NRC Generic Letter 1980-05

NRC Generic Letter 1980-005, Submittal of IE Bulletin 1979-018: Environmental Qualification of Class IE Equipment

ML031350283
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Issue date:	01/14/1980
From:	Grier B H NRC Region 1
To:
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-nr, BL-79-001B GL-80-005, NUDOCS 8001290369
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UNITED STATES NUCLEA*'REGULATORY

COMMISSION

REGION I 631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA

19406 OL -go-_g Docket Nos. 50-03 50-247 JAN 1 4 1980 Consolidated Edison Company of New York, Inc.ATTN: Mr. W. J. Cahill, Jr.Vice President 4 Irving Place New York, New York 10003 Gentlemen:

Enclosed is IE Bulletin 79-OIB which requires action by you with regard to your power reactor facility with an operating license.Should you have questions regarding this Bulletin or the actions required of you, please contact this office.Sincerely, Boyce H. Grier Director Enclosures:

1 IE Bulletin No.79-01B with Attachments

2. List of Recently Issued IE Bulletins

CONTACT

S. 0. Ebneter (215-337-5296)

cc w/encls: L. 0. Brooks, Project Manager, IP Nuclear W. Monti, Manager -Nuclear Power Generation Department M. Shatkouski, Plant Manager J. M. Makepeace, Director, Technical Engineering W. D. Hamlin, Assistant to Resident Manager (PASNY)J. 0. Block, Esquire, Executive Vice President

-Administration Joyce P. Davis, Esquire 80012 90 Aw -

ENCLOSURE

1 UNITED STATES SSINS No.: 6820 NUCLEAR REGULATORY

COMMISSION

Accessions No.: OFFICE OF INSPECTION

AND ENFORCEMENT

7910250528 WASHINGTON, D.C. 20555 IE Bulletin No. 79-O1B Date: January 14, 1980 ENVIRONMENTAL

QUALIFICATION

OF CLASS IE EQUIPMENT

Description of Circumstances

IE Bulletin No. 79-01 required the licensee to perform a detailed review of the environmental qualification of Class IE electrical equipment to ensure that the equipment 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 relating 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 contain-ment, aging, and submergence.

Attachment

4, "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 general, 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 containment, 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:

a. Identification of the protective plant systems required to function under postulated accident conditions.

The postulated accident conditions are defined as those environmental conditions resulting from both LOCA and/or HELB inside primary containment and HELB outside the primary containment.

Enclosure

1 IE Bulletin No. 79-QIB Date: January 14, 1980 b. Identification of the Class IE electrical equipment items within each of the systems identified in Item a, that are required to function under the postulated accident conditions.

c. The correlation between the environmental data requirements specified in the FSAR and the environmental qualification test data for each Class IE electrical equipment item identified in Item b above.4. Additional data not previously addressed in IE Bulletin No. 79-01 are needed to determine the adequacy of the environmental qualification of Class IE electrical equipment.

These data address component aging and operability in 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 Attachment

1)1. Provide a "master list" of all Engineered Safety Feature Systems (Plant Protection 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 Attachment

2 are standard formats to be used for the "master list" with typical information included.Electrical equipment items, which are components of systems listed in Appendix A of Attachment

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 plant was originally licensed to operate. The necessity for further up grading of nonsafety-related plant systems will be dependent on the outcome of the licensees and the NRC reviews subsequent to TMI/2.2. For each class IE electrical equipment item identified in Item 1, provide written evidence of its environmental qualification to support the capa-bility of the item to function under postulated accident conditions.

For those class IE electrical equipment items not having adequate qualifica- 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 Attachment

3.3. For equipment identifed in Items 1 and 2 provide service condition profiles (i.e., temperature, pressure, etc., as a function of time). These data should be provided for design basis accident conditions and qualification tests performed.

This data may be provided in profile or tabular form.

Enclosure

1 IE Bulletin No.79-01B Date: January 14, 1980 4. Evaluate the qualification of your Class IE electrical equipment against the guidelines provided in Attachment

4. Attachment

5, "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 Attachment

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 Attachment

3.6. 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 appropriate 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 operation is to continue following identification, provide justifi-cation 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 not require an LER.7. Complete the actions specified by this bulletin in accordance with the following schedule: (a) Submit 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.Send 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.

Enclosure

1 Approved by given under IE Bulletin No.79-01B Date: January 14, 1980 GAO, B180225 (R0072); clearance expires 7/31/80. Approval was a blanket clearance specifically for identified generic problems.Attachments:

1. List of SEP Plants 2. Master List Standard Format, Typical 3. System Component Evaluation Work Sheet 4. Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors 5. Interim Staff Position on Environmental Qualification of Safety-Related Equipment (To

Addressees

Only)

Attachment

1 to IE Bulletin 79-O1B SEP Plants Plant Region Dresden 1 III Yankee Rowe I Big Rock Point III San Onofre 1 V Haddam Neck I LaCrosse III Oyster Creek I R. E. Ginna I Dresden 2 III Millstone

1 I Palisades III

Facility:

XYZ ---.Dpcket.No.:

50-XXX .MASTER LIST--.- Attachment

'lo.-=. ->-: <t>;=m .- :~tgyp~(Typical').Pg1 f_:--.. --- -~ <C1 ass._IE Electricai Equipment Required to Function-:--Under.Postulated Accident Conditions).

.;I. SYSTEM: RESIDWUAL-HEAT

REMOVAL (RHR)-- ~.:--:.......................;:

.2 -to.E IE. Bull1et in. 79-OIB COMPONENTS

Location Plant-Identification Inside Primary Outside Primary Number Generic Name Containment Containment IPT 456 -PRESSURE

TRANSMITTER

x ILT 594 LEVEL TRANSMITTER

x.S 210 LIMIT SWITCH x II. SYSTEM: AUTOMATIC

DEPRESSURIZATION

SYSTEM (ADS)COMPONENTS

..~Locatilon-.

Plant Identifcation Inside Primary Outside Primary.Nuber Generic Name Containment Containment B21-ROOI VALVE MOTOR OPERATOR x B21-F003 -SOLENOID

VALVE x B21-FOlO PRESSURE SWITCH .x II. SYSTEM. RHR EQUIPMENT/COMOI1NENTS(Typical)

Attachment No.**COMPONENTS'.-

2 to IE Bulletin 79-01B l .k.__________________________________________________________________________

I Plant Identification Number*4 16xP455 O-RING GASKET x*EPA,- Clas~ E, Westinghouse:

E OOC ELECTRICAL

PENETRATION

ASSEMBLY X KULKA No. ET35 TERMINAL BOARD x ONKONITE, lOOOV, 3C Black POWER CABLE x x X BRAND 10W-40 LUBRICATE

OIL x 15 KB69 (Boston Wire & Cable) INSTRUMENTATION

CABLE x x Cutler Hamner TB TERMINAL BOX x N o .-6_ _ _ _ _ _ _ _ _ _ _RAYCHEM XYZ CABLE SPLICE x x Scotch No. 54 INSULATING

TAPE x T&B No. 10 INSULATE TERMINAL LUG x Y Brand Epoxy No;. SEALANT x x.ll ._________________________

• When a component is manufacturer, model** Like components may not identified number, serial be referenced.

by plant identification number, use the number, etc.

' Facility: Unit: D ocket: SYSTEM COMPONENT

EVALUATION

WORK SHEET (Typical)Attachment No. 3 to IE Bulletin No. 79-OIB Page I of 3 t'EfIVI RONMENT DOCU1MENTATION'REF*

QALFCTOOTTND

EQUIPMENT

DESCRIPTION

QUALIFICATION

OUTSTANDI pec if- ua li- Specifi- ualiti- METHOD ITEMS Pa -arameter iDra tnn -catin nn ._System: RHR Operating

15 min. 300 min. 5 Simultaneou!

None Plant ID No. IPT456 Time Test Component Temperature SEE ACCIDENT AND 5 Simultaneou!

PRESSURE TRANSMITTER.

S EST PROFILESTAN ( ) TEST PROFILES .Test None Manufacture:

PROVIDED : Fischer-Porter Co. Pressura o (PSIA) , 1 5 Simultaneou None Model Number: Test 50-EN-1071-BCXN-NS

Relative Functlon:

Humidity(%)

100% 100% 1 5 Simultaneou None Accident Monitoringi.

ii __- ' _ Test , Chemical N 3 B0 3/Accuracy:

Spec: 5% Spray NAOH 1 See Note 1 Demon: 4% NO Servi ce: RHR Pump lA 6Radiaton

4xl0 6 rads l.2xlO 8 rad 2 6 Sequential Discharge Pressure Test None S/NiO7 1 1. Seq4entf Nn Location:

Containment Aging yrs 40 yrs 3 7, 8 Test ysNone Flood Level Elev: 620' Not Not None Above Flood Level: Y Yes lSubmergence Required Required See Note 2 N o x 'j_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _I'IG;C (-uocumentation References:, Nbtes: 1.2.3.4.5.6.7..8.'tSAR Chapter 3, Paragraph

3.11 FSAR Chapter 14, Paragraph

14.2.3.1 Technical Specification 3.4.1, Paragraph A Technical.

Speciffcation

4.6.5, Paragraph B FIRL Test Report No. ?O00 dated November 2, 1972 Fischer and Porter Co. Test Report No. 2500-1 A. 0. DOD Engineering Evaluation Data.Report No. 6932 Wylie Laboratbry Report.Ro.

467 1. XYZ Letter No. 237-1, dated November 2, 1979, has been sent to MFG. requesting the qualification information.

If qualification not determined acceptable by December 15, 19791, component will be replaced during refueling outage March 1980..,.I .2. In the FSAR submergence was not considered an environmental parameter.

ABC Laboratory is to perform submergence test in April 1980..I

Attachment

3 to IE Bulletin 79-OIB SYSTEM COMPONENT

EVALUATION

WORK SHEET INSTRUCTIONS

1. Equipment 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 the component is above the flood level elevation.

In addition, provide the specified and demonstrated accuracies of all instruments for their trip functions and/or post accident monitoring requirements.

Cables, EPA's, terminal blocks, and other items shall be identified as part of the engineered safety features systems.2. Environment:

List values for each environmental parameter indicated.

List the specification values" obtained from postulated accident analysis in the "SPEC" column. List the "qualification values" obtained from test reports, engineering analysis data, etc. in the "Qual" column. Tempera-ture, pressure, etc., as a function of time shall be provided in profile or tabular form. Specify the time period that the component or equipment is required to function and identify the document which provides the basis for this time interval.It is expected that some listed parameters were not requested of the licensee 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. Documentation Reference:

Reference the documents from which information was obtained in the "Spec" column. Identify the document, paragraph, etc., that contains the postulated accident environmental specification data. In the "Qual" column identify the document, paragraph, etc., that contains the environmental qualification data.4. Qualification Method: Identify the method of qualification.

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

Words such as "test" and/or "analysis" when used alone do not adequately identify the qualification method.5. Outstanding Items: Identify parameters for which no qualification data is presently available.

Also, identify parameters, service conditions, or environments not previously addressed during FSAR environmental quali-fication analysis such as submergence, qualified life (aging), or HELB.Identify in the "Notes" section on page 1 of this attachment the actions planned for determining qualification and the schedule for completing these actions.

Attachment

3 of IE Bulletin 79-010 EQUIPMENT DESCRIPTION

NOTE 1 POSTULATED

ACCIDENT ENVIRONMENT

NOTE 2 TYPICAL-2-SERVICE CONDITION

PROF QUALIFICATION

TEST ENVIRONMENT

NOTE 3 ACCURACY ACCURACY REQUIREMENTS

DEMONSTRATED

NOTE 4 NOTE 5 EXCEPTIONS

OR REMARKS 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.3. Provide sufficient values of temperature and pressure as a function of time for which equipment was qualified to draw a characteristic profile. Present this information in tabular form.4. Provide the accuracy requirements for sensors and transmitters for trip functions and/or post accident monitori(-

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.

6. Identify any exception or deviation between specified service condition and qualification service condition and justification to explain acceptance of deviation.

.Attachment No. 4 to6 3ulTetin 1--01B- 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 SDrays 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 aHighEnergy Line Break (HELB)4.3.2 Areas Where Fluids are Recirculated From Inside C ainment to Accom'lish Lona. "er e Core Coolina Following a LOCA 1. Temoerature, Pressure and Relative Humidity 2. Radiation 3. Submercence

4. Chemical SDrays

.tAttachment No. 4 to IE Bulletin 79-01B'. -2-4.3.3 Areas Normally Mat--.talned at Room Conditions

5.0 Qualification Methods 5.1 Selection of Qualification Method 5.2 Qualification by Type Testing-l. 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 Acina 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 Attachment No. 4 to IE Bulletin 79-01B 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 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.

Attachment No. 4 to IE Bulletin 79-01B 2 Equipment 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 1 IEEE Std. 323-1974, 'IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."

.'*. .. -Attachment No. 4 to IE Bulletin 79tO1B* specified in Appendix A of 10 CFR 50. General Design Criterion 4 states in part that structures, systems and components important to safetS 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 severe postulated loss of coolant accident (LOCA) and main steanline 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 CSection 4.0).The guidelines in this document are applicable 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, Attachment No. 4 to IE Bulletin 79-O1B -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 q 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: (l11BWR Drywells .340 0 F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and C21 FWR Ice Condenser Lower Compartments

-340 0 F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.2.. ?adiation

-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 conditions for equipment located directly above the containment sump; in the vicinity of filters, or-submerced in contaminated liquids must be evaluated on a case by case basis, Guidelines for these evaluations are not provided in this document.,

, Attachment No. 4 to IE Bulletin 79-O1B Ganma Radiation Doses -A total gamma dose radiation service condition of 2 x 10 7 RADS is acceptable for Class IE equipm..at 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 XUREG-.0588

1 Ga6nna 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 containment, 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 basIs, Bet.a Radiation Doses -Beta radiation doses generally are less significant than gama 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 (etg,, cables, instrument transmitters, valve operators, containment penetrations), electrical cable is considered the most 1 NUkE-0588, Interim Staff Position on Environmental Qualification of SafetyRelated Electrical Equipment.

Attachment No. 4 to IE Bulletin 79-OIB -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 0 of NUREG 0588 would be reduced by approximately a factor of ten within 30 mils of the sur face of electrical cable insulation of unit density. An additional

40 mils of insulation (total of 70 mils) results in another actor 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 106 of the tota' garma 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 garma and beta doses.3. Submercence

-The preferred method of protection against the effects of submEergency 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 Attachment No. 4 to IE Bulletin 79-O0B 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. Ter.Derature and Pressure Steam Conditions

-Equipment qualified for a LOCA environment is considered qualified for a MSLB accident environ-rer.t in plants with automatic spray systems not subject to disabling single component failures.

This position is based on the 'Best Estim.at'e calculation of a typical plant peak temperature and pressure and a therma' analysis of typical components inside containment.

1/The 'inal 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 single failures or delayed initiation should be qualified for a MSLB accident environment determined by a plant specific analysis.

Acceptable methods See NUR E 0456, Short Term Safety Assessment on the Environmpntal Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation.

Attachment No. 4 to IE Bulletin 79-O1B 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 Eletctrical Equipment.

2. Radiation

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

3. Submercence

-Same as Section 4.1 above.4. Chemical Sprays -Same as Section 4.1 above.4.3 Seruice 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 HELB were evaluated on a plant by plant basis as part of a program initiated by the staff in Dece.mber, 1972 to evaluate the effects of a HELB. The equipment required to mitigate the event was also Identified.

This equipment should be qualified for the service conditions reviewed and approved n tne i.-. Sa-ezy Evaluation Report. for each specific plant.4.3.2 Areas Where Fluids are Recirculated from Inside Containment to Accomplish Lona-Temn Core Coolino Followina a LOCA 1. Termerature and Relative Humidity -One hundred oercent relative humidity shouTd be established as a service condition in confined spaces. The temoerature and pressure as a function of time should be based on the plant unique analysis reported in the FSAR.

Attachment No. 4 to IE Bulletin 79-O1B 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 Sorays -Not applicable.

4.3.3 Areas Normally Maintained at Room Conditions Class IE equipment located in these areas does not 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 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 the environmental qualification of Class IE equipment in these areas provided the aging requirements discussed in Section 7.0 below are satisfied and the areas are maintained at room conditions by redundant air conditioning or ventilation systemis served by the onsite emergency electrical power system.Equip.ent located irf 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 specific analysis.5.0 QJALIFICATION

METHODS

Attachment No. 4 to IE Bulletin 79-OB lo: V-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 cualification 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 Tyce Testina 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 long as the 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 Attachment No. 4 to IE Bulletin 79-01B-1 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 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 Secuence -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 Apn-endix 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 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 sare test specimen in the appropriate sequence.4. Test Soecimen Acing -Tests which were successful using test specimens which had not been preaged may be considered acceptable provided the co0cnent does not contain materials which are known to be susceptible Attachment No. 4 to IE Bulletin 79-01B v-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.S. 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.

Attachment No. 4 to IE Bulletin 79-O1B* : -13 5.3 Oualification 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 fown: acceptable on a case by case basis.1. Radiation Oualiflcation

-Some of the earlier tvop 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 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 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 SDray Qualification

-Components enclosed entirely in corrosion resistant cases (egg.1 stainless steel) may be shown to be qualified for a chemical environment by an analysis of the effects of the particular chemicals on the zarticular enclo-sure materials.

The effects of chemical sprays on the pressure inmtegrity of any gaskets or seals present should be considered in the analysis.

.Attachment No. 4 to IE Bulletin 79-O1B_14 6.0 Marcin IEEE Std. 323-1974 dC 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 conditions 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 tes: 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 satisfiedino separate margin factors are required to be added to the service conditions when specifying test conditions.

7.0 Acina Inpiicit 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 equipment is not sufficient to justify the expense for plants already constructed and operating.

This position does not, however, exclude equipment

.* Attachment No. 4 to IE Bulletin 79-O1B 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 Cornplete 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.

Attachment No. 4 to IE Bulletin 79-OlB APPENDIX A TYPICAL EQUIPMENT/FUNCTIONS

NEEDED FOR MITIGATION

OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steanrline Isolation Main Feedwater Shutdown and Isolation Emergency Power Emergency Core Cooling 1 Contairment Heat Renoval Containment Fission Product Removal Containment Conbustible 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 Radiation Monitoring3 Safety Related Display Instrumentation

3 Attachment No. 4 to IE Bulletin 79-O1B These systems will differ for PWRs and BWRs, and for older and newer plents. In each case the system features which allow fov transfer to recirculation cooling mode and establishment of long term cooling with boron prec-ipitation control are to be considered as part of the system to be evaluated.

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 volumse control system, and steam dump systems.3 More specific identification of these types of equipment can be found in the plant emergency procedures.

.* Attachment No. 4 to IE Bulletin 79-O1B-~~~~ El PEN~ v PROCEU?.ES

FOR EVALUATING

G6MfA RADIATION

SERVICE CONDITWNS Introduction and Discussion The adequacy of gamnma radiation servi-ce conditions specified for inside containment during a LOCA or FML3 accident can be verified by assuming a conservative dose at the contaTlment centerline and adjusting the dose according the plant specific parameters;

The purpose of this appendix ts to identify thase paraneters whose effect on the total gamma dose is easy to quantify with a high degree of ccnfidence 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: (l} A conservative dose at the containment centerline of 2 x 107 RADS for a LOCA and 2 x 10i RADS for a MSLE 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 WP.EG-053, "Interim Staff Position on Environrental Qualification of Safety-Related Electrical Equip-ment. " Therefore, all the llmitations listed in Appendix D of NURES-.588 apply to these procedures.

t2) The sample calculation In Appendix D of HLUREG-0588 is for a 4,000 MWth pressurized water reactor housed in a 2.52 x 1O6 ft 3 contain-ment wi.th an Iodine scrzbbing spray system. A similar calculation without Iodine scrubbint sprzys would increase the dose to equipment approxriately

150. The conservative dose o.' 2 x 107 RADS assumed S. .,'Attachment No. 4 to IE Bulletin 79-O1B-2- in the procedure below includes sufficient conservatism to account for this factor. Therefore, the proc.edure is also applicable to plants without an iodine scrubbing spray system.(3) Shielding calculations are based on an average gamma energy of 1 MEY 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 assumptions and adjustment factors developed on a case by case basis.Procedure Figures I 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.

• , ..-Attachment No. 4 to IE Bulletin 79-O1B'~. -.* , -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 ft 3 Compartment Volume -8,000 ft 3 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 10i ft 3 containment volume and read a 30-day integrated dose of 1.5 x 107 RADS.SteD 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).Stem 3 Enter Figure 3 at 8,000 ft 3 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

Attachment No. 4 to IE Bulletin 79-OlB Page-23 of 33-4-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 correction 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 (Z.O xl10 6 1 = 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.

J

.; &n FIGURE 1 K1tGAM FOR rONTAINMENT

VOLUME AND REACTOR P'- m JA DOSE CORRECTIONS*

CONTAINMEN

VOLUME (ft 3xlC 2x1C I Xi1o 5 x 10 4x10V.3x16 T 3)~6 5-MWTH 40 4o00_3000k_40DW _30 DAY INTEGRATED

YDOSE 4 x 10o Attachment No. 4 to IE Bulletin 79-OlB K 3 x 107_-1000 I I 2 x107 500%E 200 2x 10 w 1 x 107-I x 1O 5 x 1061 _4 x 106 _3x106 2.S x 106 2.0 x 106 I-1 x 106_I*ISLB ACCIDENT DOSES SHOULD BE READ AS A FACTOR OF 10 LESS

DOSE CORRECTION

FACTOR FOR CONCRETE SHIELDING Y( ONLY) Attachment NcoA to IE Bul 108 page 25 of 33 x1 S x oS 0 1 X 104 I- -l10:3 fit ', Ids , Nit~to I t* 1oC 1O7 106 10 I DOSE (RADS) WITHOUT SHIELDING (FROM FIGURE 1)letin 79-01B S

a 106 FIGURE 3 \DOSE CORRECTIN

FACTOR FOR COMPARTMENT

VOLUMBE Attachment No. 4 to IE Bulletin 79-O1B -*0 I-z Lu C 0 C;CD I 106 I I I I I I I ,I I I 0.2.4.6.8 1.0 CORRECTION

FACTOR

D URE 4 DOSE CORRECTION

FOR TIME hEQUIRED TO REMAIN FUNCTIONAL

c-V-.C.*1 a,-w U.-.o U r-.4Ju 0 4J I Ad O=: VI)al 0n.C.0 .1.0.1-.I I II hIIII.01 I a I II i fi I I A I I fia ll I I I I I i lt I I I I l I I , , I .. I ...........

....1 1.0 10 100 1000 TIME REQUIRED TO REMAIN FUNCTIONAL

MHRSP 4

• - .Attachment No. 4 to IE Bulletin 79-O0B t ' *Pale 28 of 33:APPENDI 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 45 0 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 terms used to characterize the dose effect is as follows:# 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.I 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 (NMIS), the National Aeronautics and Space Administration's Scientific and Technical Aerospace Report (STA.), NTIS Government Report Announcements and Index (GRA), and

• .Attachment No. 4 to IE Bulletin 79-O1B 2-various manufacturers data reports. The materials list is not to be considered all inclusive 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.

11/14/79 TABLE C-1 THERMAL AND RADIATION

AGING DEGRADATION

OF SELECTED MATERIALS r1 T I ALSO AS smirnICANT

AGING 10 YILS 140 YRS RtADIATION

SUSCEPTIDI

LITY TyPrs or rQtUiPI*N.tr (wrI~nll 10 WHIh iTEIAi, NAly lip tyoII))IA% y I C' "'h fl As t'HK MATlt:I At.IIAS I S I- -1~ fi -A ! I I I -I I _ _Integrated Circuits JIC)Integrated Circuits IIC)C-tiS Transltors Diodes Silicon-Controlled Rectifiers Integrated Circuits (IC)Analog Vulcanized Fiber Fish Paper Polyester (unfilled)

Nylon Polycarbondte Polywide Chlorosulfonated Poly-ethylene 8um-n Integrated Circuit. (IC)104 104 l 14 Threshold a Allowable Threshold K K I K Ix Polyamide Itypalon'tSR/ti-trile tubber)AP ik A 105 106 6 10.105 105 106 106 K K K K I x x K K K K x x K I I K I K x K K I K I x x K I I K K K x K I x x x xC DP 0 M SU 0 C* l 0 O C W =X TTL biallyl Phthalate Silicone Rubbet a.I __________

I I L *Indicates that there is data available which shows a potential for significant thermal aging of the materials when exposed to normal operating conditions for either 10 or 40 years as indicated.

11/14/79

11/14/79 I.9-v U r-MTreAL ALSO AS rOTENTIAL OR.tlCNIFICIWT

AGING i0 YM 40 YM8 RRMArloN SuscePTInILITY

I Is TYPES OF EQUIPAUrTr (WITHIN wiiiaC MATERIAL M"Y UK INwXI ./7 7 -4kv1 a I I- I I I Polysultone Reaistora

-Wire-ound Resistors

-Carbmr omposition Capacitors

-Ceramia Capacitors

-alas.Capacitora

-Rica ENA Thermosetting Lamnatee, Oar X c HEA Thermos.ttin'

Laminates, Grafe XXXP"EOA theuosetting Laminate..

Grafe XPX Nm Thermosetting Laminates, Grade XPC WMR Thermoeetting Laminates, Grata XX HEt Thermoaetting LaOinate..

Grade XXP mHE ¶termosattinq Laminate., CGra XXX MhE Therrmoetting Laminate, Graft Ce eOM Thermoaetting La"nate. GCrade C wrasde 1107 10l 19 109 109 109 109 109 109 109 l0g 109 109 109 109 109 fllowable 24% Loss of Elonga-tion rhrerhold a Allowable U X X I K K I K I X X X I I I K I I X I I I I I I N U S I I I N K K I 3.X K I K-0 :r-tC+IDOt 0 CD 0@_hC f "I s-I 40 CI 1-4 ID s-O, U a X I aa U N.K I.1 L 1. .1 I ;i11/14/79

• vI;_iTypes or rvQuirfl ("ITIN WIlc0 IIRTERIAI 4.MAT IIe 1 09flU))109 Shre0l ron 5IE."IFICAPM

SS~T1ILT AS 10 vp' 40 Tits GM BSI 1L09 Threehold t9 103 1 9 109 109 N 9 10 109 *1010 1 It 1 W ENCLOSURE

2 IE Bulletin No. 79-O0B Date: January 14, 1980 RECENTLY ISSUED IE BULLETINS Bulletin No.79-13 (Rev. 2)Subject Cracking in Feedwater System Piping Date Issued 10/17/79 Issued To All PWRs with an OL and Designated Ap-plicants (for Action), All Other Power Reactor Facilities with an Operating License (OL) or Con-struction Permit (CP)(for Information)

79-17 (Rev. 1)79-25 79-02 (Rev. 2)79-26 79-27 79-28 Pipe Cracks in Stagnant Borated Water Systems 10/29/79 All PWRs with an OL (for Action). All other Power Reactor Facilities with an OL or CP (for In-formation)

All Power Reactor Facilities with an OL or CP (for Action)Failures of Westinghouse

11/2/79 BFD Relays in Safety-Related Systems Pipe Base Plate Designs Using Concrete Expansion Bolts Boron Loss From BWR Control Blades Loss of Non-Class-1-E

Instrumentation and Con-trol Power System Bus During Operation Possible Malfunction of NAMCO Model EA180 Limit Switches at Elevated Temperatures

11/8/79 All Power Reactor Facilities with an OL or CP 11/20/79 All BWR Power Reactor Facilities with an OL 11/30/79 All Power Reactor Facilities with an OL and those nearing Licensing (for Action)All Power Reactor Facilities with a CP (for Information).

12/7/79 All Power Reactor Facilities with an OL or CP