methodolgy can be utilized for qualification testing of safety-related equipment.

3.

02finition of the nuclear radiation source based on the Regulatory Guide 1.89 accident assumptions and evaluation of the adequacy of radiation simulators.

4.3.1 Syneraistic Tests The tests were to confirm that the sequential test sequence recommended in IEEE Std. 323-1974 conservatively simulates g{[

[iG hee combined radiation and steam environment to which safety-related equipment wnuld be exposed in the unlikely event of a 2190 338

A-25 LOCA. A research program to investigate potential synergistic effects was initiated at Sandia Laboratories in FY 1975.

Preliminary evaluation of the Sandia tests which have been completed to date does not indicate a significant functional synergism for electrical cables; however, with respect to connectors, it was not possible to determine whether synergism exists because of the failures that occurred.

4.3.2 Aging Effects Considerations of aging in enviromental qualification test programs are important because of the potential to create a weakened condition in a safety-related component through some aging mechanism that may not be detected through routine periodic testing. A research program to develop a methodology that can be utilized for simulation of the natural acing process of safety-related materials on an accelerated basis was initiated in FY 1976 and is continuing.

2190 339 un 0?!S

A-26 4.3.3 Source Term Equivalences One of the exceptions taken to IEEE Std 323-1974 by the NRC is the required radiation environment to be utilized for enviromental qualification testing. LOCA radiation releases to be used by an applicant in establishing the nuclear radiation environment for type testing are defined in Regulatory Guide 1.89.

An assessment of the adequacy of currently used radiation simulators to duplicate the accident radiation environment requires additional experimental evaluation. A research program to assist in this evaluation was initiated in FY 1976 and is continuing. Progress to date has consisted of analysis to determine the post-LOCA time relationship of dose, dose rate, energy spectra and particle type.

These data show that current industry practice with regard to radiation simulation testing may be significantly different in terms of dose rate, spectrum and particle type than that described in Regulatory Guide 1.89.

The ongoing work in this area is aimed at determining the importance of these differences f'

in terms of damage to safety related equipment.

2190 340

A-27 4.4 Standards Development Programs IEEE Standards related to the environmental qualification of the following specific safety-related electrical equipment are currently under development:

(a) Fire stops (b) Fire breaks (c) Storage batteries (d)

Switchgear (e) Circuit breakers

( f) Battery chargers (g) Transformers (h) Motor control centers A general Standard addressing the environmental cualification of safety-related mechanical and electrical eouipment is being developed jointly by IEEE and ASME.

It is anticipated that NRC Regulatory Guides endorsing the above-mentioned Standards will be developed through FY 1979.

NRC Regulatory Guide 1.40 is currently being revised to reflect the updated requirements of IEEE Standard 334-1974 NRC Regulatory Guides 1.89 and 1.131 are being revised sub-sequent to their issuance for public comment.

2190 341

A-28

5.0 REFERENCES

1.

" Guide to the Organization and Contents of Safety Analysis Reports," U.S. Atomic Energy Commission, June 30, 1976.

2.

" Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Regul. tory Guide 1.70, Revision 1, U.S. Atomic Energy Commission, October 1972; Revision 2, NUREG-75/094, U.S. Nuclear Regulatory Commis-sion, September,1975.

3.

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-75-087, U.S.

Nuclear Regulatory Commission, September,1975.

4.

" Reactor Safety Stucy:

An Assessment of Accident Risks In U.S. Commerical huclear Power Plants," WASH-la00 (NUREG-75/014), U.S. Nuclear Regulatory Commission, October 1975.

2 MO 342 i s.

note I $

l_) \\ 1 J

TABLE 1 373 v

rs-Time Phasin 9 of the Development of IEEE Standards Addressing the cc-Environmental Qualification of

-13 Safety-Related Electrical Equipment for Nuclear Power Generating Stations Vt 1968 1969 1970 1971 1972 1973 1974 1975 1976 1977 1978 I

I I

i i

i i

i i

l i

IEEE Std.

IEEE Std. 279-1968 279-1971 IEEE Std.

IEEE Std. 323-1971 323-1974 IEEE Std.

IEEE Std.

IEEE Std. 317-1971 317-1972 317-1976

[3 IEEE Std.

IEEE Std. 334-1971 334-1974 o

IEEE Std. 382-1972 IEEE Std.

ps) 381-1977 hff IEEE Std. 383-1974 u

A LN

A-30 TABLE 2 Facility Groupings by the Evolutionary Stage of NRC Licensing Requirements For Environmental Qualification of Safety-Related Electrical Equipment A.

Pre-1967 Construction Permit Operating License Facility Name Issued Issued Dresden Unit No.1 5/56 9/59 Yankee-Rowe 11/57 7/60 Humboldt Bay Unit No. 3 11/60 8/62 Big Rock Point 5/60 8/62 Indian Point Unit No. 1 5/56 3/62 San Onofre Unit No.1 3/64 3/67 Connecticut Yankee 5/64 6/67 (Haddam Neck)

Lacrosse 3/63 7/67 B.

1967-1974 Oyster Creek 12/64 4/69 Nine Mile Point Unit No. 1 4/65 8/69 Dresden Unit No. 2 1/66 12/69 Ginna 4/66 9/69 Miilstone Unit No. 1 5/66 10/70 Dresden Unit No. 3 10/66 1/71 3 :

IndianLPoint Unit No. 2 10/66 10/71 Quad Cities Units Nos.1 & 2 2/67 10/71/3/72 Palidades 3/67 3/71 Robinson Unit No. 2 4/67 7/70 2190 344

A-31 Construction Permit Operating License Facility Name Issued Issued Turkey Point Units Nos. 3 & 4 4/67 7/72/4/73 Browns Ferry Units Nos.1 & 2 5/67 6/63/6/74 Monticello 6/67 9/70 Point Beach Unit No. 1 7/67 10/70 Oconee Units 1, 2 & 3 11/67 2/73/10/73/7/74 Vermont Yankee 12/67 3/72 Peach Bottom Units 2 & 3 1/68 8/73/7/74 Diablo Canyon Unit 1 a/68 Three Mile Island Unit 1 5/68 4/74 Cooper 6/68 1/74 Ft. Calhoun 6/68 5/73 Prairie Island Units 1 & 2 5/68 8/73/10/74 Surry Units 1 & 2 6/68 5/72/1/73 Point Beach Unit No. 2 7/68 11/71 Browns Ferry Unit No. 3 7/68 7/76 Kewaunee 8/68 12/72 Pilgrim Unit No. 1 8/68 6/72 Ft. St. Vrain 9/68 12/73 Crystal River Unit No. 3 9/68 12/76 Salem Unit Nos. 1 & 2 9/68 8/76 Rancho Seco 10/68 8/74 Maine Yankee 10/68 9/72 Arkansas 1 12/68 5/74 Zion Units Nos. 1 & 2 12/68 4/73/11/73 2190 345

A-32 Construction Permit Operating License Facility Name Issued Issued D. C. Cook Units 1 & 2 3/69 10/74 Calvert Cliffs Units 1 & 2 7/69 7/74/8/76 Indian Point Unit 3 8/69 12/75 Hatch Unit 1 9/69 8/74 Three Mile Island 2 11/69 Brunswick Units 1 & 2 2/70 9/76/12/74 Fitzpatrick 3/70 10/74 Sequoyah 1 & 2 5/70 Duane Arnold 6/70 2/74 Beaver Valley Unit 1 6/70 1/76 Diablo Canyon 2 12/70 St. Lucie Unit 1 7/70 3/76 Millstone Un t 2 12/70 8/75 Trojan 2/71 11/75 North Arna 1 & 2 2/71 Davis-Besse 3/71 4/77 Farley 1 & 2 8/72 6/77 Fermi 2 9/72 Zimmer 1 10/77 Arkansas 2 12/72 2190 346 Midland 1 & 2 12/72 Hatch 2 12/72 Watts Bar 1 & 2 1/73 2 h t.

Ok

A-33 Construction Permit Operating License Facility Name Issued Issued McGuire 1 & 2 3/73 Washington Nuclear 2 3/73 Summer 1 3/73 Shoreham 4/73 Forked River 7/73 LaSalle 1 & 2 7/73 San Onofre 2 & 3 10/73 Susquehanna 1 & 2 11/73 Sailly 1 5/74 Beaver Valley 2 5/74 Limerick 1 & 2 6/74 iline Mile Point 2 6/74 Vogtle 1 & 2 6/74 C.

Post-July 197a North Anna 3 & 4 7/74 Millstone 3 8/74 Grand Gulf 1 & 2 9/74 Hope Creek 1,& 2 g

(

11/74 o.-

o Waterford 3 11/74 Comancne Peak i'& 2 12/74 Surry 3 & 4 12/74 Bellefonte 1 & 2 12/74 2190 347 Catawba 1 & 2 8/75 I

A-34 Construction Pemit Operating License Facility Name Issued Issued South Texas 1 & 2 8/75 Washingtin Nuclear 1 12/75 Byron 1 & 2 12/75 Braidwood 1 & 2 12/75 Clinton 1 & 2 2/76 Seabrook 1 & 2 2/76 Callaway 1 & 2

'/76 Palo Verde 1, 2 & 3 5/76 Hartsville 1, 2, 3, & 4 5/77 Perry 1 & 2 5/77 Wnif Creek 1 5/77 River Bend 1 & 2 3/77 St. Lucie 2 5/77 Sterling 1 9/77 And all other CP acolications currently under staff review 2190 a48

(

s s +

TABLE 3 Time Phasing of the Developnent of AEC/NRC Requirements and/or Guidance for the Environmental qualification of Safety-Related Electrical Equipment 1965 1966 1967 1968 1969 1970 1971 1972 1973 1974 1975 1976 1917 Draf t Genieral Revised General General Design Criteria Design Criteria

~

De. sign Published for Published for Criteria Coninent 11/65 Consnent 7/67 Incorporated (GDC-16)

(GDC-4,23)

~'

into Regula-tions, Appen-dix A to 10 CFR Part 50 5/71 (GDC-4,23)

Revision to 10 CFR 50.55a Y

10 CFR Part 50

" Codes and Star-

!N (650.55a)," Codes dards," incorporated and Standards,"

into f.egulations 7/71 Published for (Endorsed IEEE coninent 11/69 Std. 279)

AEC R.G. 1.89 (Endorsed IEEE N

Std. 323-1974) 11/74

~D CD AEC Published a AEC R.G. 1.70, Rev. 1 R.G. 1.70 NRC R.G.

Guide for Prepara-(Standard format and Rev. 2 1.131

(*

tion of Safety Analysis Content of Safety 9/75 (endorsing Reports 6/66 Analysis Repe-ts)

IEEE Std.

so 10/72 NRC Stan-383-1974) dard Review Published Plan 9/75 for comment 8/77 1965-1966 1967 1968 1969 1970 1971 1972 1973 1974 1975 1976 1977

___t AEC I.G.2 cs Published

,c 10/71 to AEC R.G.

Revision 1 to 1.63 R.G. 1.63 (endorsing (endorsing IEEE Std. 317-IEEE Std. 317-1972) 10/73 1976) published for comment S/77 AEC R.G.

1.40 (endorsin9 IEEE Std.

?

334-1971)

E N

3/73 8

ye"a-(endorsing y

IEEE Std.

w 382-1972) o 1/74

Appendix B PROCEDURE FOR EVALUATION OF ENVIRONMENTAL OUALIFICATION OF SAFETY-RELATED ELECTRICAL EQUIPMENT AND COMPONENTS The staff has determined that it is appropriate to complete the review of this subject as the first topic of the Systematic Evaluation Program. The licensees of the eleven SEP facilities will be required to evaluate the environmental qualification of all electrical equipment they deem necessary to mitigate the consequences of Design Basis Events.

It is expectec' that within about 90 days the review effort will be sufficient to asses; any safety implications in sufficient detail to decide whether or not additional review of facilities other than those included in the SEP is required. The results of this staff review will be used to make a determination of whether it is necessary to expand this effort from the eleven SEP facilities to other operating nuclear power plants. The staff's bases for this approach are set forth in its report, titled,

" Staff Report on Environmental Qualification of Safety-Related Electrical Equipment" dated December 15, 1977.

The evaluation of these eleven older operating reactor facilities will be conducted in accordance with the following procedures:

1.

Objective of Evaluation To confirm that electrical equipment necessary to mitigate the consequences;of DBE have been demonstrated, by test, analysis, or NEL Vsia operating experience, to have the capability to perfom its design 2190 351 B-1

B-2 safety function under the environmental conditions of Design Basis Events.

To determine actions that need to be taken te qualify appropriate equipment in accordance with current requirements.

2.

Information To Be Provided By Licensees Licensees will be requested to provide the following information:

A.

Identificatior. of safety-related systems and associated electrical equipment located both inside and outside con-tainment which are required to perform a W ety function under the environmental conditions resulting from each DBE.

Briefly describe the safety function provided by each item of equipment identified.

Describe the location of the equipment.

Identify any non-safety system, equipment or components, which, if subjected to the environmental conditions associated with a DBE, could affect the safety function of any safety-related system.

Identify non-safety grade systems which could perform the function of safety systems by ameliorating the consequences of a DBE and specify electrical components required to assure function of such non-safety grade systems.

2190 352

,I

B-3 B.

Definition of the limiting service environmental conditions for operation of the equipment and components identified above.

Tho environmental parameters to be included are pressure, temperature, radiation, submergence, steam, humidity, chemicals, vibration or any combination of the above (seismic conditions are not included in this evaluation but will be considered elsewhere in the SEP). These environmental parameters should be presented as a function of time and the DBE producing the conditions should be identified.

The time period during which each item of equipment would be required to operate in a DBE environment should also be identified.

C.

Determination of the current status of environmental qualifica-tion for safety-related electrical equipment and identification of the suoporting documentation. Any evidence of environmental qualification for any environmental condition should be considered and provided.

3.

Staff Review of Previously Documented Environmental Qualification The staff will re-examine any environmental qualifications previously accomplished on these facilities, such as the information submitted by licensees on modifications performed to demonstrate compliance with Appendix K of 10 CFR Part 50.

r.

2190 a53

B-4 4.

Staff Determination of Plant Environmental Conditions The staff will review and verify the environmental conditions provided by licensees for main steam line breaks inside con-tainment, for the limiting loss of coolant accident, and for other DBEs.

5.

Site Visit by Staff Review Team Early in the review process, staff members will visit the facility to discuss with the licensee the status of his response and to discuss alternatives considered by the licensee for satisfying the environmental qualification acceptance criteria.

6.

Identification of Significant Safety Problems Following the site visit, the staff will determine if there is inadequate evidence of environmental qualification for any safety-related electrical equipment which must function in a severe environment to mitigate the consequences of a DBE.

If such 1rade-quacies are found, appropriate action will be taken to assure no undue risk to the health a,nd safety of the public.

7.

Staff Evaluation The staff will evaluate the information submitted by the licensee in acccrdance with Item 2.

If necessary, another site visit may be made at this time to clarify or amplify the licensee's subnit.al.

3d U 9 i 'i.

2190.$54

B-5 In tFe event that documentation ' environmental testing does not exist, or is insufficient t' ossure environmental qualification, the following alternatives will be considered:

A.

Evidence of qualification of identical or similar equipment, either in another nuclear facility or by another industry.

B.

Importance of the safe'tY Ifunctio$ $ssociatec with any questionable equipment will be considered. Demonstration, of adequate facility response to all DBEs without credit for the function of an unqualified component may be justification for not requiring environmental qualification of a specific component.

C.

If important safety equipment is not environnentally qualified, consideration will be given to alternate ways of performing the safety function by using different systems, including the use of non-safety grade systems.

D.

Consideration will be given to possible means of protecting ccioponents from adverse environmental conditions, such as by enclosing it, coating it or providing other protective features.

E.

Other alternatives that may be proposed by the licensee will be considered.

8.

Staff Recort A staff report will document the evaluation of this safety topic for each facility.

The associated information will be provided 2190 355

B-6 in a format that is compatible with the SEP evaluation procedure, such that the evaluation results can easily be incorporated into the integrated SEP evaluation report for each facility.

2190 556 L

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I NUCLE AR HEGUL ATORY COM: ilSSION WASHING TON, D. C.

20' 55 POST AGE AND F E ES PAID

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PEN ALTY FOR PRIV ATE 15E, $ 300

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