ML19318C533

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Advises That IE Bulletin 79-01B, Environ Qualification of Class IE Equipment, Was Sent to Listed Utils on 800116
ML19318C533
Person / Time
Site: Dresden 
Issue date: 01/30/1980
From: Carroll D
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 8007010606
Download: ML19318C533 (3)
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4 UNITED STATES

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NUCLEAR REGULATORY COMMISSION g

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799 ROOSEVELT ROAD o

GLEN ELLYN, ILLINOIS 60137 N 3 Qgg MEMORANDUM FOR: Those on Attached List FROM:

Dorothy E. Carroll, Acting Chief, Administrative Branch

SUBJECT:

IE BULLETIN NOTICE NO.79-01B The attached IE Bulletin No.79-01B titled " Environmental Qualification of Class IE Equipment" war sent to the following licensees on January 16, 1980, for action:

American Electric Power Service Corporation Indiana and Michigan Power Company D. C. Cook 1, 2 (50-315, 50-316)

Commonwealth Edison Company Dresden 1, 2, 3 (50-10, 50-237, 50-249)

Quad-Cities 1, 2 (50-254, 50-265)

Zion 1, 2 (50-295, 50-304)

Consumers Power Company Big Rock Point (50-155)

Palisades (50-255)

Dairyland Power Cooperative LACBWR (50-409)

Iowa Electric Light & Power Company Duane Arnold (50-331)

Northern States Power Company Monticello (50-263)

Prairie Island 1, 2 (50-282, 50-306)

Toledo Edison Company

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Davis-Besse 1 (50-346) 8007010 M

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Those on Attached List 2-Wisconsin Electric Power Company-Point Beach 1, 2 (50-266, 50-301)-

Wisconsin Public Service Corporation Kewaunee (50-305) l l

(L D rothy E.

arroll, Acting Chief Administrative Branch

Attachment:

IE Bulletin No.79-01B t

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7, Addressees - Memorandum dated January 18, 1980 IE Files MPA/DTS Central Files Reproduction Unit NRC 20b e

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UNITED STATES SSINS No.: 6820

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NUCLEAR REGULATORY COMMISSION Accessions No.:

OFFICE OF INSPECTION AND ENFORCEMENT 7910250528 WASHINGTON, D.C.

20555 January 14, 1980 IE Bulletin No.79-01B ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT Description of Circumstances:

IE Bulletin No. 79-01 required the licensee to perform a detailea review of the environmental qualification of Class IE electrical equipment to ensure that the equipnent will function under (i.e. during and following) postulated accident conditions.

The NRC staff has completed the initial review of licensees' responses to Bulletin No. 79-01. Based on this review, additional information is needed to facilitate completion of the NRC evaluation of the adequacy of environmental qualification of Class IE electrical equipment in the operating facilities.

In addition to requesting more detailed information, the scope of this Bulletin is expanded to resolve safety concerns reitting to design basis environments and current qualification criteria not addressed in the facilities' FSARS. These include high energy line breaks (HELB) inside and outside primary containment, lying, and submergence., " GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS", provides the guidelines and criteria the staff will use in evaluating the adequacy of the licensee's Class IE equipment evaluation in response to this Bulletin.

In ger.eral, the reporting problems encountered in the original responses and the additional information needed can be grouped into the following areas:

1.

All Class IE electrical equipment required to function under the postulated accident conditions, both inside and outside primary contain-

's cent, was not included in the responses.

2.

In many cases, the specific information requested by the Bulletin for each component of Class IE equipment was not reported.

3.

Different methods and/or formats were used in providing the written evidence of Class IE electrical equipment qualifications.

Some licensees used the System Analysis Method which proved to be the most effective approach. This method includes the following information:

Identification of the protective plant systems required to function

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under postulated accident conditions. The postulated acciden't conditions are defined as those environmental conditions resulting from both LOCA and/or HELB inside primary containment and HELB outside the primary containment.

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January if, 1980 IE Bulletin No. 79-018 Page 2 of 3 i.'

Identification of the Class IE electrical equipment hans within b.

each of the systems identified in Item a, that are required to function under the postulated accident conditions.

The correlation between the environmental data requirements specified c.

in the FSAR and the environmental qualification test data for each Class IE electrical equipment item identified in Item b above.

t Additional data not previously addressed in IE Bulletin No. 79-01 are needed 4.

to determine the adequacy of the environmental qualification of Class IE These data address component aging and operability in electrical equipment.

a submerged condition.

Action To Be Taken By Licensees Of All Power Reactor Facilities With An Operating License (Except those 11 SEP Plants Listed on Enclosure 1)

Provide a " master list" of all Engineered Safety Feature Systems (Plant Pro-1.

tection Systems) required to function under postulated accident conditions.

Accident conditions are defined as the LOCA/HELB inside containment, and HELB outside containment. For each system within (including cables, EPA's terminal blocks, etc.) the master list identify each Class IE electrical equipment item that is required to function under accident conditions.

Pages 1 and 2 of Enclosure 2 are standard formats to be used for the " master l

list" with typical information included.

Electrical equipment items, which are components of systems listed in Appendix A of Enclosure 4, which are assumed to operate in the FSAR safety analysis and are relied on to mitigate design basis events are considered within the scope of this Bulletin, regardless whether or not they were classified as part of the engineered safety features when the The necessity for further j

plant was orginally licensed to operate.

i up grading of nonsafety-related plant systems will be dependent on the outcome of the licensees and the NRC reviews subsequent to THI/2.

For each class IE electrical equipment item identified in Item 1, provide 2.

written evidence of its environmental qualification to support the capability of the item to function under postulated accident conditions.

For those class IE electrical equipment items not having adequate qualifica-t tion data available, identify your plans for determining qualifications of these items and'your schedule for completing this action Provide this in the format of Enclosure 3.

a For equipment identifed in Items 1 and 2 provide service condition profiles 3.

These data (i.e., temperature, pressure, etc., as a function of time).

.should be provi.ied for design basis accident conditions and qualification tests performed. This data may be provided in profile or tabular form.

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IE Bulletin No. 79-018 January 14, 1980 Page 3 of 3 4.

Evaluate the qualification of your Class IE electrical equipment against the guidelines provided in Enclosure 4., " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,"

provides supplemental information to be used with these guidelines.

For the equipment identified as having " Outstanding Items" by Enclosure 3, provide a detailed " Equipment Qualification Plan." Include in this plan specific actions which will be taken to determine equipment qualification and the schedule for completing the actions.

5.

Identify the maximum expected flood level inside the primary containment resulting from postulated accidents. Specify this flood level by elevation such as the 620 foot elevation.

Provide this information in the format of.

6.

Submit a " Licensee Event Report" (LER) for any Class IE electrical equipment item which has been determined as not being capable of meeting environmental qualification requirements for service intended.

Send the LER to the approp-riate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification.

If plant opera-tion is to continue following identification, provide justification for such operation in the LER.

Provide a detailed written report within 14 days of identification to the appropriate NRC Regional Office. Those items which were previously reported to the NRC as not being qualified per IEB-79-01 do r.ot require an LER.

7.

Cocplete the actions specified by this bulletin in accordance with the following schedule:

(a) Subcit a written report required by Items 1, 2, and 3 within 45 days from receipt of this Bulletin.

(b) Submit a written report required by Items 4 and 5 within 90 days from receipt of this Bulletin.

This information is requested under the provisions of 10 CFR 50.54(f).

Accordingly, you are requested to provide within the time periods specified in Items 7.a and 7.b above, written statements of the above information, signed under oath or affirmation.

Submit the reports to the Director of the appropriate NRC Regional Office.

Sand a copy of your report to the U.S. Nuclear Regulatory Commission, Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C.

20555.

Approved by GAO, B180225 (R0072); clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.

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IE Bulletin No. 79-018 Enclosure Januaryfy,1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.

79-28 Possible Malfunction of 12/7/79 All power reactor Namco Model EA 180 Limit facilities with an Switches at Elevated OL or a CP Temperatures 79-27 Loss Of Non-Class-1-E 11/30/79 All power reactor facilities holding Instrumentation and Control Power System Bus OLs and to those During Operation nearing licensing 79-26 Boron Loss From BWR 11/20/79 All BWR power reactor Control Blades facilities with an OL 79-25 Failures of Westinghouse 11/2/79 All power reactor BFD Relays In Safety-Related facilities with an r

Systems OL or CP 79-17 Pipe Cracks In Stagnant 10/29/79 All PWR's with an (Rev. 1)

Borated Water System At OL and for information PWR Plants to other power reactors 79-24 Frozen Lines 9/27/79 All power reactor facilities which have either OLs or cps and are in the late stage of construction 79-23 Potential Failure of 9/12/79 All Power Reactor Emergency Diesel Facilities with an Generator Field Operating License or 1

Exciter Transformer a construction permit 79-14 Seismic Analyses For 9/7/79 All Power Reactor (Supplement 2) As-Built Safety-Related Facilities with an Piping Systems 0:. or a CP 79-22 Possible Leakage of Tubes 9/5/79 To Each Licensee I

of Tritium Gas in Time-who. Receives Tubes pieces for Luminosity of Tritium Gas Used in Timepieces for Luminosity

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SEP Plants Region p7 pg 111 t W ien 1 I

Yankee Rowe III Big Rock Point Y

San Onofre 1 I

Haddam Neck III Lacrosse I

Oyster Creek I

R. E. Ginna III Dresden 2 I

Millstone III Palisades f

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Facility: XYZ Enc 1:sure 2 Docket No.: 50-XII*

MASTER LIST

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(Class IE Electrical Equipment Required to Function

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Under Postulated Accident Conditions)

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SYSTEM: RESIDUAL HEAT REMOVAL (RHR) i COMPONENTS Location l

Plant Identification Inside Primary Outside Primary Nunber Generic Name Containment Containment 1PT 456 PRESSURE TRANSMITTER x

1LT 594 LEVEL TRANSMITTER x

1LS 210 LIMIT. SWITCH x

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SYSTEM: AUT0KATIC DEPRESSURIZATION SYSTEM (ADS) 1 COMPONENTS 1

Location l

Plant Identification Inside Primary Outside Primary Nur.ber Generic Name Containment Containment

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B21-R001 VALVE MOTOR OPERATOR x

B21-F003 50LEN0I0 VALVE x

B21-F010 PRESSURE SWITCH x

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l,*,,,, III. SYSTEM: RHREhUIPMENT/ COMP 0NENTS(Typical) j

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Location Plant Identification Inside Primary -

Outside Primary Number

  • Generic Name Containment Containment 16xP455 0-RING GASKET x

EPA, Class E.

Westinghouse,100C ELECTRICAL PENETRATION ASSEMBLY X KULKA No. ET35 TERMINAL BOARD x

ONK0 NITE,1000V, 3C Black POWER CABLE x

x 1

X BRAND 10W-40 LUBRICATE OIL x

15 KB69 (Boston Wire & Cable)

INSTRUMENTATION CABLE x

x Cutler Hamer TB TERMINAL BOX x

No. 6 RAYCHEM XYZ CABLE SPLICE x

x Scotch No. 54 INSULATING TAPE x

T&B No.10 INSULATE ] TERMINAL LUG x

i Y Brand Epoxy No.,

SEALANT x

x 111 I

  • When a component is not identified by plant identification number, use the manufacturer, model number, serial number, etc.
    • Like components may be referenced.

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SYSTEM COMPONENT EVALUATION WORK SHEET INSTRUCTIONS

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1.

Equioment

Description:

Provide the specific information requested for each Class. IE electrical component.

Provide component location, specific information such as the building, access floor elevations, and whether In addition, provide the component is above the flood level elevation.

the specified and demonstrated accuracies of all instruments for their

Cables, trip functions and/or post accident scattoring requirements.

EPA's, terminal blocks, and other items shall be identified as part of the engineered safety features systems.

List values for each environmental parameter indicated.

Environment:

2.

List the " specification values" obtained from postulated accident analysis in the " SPEC" column.

List the " qualification values" obtained from test data, etc. in the " Qual" column. Tempera-reports, engineering. analysis ure, pressure, etc., as a function of time shall be provided in profile cr tabular form. Specify the time period that the component or equipment

!s required to function and identify the dacument which provides the basis for this time interval.

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It is expected that come listed parameters were not requested of the 1

lice::see at the time of their license issuance.

Address each parameter condition during this review.

If it is determined that a parameter such as submergence or a service condition such as aging was not previously considered, identify it as an " Outstanding Item."

3.

Docu::entation

Reference:

Reference the documents from which information in the " Spec" column.

Identify the document, paragraph, was obtained etc., that contains the postulated accident environmental specification da.a. In the " Qual" column identify the document, paragraph, etc., that contains the environmental qualification mata.

Qualification Method:

Identify the method of qualification.

To describe t.

4.

the qualification method use words such as simultaneous test, comparison Words test, sequential test, and/or engineering / mathematical analysis.

such as " test" and/or " analysis" when used alone do not adequately identify the qualification method.

5.

Dutstanding ItMns:

Identify parameters for which no qualification data is presently available. Also, identify parameters, service conditions, or environments not previously addressed during FSAR environmental qualification analysis such as submergence, qualified life (aging), or HELB.

Identify in the " Notes" section on page 1 of this enclosure the

. actions planned for determining qualification and the schedule for completing these actions.

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Facility SYSTEM COMPONENT EVALUATION WORK SHEET

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  • I ENVIRONMENT DOCUMENTATION REF*

QUALIFICATI ON OUTSTANDI NG EQUIPMENT DESCRIPTION Specifi-Quali fi-Specifi-yualiti-METHOD ITEMS Parameter catinn ra tinn ca tion cation System: RHR Operating 15 min.

300 min.

1 5

Simultaneous None i

Plant ID No. IPT456 Time Test Component:

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Simultaneou!

PRESSURE TRANSMITTER -

Temgerate SEE ACCIDENT AND Test None jl

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TEST PROFILES Manufacture:

PROVIDED l

Fischer-Porter Co.

Pressure (PSIA) 1 5

Simultaneous None Model Number:

Test 50-EN-1071-BCXN-NS Relative Function:.

Humidity (%)

100%'

100%

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Simultaneou!

None Test y

Accident Monitoring i q Chemical N 00 #

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See Note 1

ij Accuracy
Spec: 5%

Spray NA0H

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Demon: 4%

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Sequential Service. RHR % 1A Radiation 4x10 rads 1.2x10 rad ;

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Test None r

Discharge-Pressure S/N107

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Location: Containment Aging 40 yrs '

40 yrs 3

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2. En na' ysi Flood Level Eley: 620' 3 Not Not None Abova Flood Level: Yes l Submergence Required Required See Note 2 No x'

b Notes:

  • Documentation

References:

1.

XYZ Letter No. 237-1, dated Novamber 2, 1979,

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1.

FSAR Chapter 3. Paragraph 3.11 has been sent to MFG. requesting the qualification -

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FSAR Chapter 14. Paragraph 14.2.3.1 informa tion.

If qualification not determined 3.

Technical Specification 3.4.1, Paragraph A 1

acceptable by December 15, 1979, component 4.

Technical Specification 4.6.5, Paragraph B will be replaced during refueling outage March 1980.

Fischer and Porter Co.'$00 dated November 2,1972 FIRL Test Report No. 3 5.

Test Report No. 2500-1 6.

7.

.A. B. D00 Engineering Evaluation Data. Report Nn. 6932 2.

In the FSAR submergence was not considered l

.8.

Wylie Laboratory Report.Ro. 467 an environmental parameter. ABC Laboratory is to perform submergence ' test in April 1980.

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4 Page 2 of Enclosure 3 TYPICAL SERVICE CONDITION PROFILES EXCEPTIONS POSTULATED QUALIFICATION EQUIPMENT ACCIDENT TEST ACCURACY ACCURACY OR l

DESCRIPTION ENVIRONMENT ENVIRONMENT REQUIREMENTS DEMONSTRATED REMARKS 4

NDTE 1 NOTE 2 NOTE 3 NOTE 4 NOTE 5 NOTE 6 NOTES:

1.

Refer to " Equipment Description" on Page 1 of this Enclosure.

2.

Provide sufficient values of temperature and pressure as a function of time in tabular form to draw a characteristic profile.

Provide sufficient values of temperature and pressure as a function of time for which equ'pment was qualified 3.

to draw a characteristic profile.

Present this information in tabluar form.

i Provide the accuracy requirements for sensors and transmitters for trip functions and/or post accident monitoring 4.

as used in the plant safety analysis.

5.

Provide the accuracy demonstrated by sensors and transmitters during the qualification test regarding the trip functions and/or post accident monitoring as applicable.

Identify any exception or deviation between specified service condition and qualification service condition and i

6.

justification to explain acceptance of deviation.

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GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION 0F CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS 1.0 Introduction 2.0 Dis cussion 3.0 Identification of Class IE Ecufoment 4.0 Service Conditions 4.1 Service Conditions Inside Containment for a Loss of Coolant Accident (LOCA) 1.

Temperature and Pressure Steam Conditions i

2.

Radiation 3.

Submergence 4.

Chemical Sorays 4.2 Service Conditions for a PWR Main Stean Line Break (MSLB)

Insice Containmen 1.

Temoerature and Pressure Steam Conditions 2.

Radiation 3.

Submergence 4

Chemical Sorays 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 Lona-Tert-Eme-cency Core Cooling Following a LOCA 1.

Temoerature, Pressure and Relative Humidity 2.

Radiation 3.

Submercence 4.

Chemical Sorays

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2-4.3.3 Areae Normally Mat.tained at Room Conditions 5.0 Qualification Methods 5.1 Selection of Qualification Method 5.2 Qualification by Type Testino 1.

Simulated Service Conditions and Test Duration 2.

Test Soecimen 3.

Test Seouence 4.

Test Specimen Aoing 5.

Functionai Testino and Failure Criteria 6.

Installation Interfaces 5.3 Qualification by a Combination of Methods (Test, Evaluation, Analysis )

6.0 Marcin r

7.0 Ag 8.0 Documentation Appendix A - Typical Equipment / Functions Needed for Mitigation of

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a LOCA or MSLB Accident

' Appendix B - Guidelines for Evaluating Radiation Service Conditions Inside Containment for a LOCA and MSLB Accident Appendix C - Thennal and Radiation Aging Degradation of Selected Materials

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GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION 0F CLASS IE ELECTRIC'LL EQUIPMENT IN OPERATING REACTORS 1.0 INTRODL'CTION Dn 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 IC Circular 78-08 was to initiate a review by the licensees to detemine whether preper docu.entation existed to verify that all Class IE electrical equipment would function as required in the hostile environment which could result fror design basis events.

The licer. sees' reviews are now essentially complete and the NRC ste ff has begur to evaluate the results. This document sets forth guidelined for the NRC staff to use in its evaluations of the ifceasees' responses te IE Bulletin 79-01 and selected associated qualification documentatior. The objective of the evaluations using these guidelines is to identif.f Class IE equipment whose documentation does not provide reasonable assuran:e of environ-mental qualification. All such equipment identified will then b>! subjected to a plant application specific evaluation to detemine whether st should be requalified or replaced with a component whose qualification has been adequately verified.

These guidelines 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.

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6, s Equipent in other classes of plants not yet licensed to operate, or replacerent equipment for operating reactors, may be subject to dfferent requirements such as those set ferth in NUREG-05BB, Interim Staff Position on Envircrrnental 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 deterrination will be rade as part of that related generic review.

2.0 DISCUSSION IEEE Std. 323-1974 is the current industry standard for environmental I

This standard was qualification of safety-related electrical equipment.

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 dhcument 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 comitted to comply with any particular industry standard for electrical equipment qualification. However, all of

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the operating reactors are required to comply with the General Design Criteria IEEE Std. 323-1974, "!EEE Standard for Qualifying Class IE Equipment for I

fluclear Power Generating Stations."

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-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 secomodate the affects of and to be compatible with the environmental conditions associated with nonnal 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 r'eactors are in compliance with General Design Critarion 4.

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3.0 IDEhTIFICATION 0~ CLASS IE E001PMENT Class IE equipment includes all electrical equipment needed to achieve

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emergency reactor shutdown, containment isolation, reactor core cooling, l

i 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 nost severe postulated loss of coolant accident (LOCA) and main steanline break accident (MSLB) are listed in Appendix A.

More detailed descript.ons 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).

The guidelines in tnis document are applicable to all components neces'sary for operation of the systems listed in Appendix A including but not limited to valves, actors, cables, connectors, relays, switches, transmitters and valva position indicators.

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4.0 SERVICE CONDITIONS In order to detemine 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 (LCCA) l.

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 ured: (.1} BWR Drywells - 3400F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and (21 PWR Ice Condenser Lower Compartments + 3400F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

2.

Radhtien - When specifying radiation service conditions for equipment l

exposed to radiation during nonnal operating and accident conditions, the nonnal operating dose should be added to the dose received during f

the course of an accident. Guidelines for evaluating beta and gamma radiation service conditions for general areas inside containment are provided below. Radiation service conditions for equipment located directly above the containment sump, in the vicinity of filters, or submerged in contaminated liquids must be evaluat'ed on a case by case basis. Guidelines for these evaluations are not provided in this docuent.

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.5 Gamma Radiation Doses - A total gama dose radiation service condition 7

of 2 x 10 RADS is acceptable for Class IE equiprr...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 perfomed to detemine if the dose specified is acceptable.

Procedures for evaluating radiation service conditions i

in such cases are provided in Appendix B.

The procedures in Appendix B are based on the calculation for a typical PWR report'ed in Appendix D of 10JREG45881 i

Ga.ma dose radiation service conditions for BWRs and PWRs with ice cender.ser containments must be evaluated on a case by case basis,

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Since the procedures in Appendix B are based on a calculation for a typical PWR with a dry type contatnment, 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 censervative dose assumptions and adjustment factors developed on a case by case basis, Beta Radiation Doses - Beta radia. tion doses generally are less significant than ganrna radiation doses for equipment qualification. This is due to the low penetratYng 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 transmitters, valve operators, containment penetrations), electrical cable is considered the most j

I 'JRES-0583, Interim Staff Position on Environmental Qualification of N

Safety,Related Electrical Equipment.

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6-vulnerable to damage from beta radiation. Assuming a TID 14844 source term, t'ie 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 8

beta surface dose of 1.40 x x 10 RADS reported in Appendix 0 of NUREG 0588 would be reduced by approximately a factor of ten w"hin 30 mils cf 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-8 RADS and considering tive unshielded surface beta dose of 2.0 x 10 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 gama dose to which an item of equipment has been qualified, then that equipment may be considered qualified for the total radiation environment (gama plus beta).

If this criterion is not satisfied the radiation service condition should be determined by the sum of the gama and beta doses.

3.

Submereence - The preferred method of protection against the effects cf s@.ergency is to locate equipment above the water flooding level.

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

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4.

Containment Sorays - Equipment exposed to chemical sprays should be qualifiedforthemostseverechemicalenvironment(acidicor

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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

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Equipment required to function in a steam line break environment must be qualified for the high temperature and pressure that could result.

In scrne 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.

Tencerature and Pressure Steam Conditions - Equipment qualified for a LO~A environment is considered qualified for a M5LB accident environ-ment in plants with automatic spray systems not subject to disabling sing:e component failures. This position is based on the "Best

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Estimate" ca:culation of a typical plant peak temperature and pressure and a therma'. analysis of typical components inside containment.M 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 spray systems subject to disabling s'.ngle failures or delayed initiation should be qualified for a MSLB accf Jent environment determined by a plant specific analysis. Acceptabl.e" methods ISee NUREG 0458, Short Term Safety Assessment on the Environmental Ouslification of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation.

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

2.

Radiation - Sane as Section 4.1 above except that a conservative 6

gamma dose of 2 x 10 RADS is acceptable.

3.

Submeroence - Same as Section 4.1 above.

4.

Chemical Sprays - Same as Section 4.1 abo ~e.

4.3 Service Conditions Outside 'of Containment 4.3.1 Areas Sub3e:t to a Severe Environment as a Result of a Hich Eneroy Line Break (HEL3)

Service conditions for areas outside containment expcsed to a HELB were

' evaluated on a plant by plant basis as part of a program initiated by the staff in De: ember,1972 to evaluate the effects of a HELB. The equipment required to mitigate the event was also identified. This equipnent should be qualified for the service conditions reviewed and approved in tne ti-i5 Sa'e:y Evaluation Report for each specific plant.

4.3.2 Areas Where Fluids are Retirculated from Inside Containment to Accomolish Lono-Term Core Coolino Followina a LOCA 1.

Tennerature and Relative Humidity - One hundred percent relative humidity shouTd be established as a service condition in confined spaces. The tecoerature and pressure as a function of tim e should be based on the p1, ant unique analysis reported in the FSAR.

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Radiation - Due to differences in equipment arrangement within l

these areas and the significant effect of this factor on doses, i

radiation service conditions must be evaluated on a case by case 6 RADS would be basis.

In general, a dose of at least 4 x 10 expected.

3.

Submergence - Not applicable.

4.

Chemical Sorays - Not applicable.

4.3.3 Areas Normally Maintained at Room Conditions Class IE equipcent located in these areas does nqt 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., AN31, NEMA, National Elec ric Code). Based on these factors, failures of equipment in these areas during a design basis event are expected to be random except to the extent that they may be due to aging or failures of air conditioning or ventilation systems. Therefore, no special consideration need be given to

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the environmental qualification of Class IE equipmen't in these areas provided the aging recuirements discussed in Section 7.0 below are satisfied and the areas are raintained at room conditions by redundant air conditioning or ventilation systems served by the onsite emergency electrical power system.

Equipment located ih 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 detemined from a plant specific analysis.

5.0 00A!.IFICATION METHODS I

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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 J

factors as: (1) the severity of the service conditions; (2) the structural and saterial 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 cualification for severe temperature, pressure, and steam service conditions for Class IE equipment should be based on type testing.

tQualification 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 Ouali'i:ation bv Tyoe Testino 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.

Tne time duration of the test should be at least as long as the period from the initiation of the accident until the temperature and pressure

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service conditions return to essentially the same levels that existed before the postulated accident. A shorter test duration may be accepthble t

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

2.

Test Soecimen - The test specimen s~ ould be the same model as the n

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 Seouence - 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 apolied at any time during the test sequence provided the component does not contain any materials which are known to be susceptible to sienificant radiation damage at the service condition levels or ma erials whose susceptibility to radiation damage is not known (see Ap;endix C).

If the component contains any such materials, the radiation dose should be applied prior to or concurrent with exposure to the elevated terperature and pressure steam / air environment. The same test specimen sh:uld be used throughout 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 Scecimen Aging - Tests which were successful using test specimens which had not been preaged may be considered acceptable provided the co:senen: does not contain materials which are known to be susceptible l

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to significant degradation due to thermal and radiation agir.. (see Section i

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 Testino and Failure Criteria - Operational modes tested should be representative of the actual application requirements (e.g., cor:ponents 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 ac:uracy recairements based on the rt.aximum 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 acuipment mounting and electrical or nechanical 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 h

inspection in the field to verify that equipment was installed 1

as it was tested. Particular emphasis should be placed on connon

. problems such as protective enclosures installed upside down with, drain holes at the top and penetrations in equipment housings for L

electrical connections being left unsealed or susceptible to l

aoisture incursion through stranded con,ductors.

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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 quali,fication 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 b'y a combination of methods.

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

1.

Radiatior Dualification - Some of the earlier tvoa tests eerformed for operating reactors did not include radiation as a service condition.

In these cases the equipment may be shown to be radiation qualified by performing a calculation of the dose 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

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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 corrosi on resistant cases (e.g., stainless steel) may be shown i

to be qualified for a chemical environment by an analysis of' the effects of the particular chemicals on the,, articular enclo-sure materials. The effects of chemical sprays on the pressure

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integrity of any gaskets or seals'present should be considered l

in the analysis.

_ :_.u._ _ 2.z 6.0 Marcin IEEE Std. 323-1974 & 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 nomal variations in cocinercial production i

of equipment and reasonable errors in defining satisfactory pr$formance.

Section 6.3.1.5 of the standard provides suggest,ed. factors to be applied l

to the service conditio*ns to assure adequate margins. The factor applied to the time equipment is requi' 3d 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 ba:is event. Therefore, if the guidelines in Section 4.0 and 5.2 are satisfied,no separate margin factors are required to be added to the service conditions when specifying test conditions.

A ing 7.0 l

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 fer all Class IE equipment is not sufficient to justify the expense for plants already constructed and operating. This position does not, however, exclude equipment G

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1 using materials that have been identified as being susceptible to significant degradation due to thernal and radiation aging. Component maintenance or replacement schedules should include considerations of the specific aging characteristics of the component materials. Ongoing progrars 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 day be found in nuclear power plants along with an indication of the material susceptability to thermal and radiation aging.

8.0 Documentation

. Car.plete and auditable records must be available for qualification by any of the methods described in Section S.O above to be considered valid.

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These records should describe the qualification method in sufficient detail to veri.fy that all of the guidelines have been satisfied.

A simple vendor certification of compliance with a design spe:ification should not be considered adequate.

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-l APPENDIX A TYPICAL EQUIPMENT / FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR P.5LB ACCIDENT Engine &ed Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power l

Emergency Core Cooling Contairment Heat Removal Contairment Fission Product Removal Contairment Combustible Gas Control Auxiliary Feedwater Contairment Yantilation Contain. ment Radiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)

Ventilation for Areas Containing Safety Equipnent Component Coolir.g Service Water 2

Emergency Shutdcwn 3

Post Ac:1 dent Sampling and Monitoring 3

Radiation Monitoring 3

Safety Related Display Instranentation

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IThese systems will diff'er for PWRs and BWRs, and for oldar and newer plants.

In each case the system features which allow fo transfer to recirculation cooling mode and establishment of long term cooling with boren 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 resJit in a breach of the reactor coolant pressure boundary together i

with c rapid depressurization of the reactor coolant system.

Examples i

of such systems and equipment are the RHR system, PORVs, RCIC, pressurizer spr.=ys, chemical and volume control system, and steam dump systems.

Mort specific identification of these types of equipment 'can be found 3

in.he plant emergency procedures.

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APPENDIX B PROCEDURES FOR EVALUATING GAMMA RADIATION SERVICE CONDITIOh5 Introduction and Discussion 1}m 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 Ap~pendix 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 NURES 0588 apply to these procedures.

(2) The sample ca,1culation in Appendix D of NUREG-0588 is for a 4,000 6 ft3 MWth pressurized water reactor housed in a 2.52 x 10 contain-nent with an iodine scrubbing spray system. A similar calculation without iodine scrubbing sprays would increase the dose to equipment 7

approximately 15..

The conservative dose of 2 x 10 RADS assumed

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

(3) Shielding calculations are based on an average gama 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 beilir.g water reactors or other containment types.

However, 1

deses for these other plant configurations may be evaluated using similar procedures wif.h conservative dose assumptions and adjustment factors developed on a case by case basis.

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Pr:cedure 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) cocpartment volume; and (5) time equipment is required to remain functional.

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"s, The procedure for using the figures is best illustrated by an example.

Coisider the following case. The radiation service condition for a l

6 particular item of equipment has been specified as 2 x 10 RADS. The 1

application specific parameters are:

Reactor power level - 3,000 MWth 3

6 ft Containment volume - 2.5 x 10 3

Compartment Volume - 8,000 ft Thickness of compartment shield wall (concre'te) - 24" Time equipment is required to remain functional - 1 hr.

The prcblem 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.

Steo 1 En er the nomogram in Figure 1 at 3,000 MWth reactor power level and 2.5 x 10 ft3 containment volume and read a 30-day integrated dose of 6

7 1.5 x 10 RADS.

Steo 2 7

Enter Figure 2 at a dose of 1.5 x 10 RADS and 24" of concrete shielding 4

for the compartment the equipment is located in and read 4.5 x 10 RADS.

This is the dose the equipment receives from sources outside the compart-mer.t. To this mu'st be added the dose fmm sources inside the compartment (Step 3).

Steo 3 3

En'ter Figure 3 at 8,000 ft and read a correction factor of 0.13. The dose due to sources inside the compartment would then be 'O.13 (1.5 x 10 )

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7 6 RADS. Th? sums of the doses from steps 2 and 3 equals:

= 1.95 x 10 7

6 RADS 4.5 x 104 RADS + 0.13 (1.5 x 10 ) RADS = 2.0 x 10 4

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Steo 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 correction factor of 0.15. Apply this factor to the sum of the doses detemined 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 />, t

6 5

0.15 (R.0 x 10 ) = 3 x 10 RADS 6

In this particular example the service condition of 2 x 10 RADS specified is conservative with respect to the estimated dose of 3 x 5

10 RA:5 calculated in steps 1 through 4 and 'is, therefore, acceptable.

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FIGURE 1 NOMOGRAM FOR CONTAINMENT VOLUME AND REACTOR POWER LOCA DOSE CORRECTIONS

  • CONTAINMENT -

VOLUME (ft3) 3 x 106 2 x 105 30 DAY MWTH INTEGRATED yDOSE 4000 -

1 x 105 7

3000 4 x 10 2000 3 x 107

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2 x 10 g

3 x 105 200 1 x 107 2 x 105 1 x 105 5 x 108 4 x 106

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  • MSLB ACCIDENT DOSES SHOULD SE READ AS A FACTOR OF 10 LESS

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DOSE CORRECTION FAdi65YOR CONCRETE SHIELDING

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APPENDIX C TkERMAL 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 themal aging.

0 Susceptibility to significant themal aging in a 45 C 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 tems used to characterize the dose effect is as follows:

Threshold - Refers to damage threshold, which is the radiation a

exposure required to change at least one physical property of the material.

Percent Change of Property - Refers to the radiation exposure e

required to change the physical property noted by the percent.

Allowable - Refers to the radiation which can be absorbed before e

serious degradation orsars.

The infomation in this appendix is based on a literature search of sources including the National Technical Information Service (NTIS), the National

' Aeronautics and Space Adninistration's Scientific and Technical Aerospace Report (STAR), NTIS Govemment Report Announcements and Index (GRA), and 4

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2 various manufacturers data reports. The materials list is not to be considered all inc.lusive 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

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due to radiation or thermal aging.

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

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11/14/79 TABLE C-1 2

THERMAL AND RADIATION AGING DEGRADATION 0F SELECTED MATERIALS l

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l NUREGWA8 For Comm.ent Interim Staff Position on Environmental Qua'ification of Safety-Re ated Electrical Equipment Resolution of Generic Technical Activity A-24 Manuscript Completed: August 1979 Data Published: Demmber 1979 A. J. Sruklewicz, Task Manager Division of Systems Safety Offico of Nuclear Reai: tor Regulation i

U.S. Nuclear Regulatory Commission Washington, D.C. 20565 f* "%

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