ML13333A540

Requests That Util Perform Review of Environ Qualification of safety-related Electrical Equipment Against Encl Guidelines.Also Requests That Util Attend 800221 Meeting in Bethesda,Md Re Qualification Review Process

ML13333A540
Person / Time
Site:	San Onofre
Issue date:	02/15/1980
From:	Ziemann D Office of Nuclear Reactor Regulation
To:	Dietch R SOUTHERN CALIFORNIA EDISON CO.
References
TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8003200150
Download: ML13333A540 (40)

Text

1e TEGULATORY ROcET FILE CY Docket No.

50-206 Mr. R. Dietch Vice President Nuclear Engineering and Operations Southern Calfironia Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770

Dear Mr. Dietch:

RE: ELECTRICAL EQUIPMENT ENVIRONMENTAL QUALIFICATION SAN ONOFRE NUCLEAR GENERATION STATION, UNIT NO. 1 Reviews of the environmental qualification of safety-related electrical equipment have been initiated under Topic 111-12 of the SEP program.

The reviews are being conducted in accordance with Enclosure 1, "Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors." The documents considered in the reviews include licensee submittals on SEP Topic 111-12 and supporting material including equipment qualification records.

Preliminary results of our review indicate deficiencies affecting the scope of equipment addressed, the definition of its environment, and the completeness of the supporting documentation.

Accordingly, it is dce;ed necessary to increase the level of effort applied to thesc reviews by th-ie SEP licensees.

The SEP licensees are asked to perform the review of their safety related equipment against the Guidelines of Enclosure 1. This program radircction will accoleratc tho qualification revicw process by assuring that the site visit by the NRC staff, which heretofore has served largely to generate questions concerning deficiences in documentation, can instead concentrate on (1) review of discrepancies previously identified by the licensee and (2) auditing the basis for accepting the qualification of other equipment.

All SEP licensees are requested to attend a meeting at the NRC Headquarters in Bethesda, Maryland, at 10:00 a.m., on February 21, 1980 in Room P-118 of the Phillips Buliding. The purpose of this meeting is to ensure full OFFICE SURNAM E D A T E

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NRC FORM 318 (9-761 NRCM 0240

',-U.S. GOVERNMENT PRINTING OFFICE: 1979-289-369 90&

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-2 understanding of the qualification review process.

Necessary licensee submittals, site visits, and related schedules will be covered. In order that you may begin your activity on this accelerated effort, we are providing Enclosure 2, "Guidelines for Identification of That Safety Equipment of SEP Operating Reactors for Which Environmental Qualification is to be Addressed."

Sincerely, Dr818 Signed by homaa V. fambaeh Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors

Enclosures:

1. Guidelines for Evaluating

2. Guidelines for Identification DISTRIBUTION Docket (URC PDR Local PDR ORB-Reading NIRR REading DEisenhut RVollmer OELD OISE (3)

DLziemann PO"Connor HSmith HISIC TERA ACRS (16)

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LPO'Connor:c 1ziemann SURNAME DATE NUC FORK 318 (9-76) NRCM 0240 u..

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1979 75 789

Mr.

FEbruary 15, 1980 cc Charles R. Kocher, Assistant Director, Technical Assessment General Counsel Division Southern California Edison Company Office of Radiation Programs Post Office Box 800 (AW-459)

Rosemead, California 91770 U. S. Environmental Protection Agency David R. Pigott Crystal Mall #2 Samuel B. Casey Arlington, Virginia 20460 Chickering & Gregory Three Embarcadero Center U. S. Environmental Protection Twenty-Third Floor Agency San Francisco, California 94111 Region IX Office ATTN: EIS COORDINATOR Jack E. Thomas 215 Freemont Street Harry B. Stoehr San Francisco, California 94111 San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Resident Inspector c/o U. S. NRC P. 0. Box 3550 San Onofre, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego San Diego, California 92101 California Department of Health ATTN:

Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814

W ENCLOSURE 1 GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS 1.0 Introduction 2.0 Discussion 3.0 Identification of Class IE Equipment 4.0 Service Conditions 4.1 Service Conditions Inside Containment for a Loss of Coolant Accident (LOCA)

1. Temperature and Pressure Steam Conditions

2. Radiation

3. Submergence

4. Chemical Sprays 4.2 Service Conditions for a PWR Main Steam Line Break (MSLB)

Inside Containment

1. Temperature'and Pressure Steam Conditions

2. Radiation

3. Submergence

4. Chemical Sprays 4.3 Service Conditions Outside Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a High Energy Line Break (HELB) 4.3.2 Areas Where Fluids are Recirculated From Inside Containment to Accomplish Long-Term Emergency Core Cooling Following a LOCA

1. TemDerature, Pressure and Relative Humidity

2. Radiation

3. Submergence

4. Chemical Sprays

-2 4.3.3 Areas Normally Mai tained at Room Conditions 5.0 Qualification Methods 5.1 Selection of Qualification Method 5.2 Qualification by Type Testing

1. Simulated Service Conditions and Test Duration

2. Test Specimen

3. Test Sequence

4. Test Specimen Aging

5. Functional Testing and Failure Criteria

6. Installation Interfaces 5.3 Qualification by a Combination of Methods (Test, Evaluation, Analysis) 6.0 Margin 7.0 Aging 8.0 Documentation Appendix A - Typical Equipment/Functions Needed for Mitigation of a LOCA or MSLB Accident Appendix B - Guidelines for Evaluating Radiation Service Conditions Inside Containment for a LOCA and MSLB Accident Appendix C - Thermal and Radiation Aging Degradation of Selected Materials

GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS

1.0 INTRODUCTION

On February 8, 1979, the NRC Office of Inspection and Enforcement issued IE Bulletin 79-01, entitled, "Environmental Qualification of Class IE Equipment."

This bulletin requested that licensees for operating power reactors complete within 120 days their reviews of equipment qualification begun earlier in connection with IE Circular 78-08. The objective of IE Circular 78-08 was to initiate a review by the licensees to determine whether proper documentation existed to verify that all Class IE electrical equipment would function as required in the hostile environment which could result from design basis events.

The licensees' reviews are now essentially complete and the NRC staff has begun to evaluate the results.

This document sets forth guidelines for the NRC staff to use in its evaluations of the licensees' responses to IE Bulletin 79-01 and selected associated qualification documentation. The objective of the evaluations using these guidelines is to identify Class IE equipment whose documentation does not provide reasonable assurance of environ mental qualification. All such equipment identified will then be subjected to a plant application specific evaluation to determine whether it should be requalified or replaced with a component whose qualification has been adequately verified.

These cuidel ines are intended to be used by the NRC staff to evaluate the qualification methods used for existing equipment in a particular class of plants, i.e., currently operating reactors including SEP plants.

0 Equipmrnt in other classes of plants not yet licensed to operate, or replacement equipment for operating reactors, may be subject to different requirements such as those set forth in NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.

In addition to its reviews in connection with IE Bulletin 79-01 the staff is engaged in other generic-reviews that include aspects of the equipment qualification issue. TMI-2 lessons learned and the effects of failures of non-Class IE control and indication equipment are examples of these generic reviews. In some cases these guidelines may be applicable, however, this determination will be made as part of that related generic review.

2.0 DISCUSSION IEEE Std. 323-19741 is the current industry standard for environmental qualification of safety-related electrical equipment. This standard was first issued as a trail use standard, IEEE Std. 323-1971, in 1971 and later after substantial revision, the current version was issued in 1974. Both versions of the standard set forth generic requirements for equipment quali fication but the 1974 standard includes specific requirements for aging, margins, and maintaining documentation records that were not included in the 1971 trial use standard.

The intent of this document is not to provide guidelines for implementing either version of IEEE Std. 323 for operating reactors. In fact most of the operating reactors are not committed to comply with any particular industry standard for electrical equipment qualification. However, all of the operating reactors are required to comply with the General Design Criteria lIEEE Std. 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."

3 specified in Appendix A of 10 CFR 50. General Design Criterion 4 states in part that "structures, systems and components important to safet. shall be designed to accomodate the affects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents."

The intent of these guidelines is to provide a basis for judgements required to confirm that operating reactors are in compliance with General Design Criterion 4.

3.0 IDENTIFICATION OF CLASS IE EQUIPMENT Class IE equipment includes all electrical equipment needed to achieve emergency reactor shutdown, containment isolation, reactor core cooling, containment and reactor heat removal, and prevention of significant release of radioactive material to the environment. Typical systems included in pressurized and boiling water reactor designs to perform these functions for the most s.evere postulated loss of coolant accident (LOCA) and main steamline break accident (MSLB) are listed in Appendix A.

More detailed descriptions of the Class IE equipment installed at specific plants can be obtained from FSARs, Technical specifications, and emergency procedures. Although variation in nomenclature may exist at the various plants, environmental qualification of those systems which perform the functions identified in Appendix A should be evaluated against the appropriate service conditions (Section 4.0),

Tne go belines in this document are appiicabie to all components necessary for operation of the systems listed in Appendix A including but. not limited to valves, motors, cables, connectors, relays, switches, transmitters and valve position indicators.

0

-4 4.0 SERVICE CONDITIONS In order to determine the adequacy of the qualification of equipment it is necessary to specify the environment the equipment is exposed to during normal and accident conditions with a requirement to remain functional, These environments are referred to as the "service conditions."

The approved service conditions specified in the FSAR or other licensee submittals are acceptable, unless otherwise noted in the guidelines discussued below.

4.1 Service Conditions Inside Containment for a Loss of Coolant Accident (LOCA)

1. Temperature and Pressure Steam Conditions - In general, the containment temperature and pressure conditions as a function of time should be based on the analyses in the FSAR, In the specific case of pressure suppression type containments, the following minimum high tempeature

.conditions should be used; (1) BWR Drywells. 340 0F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and (2) PWR Ice Condenser Lower Compartments, 340 0F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

2. Radiation - When specifying radiation service conditions for equipment exposed to radiation during normal operating and accident conditions, the normal operating dose should be added to the dose received during the course of an accident. Guidelines for evaluating beta and gamma radiation service conditions for general areas inside containment are provided below. Radiation service condittons for equipment located directly above the containment sump, in the vicinity of filters, or submerged in contaminated liquids must be evaluated on a case by case basis.

Guidelines for these evaluations are not provided in this document.

5 Gamma Radiation Doses - A total gamma dose radiation servtce condition of 2 x 107 RADS is acceptable for Class IE equipm..it located in general areas inside containment for PWRs with dry type containments, Where a dose less than this value has been specified, an application specific evaluation must be performed to determine if the dose specified is acceptable. Procedures for evaluating radiation service conditions in such cases are provided in Appendix B, The procedures in Appendix B are based on the calculation for a typical PWR reported in Appendix D of NUREG,0588 Gamma dose radiation service conditions for BWRs and PWRs with ice condenser containments must be evaluated on a case by case basis, Since the procedures in Appendix B are based on a calculation for a typical PWR with a dry type contai'nment, they are not directly applicable to BWRs and other containment types, However, doses for these other plant configurations may be evaluated using similar procedures with conservative dose assumptions and adjustment factors developed on a case by case basts.

Beta Radiation Doses -

Beta radiation doses generally are less significant than gamma radiation doses for equipment qualification.. This is due to the low penetrating power of beta particles in comparison to gamma rays of equivalent energy, Of the general classes of electrical equipment in a plant (e~g,, cables, instrument transmi'tters, valve operators, containment penetrations), electrical cable is considered the most NUREG-0588, Interim Staff Position on Environmental Qualification of Safety.Related Electrical Equipment.

6 vulnerable to damage from beta radiation. Assuming a TID 14844 source term, the average maximum beta energy and isotopic abundance will vary as a function of time following an accident. If these parameters are considered in a detailed calculation, the conservative beta surface dose of 1.40 x x 108 RADS reported in Appendix D of NUREG 0588 would be reduced by approximately a factor of ten within 30 mils of the surface of electrical cable insulation of unit density. An additional 40 mils of insulation (total of 70 mils) results in another factor of 10 reduction in dose. Any structures or other equipment in the vicinity of the equipment of interest would act as shielding to further reduce beta doses. If it can be shown, by assuming a conserva tive unshielded surface beta dose of 2.0 x 108 RADS and considering the shielding factors discussed here, that the beta dose to radiation sensitive equipment internals would be less than or equal to 10% of the total gamma dose to which an item of equipment has been qualified, then that equipment may be considered qualified for the total radiation environment (gamma plus beta). If this criterion is not satisfied the radiation service condition should be determined by the sum of the gamma and beta doses.

3. Submergence - The preferred method of protection against the effects of submergency is to locate equipment above the water flooding level.

Specifying saturated steam as a service condition during type testing of equipment that will become flooded in service is not an acceptable alternative for actually flooding the equipment during the test.

7

4.

Containment Sprays - Equipment exposed to chemical sprays should be qualified for the most severe chemical environment (actdic or basic) which could exist. Demineralized water sprays should not be exempt from consideration as a potentially adverse service condition..

4.2 Service Conditions for a PWR Main Steam Line Break (MSLB) Inside Containment Equipment required to function in a steam line break environment must be qualified for the high temperature and pressure that could result.

In some cases the environmental stress on exposed equipment may be higher than that resulting from a LOCA, in others it may be no more severe than for a LOCA due to the automatic operation of a containment spray system.

1. Temperature and Pressure Steam Conditions - Equipment qualified for a LOCA environment is considered qualified for a MSLB accident environ ment in plants with automatic spray systems not subject to disabling single component failures. This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a thermal analysis of typical components inside containment./

The final acceptability of this approach, i.e., use of the "Best Estimate",

is pending the completion of Task Action Plan A-21, Main Steamline Break Inside Containment.

Class IE equipment installed in plants without automatic spray systems or plants with 4pray systems subject to disabling single failures or delayed initiation should be qualified for a MSLB accident environment determined by a plant specific analysis. Acceptable methods ISee NUREG 0458, Short Term Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reac:ors, for a more detailed discussion of the best estimate calculation.

for performing such an analysis for operating reac'-ors are provided in Section 1.2 for Category II plants in NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Elitctrical Equipment.

2. Radiation - Same as Section 4.1 above except that a conservative gamma dose of 2 x 106 RADS is acceptable.

3. Submergence - Same as Section 4.1 above,

4. Chemical Sprays - Same as Section 4.1 above.

4.3 Service Conditions Outside of Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a High Energy Line Break (HELB)

Service conditions for areas outside containment exposed to a HEL8 were evaluated on a plant by plant basis as part of a program initiated by the staff in December, 1972 to evaluate the effects of a HELB. The eqjipment required to mitigate the evert vas als; dertlf-ed.

This equipment should be qualified for the service conditions reviewed and approved in the HtLB SaFety Evaluation Report for each specific plant.

4.3.2 Areas Where Fluids are Recirculated from Inside Containment to Accomplish Lono-Term Core Coolina Followina a LOCA

1. Temperature and Relative Humidity - One hundred percent relative humidity should be established as a service condition in confined spaces.

The temperature and pressure as a function of time should be based on the uinuje analysis reported in the FSAR.

2. Radiation - Due to differences in equipment arrangement within these areas and the significant effect of this factor on doses, radiation service conditions must be evaluated on a case by case basis.

In general, a dose of at least 4 x 106 RADS would be expected.

3. Submergence - Not applicable.

4. Chemical Sprays -

Not applicable.

4.3.3 Areas Normally Maintained at Room Conditions Class IE equipment located in these areas does no experience significant stress due to a change in service conditions during a design basis event.

This equipment was designed and installed using standard engineering practices and industry codes and standards (e.g.,

ANSI, NEMA, National Electric Code). Based on these factors, failures of equipment in these a-eas during a design basis event are expected to be random except to t

-e extent that the may be due t-agi,: or failures of ai-cOr ti-ii7 or ventilation systems. Therefore, no special consideration need be given to tne environmental qualification of Class IE equipment in these areas provided tne aging requirements discussed in Section 7.0 below are satisfied and the areas are maintained at room conditions by redundant air conditioning or ventilation systems served by the onsite emergency electrical power system.

Equipment located in areas not served by redundant systems powered from onsite emergency sources should be qualified for the environmental extremes which could result from a failure of the systems as determined from a plant s:e:ific analysis.

5.0 QLIFICATION METHODS

-10 5.1 Selection of Qualification Method The choice of qualification method employed for a particular application of equipment is largely a matter of technical judgement based on such factors as: (1) the severity of the service conditions; (2) the structural and material complexity of the equipment; and (3) the degree of certainty required in the qualification procedure (i.e., the safety importance of the equipment function). Based on these considerations, type testing is the preferred method of qualification for electrical equipment located inside containment required to mitigate the consequences of design basis events, i.e., Class IE equipment (see Section 3.0 above). As a minimum, the qualification for severe temperature, pressure, and steam service conditions for Class IE equipment should be based on type testing.

Qualification for other service conditions such as radiation and chemical sprays may be by analysis (evaluation) supported by test data (see Section 5.3 below).

Exceptions to these general guidelines must be justified on a case by case basis.

5.2 Oualification by Type Testing The evaluation of test plans and results should include consideration of the following factors:

1. Simulated Service Conditions and Test Duration - The environment in the test chamber should be established and maintained so that it envelopes the service conditions defined in accordance with Section 4.0 above.

The time duration of the test should be at least as 'I-g as :.? period from the initiation of the accident until the temperature and pressure service conditions return to essentially the same levels that existed before the postulated accident. A shorter test duration may be acceptable

if specific analyses are provided to demonstrate that the materials involved v 11 not experience significant accelerated thermal aging during the period not tested.

2. Test Specimen - The test specimen should be the same model as the equipment being qualified. The type test should only be considered valid for equipment identical in design and material construction to the test specimen.. Any deviations should be evaluated as'part of the qualifica tion documentation (see also Section 8.0 below).

3. Test Sequence - The component being tested should be exposed to a steam/air environment at elevated temperature, and pressure in the sequence defined for its service conditions. Where radiation is a service condition which is to be considered as part of a type test, it may be applied at any time during the test sequence provided the component does not contain any materials which are known to be susceptible to significant radiation damage at the service condition levels or materials whose susceptibility to radiation damage is not known (see Appendix C).

If the component contains any such materials, the radiation dose should be applied prior to or concurrent with exposure to the elevated temperature and pressure steam/air environment. The same test specimen should be used throuchout the test sequence for all service conditions the equipment is to be qualified for by type testing. The type test should only be considered valid for the service conditions applied to the same test specimen in the appropriate sequence.

4. Test Specirnen Acinc - Tests which were successful using test specimens which had not been oreaged may be considered acceptable provided the component does not contain materials which are known to be susceptible

-12 to significant degradation due to thermal and radiation agir (see Section 7.0).

If the component contains such materials a qualified life for the component must be established on a case by case basis. Arrhenius techniques are generally considered acceptable for thermal aging.

5. Functional Testing and Failure Criteria - Operational modes tested should be representative of the actual application requirements (e.g., components which operate normally energized in the plant should be normally energized during the tests, motor and electrical cable loading during the test should be representative of actual operating conditions).

Failure criteria should include instrument accuracy requirements based on the maximum error assumed in the plant safety analyses. If a component fails at any time during the test, even in a so called "fail safe" mode, the test should be considered inconclusive with regard to demonstrating the ability of the component to function for the entire period prior to the failure.

6. Installation Interfaces - The eQuipment mounting and electrical or mechanical seals used during the type test should be representative

• of the actual installation for the test to be considered conclusive.

The equipment qualification program should include an as-built inspection in the field to verify that equipment was installed as it was tested. Particular emphasis should be placed on common problems such as protective enclosures installed upside down with drain holes at the top and penetrations in equipment housings for electrical connections being left unsealed or susceptible to moisture incursion through stranded conductors.

-13 5.3 Qualification by a Combination of Methods (Test, Evaluation, Analysis As discussed in Section 5.1 above, an item of Class IE equipment may be shown to be qualified for a complete spectrum of service conditions even though it was only type tested for high temperature, pressure and steam. The qualification for service conditions such as radiation and chemical sprays may be demonstrated by analysis (evaluation).

In such cases the overall qualification is said to be by a combination of methods.

Following are two specific examples of procedures that are considered acceptable. Other similar procedures may also be reviewed and found acceptable on a case by case basis.

1. Radiation Oualification - Some of the earlier tvDe tests performed for operating reactors did not include radiation as a service condition. In these cases the equipment may be shown to be radiation qualified by perfoinr.inc a calcjlation of thi d:se expected, taking into account the time the equipment is required to remain functional and its location using the methods described in Appendix B, and analyzing the effect of the calculated dose on the materials used in the equipment (see Appendix C).

As a general rule, the time required to remain functional assumed for dose calculations should be at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2. Chemical Spray Qualification - Components enclosed entirely in corrosion resistant cases (e.g.,

stainless steel) nay be shown to be qualified for a chem'Icfl environment by an aralysis of the effects of the particular chemicals on the pari:Jar enclo sure materials..

The effects of che-ical sprays or the pressure integrity of any gaskets or seals present should b4: considered in the analysis.

I..

-14 6.'0 Margin IEEE Std. 323-1974 d ines margin as the difference between the most severe specified service conditions of the plant and the conditions used in type testing to account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance.

Section 6.3.1.5 of the standard provides suggested factors to be applied to the service conditiohs to assure adequate margins. The factor applied to the time equipment is required to remain functional is the most significant in terms of the additional confidence in qualification that is achieved by adding margins to service conditions when establishing test environments.

For this reason, special consideration was given to the time required to remain functional when the guidelines for Functional Testing and Failure Criteria in Section 5.2 above were established. In addition, all of the guidelines in Section 4.0 for establishing service conditions include conservatisms which assure margins between the service conditions specified and the actual conditions which could realistically be expected in a design basis event. Therefore, if the guidelines in Section 4.0 and 5.2 are satisfiedzno separate margin factors are required to be added to the service conditions when specifying test conditions.

7.0 Aging Implicit in the staff position in Regulatory Guide 1.89 with regard to backfitting IEEE Std. 323-1974 is the staff's conclusion that the incremental improvement in safety from arbitrarily requiring that a specific qualified life be demonstrated for all Class IE equipmeit is not sufficient to justify the exp nse for plants already constructed and operating. This position does not, however, exclude equipment

15 using materials that have been identified as being susceptible to significant degradation due to thermal and radiation aging. Component maintenance or replacement schedules should include considerations of the specific aging characteristics of the component materials. Ongoing programs should exist at the plant to review surveillance and maintenance records to assure that equipment which is exhibiting age related degrada tion will be identified and replaced as necessary. Appendix C contains a listing of materials which may be found in nuclear power plants along with an indication of the material susceptability to thermal and radiation aging.

8.0 Documentation Complete and auditable records must be available for qualification by any of the methods described in Section 5.0 above to be considered valid.

These records should describe the qualification method in sufficient detail to verify that all of the guidelines have been satisfied. A simple vendor certification of compliance with a design specification should not be considered adequate.

APPENDIX A TYPICAL EQUIPMENT/FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and -Isolation Emergency Power Emergency Core Cooling1 Containment Heat Removal Containment Fission Product Removal Containment Combustible Gas Control Auxiliary Feedwater Containment Ventilation Containment Radiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)

Ventilation for Areas Containing Safety Equipment Component Cooling Service Water Emergency Shutdown 2 Post Accident Sampling and Monitoring 3 Radiation Monitorinc3 Safety Related Display Instrumentation 3

-2 1Ths pThesesystems will differ for PWRs and BWRs, and for oldcr and newer plants. In each case the system features which allow fo'.

transfer to recirculation cooling mode and establishment of long term cooling with boron precipitation control are to be considered as part of the system to be evaluated.

2Emergency shutdown systems include those systems used to bring the plant to a cold shutdown condition following accidents which do not result in a breach of the reactor coolant pressure boundary together with a rapid depressurization of the reactor coolant system. Examples' of such systems and equipment are the RHR system, PORVs, RCIC, pressurizer sprays, chemical and volume control system, and steam dump systems.

3More specific identification of these types of equipment can be found in the plant emergency procedures.

APPENDIX B PROCEDURES FOR EVALUATING GAMMA RADIATION SERVICE CONDMTONS Introduction and Discussion The adequacy of gamma radiation service conditions specified for inside containment during a LOCA or MSLB accident can be verified by assuming a conservative dose at the containment centerline and adjusting the dose according the plant specific parameters. The purpose of this appendix is to identify those parameters whose effect on the total gamma dose is easy to quantify with a high degree of confidence and describe procedures which may be used to take these effects into consideration.

The bases for the procedures and restrictions for their use are as follows:

(1) A conservative dose at the containment centerline of 2 x 107 RADS for a LOCA and 2 x 106 RADS for a MSLB accident has been assumed.

This assumption and all the dose rates used in the procedure out lined below are based on the methods and sample calculation described in Appendix D of NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equip ment.,"

Therefore, all the limitations listed in Appendix D of NUREG-0588 apply to these procedures.

(2) The sample calculation in Appendix D of NUREG-0588 is for a 4,000 MWth pressurized water reactor housed in a 2.52 x 106 ft3 contain ment with an iodine scrubbing spray system. A similar calculation without iodine scrubbing sprays ul increase the dose to equipment approximately 155.

The conservative dose of 2 x 107 RJDS assumed

(21 in the procedure below includes sufficient conservatism to account for this factor. Therefore, the pro Adure is also applicable to plants without an iodine scrubbing spray system.

(3) Shielding calculations are based on an average gamma energy of 1 MEV derived from TID 14844.

(4) These procedures are not applicable to equipment located directly above the containment sump, submerged in contaminated liquids, or near filters. Doses specified for equipment located in these areas must be evaluated on a case by case basis.

(5) Since the dose adjustment factors used in these procedures are based on a calculation for a typical pressurized water reactor with a dry type containment, they are not directly applicable to boiling water reactors or other containment types. However, doses for these other plant configurations may be evaluated using similar procedures with conservative dose assumotions and adjustment factors developed on a case by case basis.

Procedure Figures 1 through 4 provide factors to be applied to the conservative dose to correct the dose for the following plant specific parameters:

(1) reactor power level; (2) containment volume; (3) shielding; (4) compartment volume; and (5) time equipment is required to remain functional.

-3 The procedure for using the figures is best illustrated by an example.

Consider the following case. The radiation service condition for a particular item of equipment has been specified as 2 x 106 RADS. The application specific parameters are:

Reactor power level -

3,000 MWth Containment volume -

2.5 x 106 ft3 Compartment Volume -

8,000 ft3 Thickness of compartment shield wall (concrete) -

24" Time equipment is required to remain functional - 1 hr.

The problem is to make a reasonable estimate of the dose that the equipment could be expected to receive in order to evaluate the adequacy of the radiation service condition specification.

Step 1 Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and 2.5 x 106 ft3 containment volume and read a 30-day integrated dose of 1.5 x 107 RADS.

Step 2 Enter Figure 2 at a dose of 1.5 x 107 RADS and 24" of concrete shielding for the compartment the equipment is located in and read 4.5 x 104 RADS.

This is the dose the equipment receives from sources outside the compart ment. To this must be added the dose from sources -inside the compartment (Step 3).

Step 3 Enter Figure 3 at 8,000 ft3 and read a correction factor of 0.13.

The dose due to sources inside the compartment would then be 0.13 (1.5 x 107)

= 1.95 x 106 RADS. The sums of the doses from steps 2 and 3 equals:

4.5 x 104 RADS + 0.13 (1.5 x 107) RADS = 2.0 x 106 RADS

()

0 Step 4 Enter Figure 4 at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and read a ;orrection factor of 0.15. Apply this factor to the sum of the doses determined from steps 2 and 3 to correct the 30 day total dose to the equipment inside the compartment to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

0.15 (2.0 x 106) = 3 x 105 RADS In this particular example the service condition of 2 x 106 RADS specified is conservative with respect to the estimated dose of 3 x 105 RADS calculated in steps 1 through 4 and is, therefore, acceptable.

FIGURE 1 NOMOGRAMR CONTAINMENT VOLUME AND REACTOR OWER LOCA DOSE CORRECTIONS*

ZONTAINMENT VOLUME (ft3) 3 x 106 2 x 106 30 DAY MWTH INTEGRATED 1 x 10 -

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APPENDIX C THERMAL AND RADIATION AGING DEGRADATION OF SELECTED MATERIALS Table C-1 is a partial list of materials which may be found in a nuclear power plant along with an indication of the material susceptibility to radiation and thermal aging.

Susceptibility to significant thermal aging in a 450C environment and normal atmosphere for 10 or 40 years is indicated by an (*) in the appro priate column. Significant aging degradation is defined as that amount of degradation that would place in substantial doubt the ability of typical equipment using these materials to function in a hostile environment.

.Susceptibility to radiation damage is indicated by the dose level and the observed effect identified in the column headed BASIS. The meaning of the terms used to characterize the dose effect is as follows:

a Threshold - Refers to damage threshold, which is the radiation exposure required to change at least one physical property of the material.

@ Percent Change of Property - Refers to the radiation exposure required to change the physical property noted by the percent.

o Allowable - Refers to the radiation which can be absorbed before serious degradation occurs.

The information in this appendix is based on a literature search of sources including the National Technical Information Service (NTIS), the National Aeronautics and Space Administration's Scientific and Technical Aerospace Report (STAR), NTIS Government Report Announcements and Index (GRA), and

-2 various manufacturers data reports. The materials list is not to be considered all inciusive neither is it to be used as a basis for specifying materials to be used for specific applications within a nuclear plant.

The list is solely intended for use by the NRC staff in making judgements as to the possibility of a particular material in a particular application being susceptible to significant degradation due to radiation or thermal aging.

The data base for thermal and radiation aging in engineering materials is rapidly expanding at this time. As additional information becomes available Table C-1 will be updated accordingly.

TABLE C-1 THERMAL AND) RADIATION AGING DEGRADATION OF SELECTED MATERIALS

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ENCLOSURE I*

GUIDELINES FOR IDENTIFICATION OF THAT SAFETY EQUIPMENT OF SEP OPERATING REACTORS FOR WHICH ENVIRONMENTAL QUALIFICATION IS TO BE ADDRESSED For operating reactors, all electrical equipment needed to mitigate high energy line breaks (LOCA, MSLB, FWLB) inside or outside containment should be qualified to perform the required safety function in the accident environment.

To this end, it is necessary that for each operating reactor the electrical equipment needed to achieve emergency reactor shutdown, containment isolation, reactor core cooling, containment and reactor heat removal, and prevention of significant release of radioactive material to the environment be identified and evaluated against relevant qualification criteria.

The following process should be used by licensees to establish SEP plant specific lists of the electrical equipment for which environmental qualification must be addressed.

(1) Safety functions typically performed by plant safety systems are listed in Appendix A. For each safety function identified in Appendix A, list the systems, sub-systems, or components assumed available in the plant FSAR or emergency procedures to perform that function during a LOCA, MSLB inside or outside containment, or FWLB inside or outside containment. If a plant specific safety function not listed in Appendix A is identified, that function and the corresponding systems or equipment to perform the function should be added to the licensee's list.

(2) All the systems or equipment implied in Item (1) should be identified regardless of the original classification of the equipment when the plant received its operating license; i.e., some control grade equipment will probably be named in emergency procedures.

However, if plant emergency procedures specify a preferred mode of accident mitigation involving equipment recognized by the licensee as unlikely to meet environmental qualification criteria, an alternate mode of performing the safety function with potentially qualifiable equipment may be identified. In such cases, the emergency procedures must clearly indicate how the operator is to use environmentally qualified safety related display instrumentation to diagnose failure to perform safety functions and how the alternate equipment can be used to perform such safety functions.

-2 (3) Plant emergency procedures typically include provisions for the operator to sample or monitor radioactivity levels or combustible gas levels, to confirm that valves are in the correct position, to monitor flow or temperature, etc. Some of these functions are essential for correct operator action, to mitigate accidents, and prevent radioactive releases. When this is the case, the radiation sensors, valve position indicators, pressure transmitters, thermo couples, etc. should be qualified to function in the relevant accident environment.

Licensees should, therefore, review their emergency procedures to determine the electrical components needed to perform the functions of Safety Related Display Information, Post Accident Sampling and Monitoring, and Radiation Monitoring. When equipment implied by the emergency procedures is not listed, justification must be provided that failure of such equipment would not prevent accident mitigation or radioactivity release.

Equipment now indicated in emergency procedures in response to TMI-2 lessons learned should be listed.

.Environmental Qualification of equipment to be installed for

.implementation of Regulatory Guide 1.97, " Instrumentation for Post Accident Monitoring,"jwill be addressed at the time of implementation.

(4) To support the licensee's conclusions regarding the identification of equipment for which environmental qualification should be addressed, the plant emergency procedures for mitigation of LOCAs, MSLBs, or FWLBs should be submitted. The submittals should include all secondary procedures referenced in.the emergency procedures.

For example, procedures for attaining cold shutdown, radiation monitoring, sampling, etc. should be included.

(5) The licensee should document anticipated service conditions in every portion of the plant where the environment could be influenced by the accident or its consequences, or where redundant air conditioning or ventilation is not provided.

These service conditions should also be correlated with the safety-related systems and sub-systems identified in steps (1) through (4).

Whenever an item of safety-related equip ment may be located in an environment outside the range of normal conditions, place it on the list of equipment in a potentially hostile environment. Equipment subject to service conditions that are at no time more severe than the normal environment for which the equipment has been designed and under which its proper operation has been verified, can be regarded as qualified by experience and need not be considered further except with respect to the aging requirements of the qualification guidelines; however, the identi fication of this equipment and the justification for excluding it from further review shall be documented.

APPENDIX A TYPICAL EQUIPMENT/FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power Emergency Core Cooling1 Containment Heat Removal Containment Fission Product Removal Containment Combustible Gas Control Auxiliary Feedwater Containment Ventilation Containment Radiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)

Ventilation for Areas Containing Safety Equipment Corponent Cooling Service Water Emergency Shutdown2 Post Accident Sampling and Monitoring 3 Radiation Monitoring3 Safety Related Display Instrumentation 3

-2 These systems will differ for PWRs and EWRs, and for old'r and newer plants. In each case the system features which allow f*z transfer to recirculation cooling mode and establishment of long term cooling with boron precipitation control are to be considered as part of the system to be evaluated.

2 Emergency shutdown systems include those systems used to bring the plant to a cold shutdown condition following accidents which do not result in a breach of the reactor coolant pressure boundary together with a rapid depressurization of the reactor coolant system. Examples of such systems and equipment are the RHR system, PORVs, RCIC, pressurizer sprays, chemical and volume control system, and steam dump systems.

3More specific identification of these types of equipment can be found in the plant emergency procedures.