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 STATESNUCLEA*'REGULATORY COMMISSIONREGION I631 PARK AVENUEKING OF PRUSSIA, PENNSYLVANIA 19406OL -go-_gDocket Nos. 50-0350-247JAN 1 4 1980Consolidated Edison Company ofNew York, Inc.ATTN: Mr. W. J. Cahill, Jr.Vice President4 Irving PlaceNew York, New York 10003Gentlemen:Enclosed is IE Bulletin 79-OIB which requires action by you with regard toyour power reactor facility with an operating license.Should you have questions regarding this Bulletin or the actions required ofyou, please contact this office.Sincerely,Boyce H. GrierDirectorEnclosures:1 IE Bulletin No.79-01B with Attachments2. List of Recently Issued IE Bulletins

CONTACT

S. 0. Ebneter(215-337-5296)cc w/encls:L. 0. Brooks, Project Manager, IP NuclearW. Monti, Manager -Nuclear Power Generation DepartmentM. Shatkouski, Plant ManagerJ. M. Makepeace, Director, Technical EngineeringW. D. Hamlin, Assistant to Resident Manager (PASNY)J. 0. Block, Esquire, Executive Vice President -AdministrationJoyce P. Davis, Esquire80012 90Aw -

ENCLOSURE 1UNITED STATES SSINS No.: 6820NUCLEAR REGULATORY COMMISSION Accessions No.:OFFICE OF INSPECTION AND ENFORCEMENT 7910250528WASHINGTON, D.C. 20555IE Bulletin No. 79-O1BDate: 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 ofthe environmental qualification of Class IE electrical equipment to ensurethat the equipment will function under (i.e. during and following) postulatedaccident conditions.The NRC staff has completed the initial review of licensees' responses toBulletin No. 79-01. Based on this review, additional information is needed tofacilitate completion of the NRC evaluation of the adequacy of environmentalqualification of Class IE electrical equipment in the operating facilities.In addition to requesting more detailed information, the scope of this Bulletinis expanded to resolve safety concerns relating to design basis environmentsand 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 CLASSIE ELECTRICAL EQUIPMENT IN OPERATING REACTORS", provides the guidelines andcriteria the staff will use in evaluating the adequacy of the licensee's ClassIE equipment evaluation in response to this Bulletin.In general, the reporting problems encountered in the original responses andthe additional information needed can be grouped into the following areas:1. All Class IE electrical equipment required to function under the postulatedaccident conditions, both inside and outside primary containment, was notincluded in the responses.2. In many cases, the specific information requested by the Bulletin foreach component of Class IE equipment was not reported.3. Different methods and/or formats were used in providing the writtenevidence of Class IE electrical equipment qualifications. Some licenseesused the System Analysis Method which proved to be the most effectiveapproach. This method includes the following information:a. Identification of the protective plant systems required to functionunder postulated accident conditions. The postulated accidentconditions are defined as those environmental conditions resultingfrom both LOCA and/or HELB inside primary containment and HELBoutside the primary containment.

Enclosure 1 IE Bulletin No. 79-QIBDate: January 14, 1980 b. Identification of the Class IE electrical equipment items withineach of the systems identified in Item a, that are required tofunction under the postulated accident conditions.c. The correlation between the environmental data requirements specifiedin the FSAR and the environmental qualification test data for eachClass IE electrical equipment item identified in Item b above.4. Additional data not previously addressed in IE Bulletin No. 79-01 areneeded to determine the adequacy of the environmental qualification ofClass IE electrical equipment. These data address component aging andoperability in a submerged condition.Action To Be Taken By Licensees Of All Power Reactor Facilities With An OperatingLicense (Except those 11 SEP Plants Listed on Attachment 1)1. Provide a "master list" of all Engineered Safety Feature Systems (PlantProtection Systems) required to function under postulated accident conditions.Accident conditions are defined as the LOCA/HELB inside containment, andHELB outside containment. For each system within (including cables,EPA's terminal blocks, etc.) the master list identify each Class IEelectrical equipment item that is required to function under accidentconditions. Pages 1 and 2 of Attachment 2 are standard formats to be usedfor the "master list" with typical information included.Electrical equipment items, which are components of systems listed inAppendix A of Attachment 4, which are assumed to operate in the FSARsafety analysis and are relied on to mitigate design basis events areconsidered within the scope of this Bulletin, regardless whether or notthey were classified as part of the engineered safety features when theplant was originally licensed to operate. The necessity for further upgrading of nonsafety-related plant systems will be dependent on theoutcome of the licensees and the NRC reviews subsequent to TMI/2.2. For each class IE electrical equipment item identified in Item 1, providewritten evidence of its environmental qualification to support the capa-bility of the item to function under postulated accident conditions. Forthose class IE electrical equipment items not having adequate qualifica-tion data available, identify your plans for determining qualificationsof these items and your schedule for completing this action. Providethis 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 datashould be provided for design basis accident conditions and qualificationtests performed. This data may be provided in profile or tabular form.

Enclosure 1 IE Bulletin No. 79-01BDate: January 14, 1980 . Evaluate the qualification of your Class IE electrical equipment againstthe guidelines provided in Attachment 4. Attachment 5, "Interim StaffPosition on Environmental Qualification of Safety-Related ElectricalEquipment," provides supplemental information to be used with theseguidelines. 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 determineequipment qualification and the schedule for completing the actions.5. Identify the maximum expected flood level inside the primary containmentresulting from postulated accidents. Specify this flood level by elevationsuch as the 620 foot elevation. Provide this information in the formatof Attachment 3.6. Submit a "Licensee Event Report" (LER) for any Class IE electrical equipmentitem which has been determined as not being capable of meeting environmentalqualification requirements for service intended. Send the LER to theappropriate 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. Ifplant operation is to continue following identification, provide justifi-cation for such operation in the LER. Provide a detailed written reportwithin 14 days of identification to the appropriate NRC Regional Office.Those items which were previously reported to the NRC as not being qualifiedper IEB-79-01 do not require an LER.7. Complete the actions specified by this bulletin in accordance with thefollowing schedule:(a) Submit a written report required by Items 1, 2, and 3 within 45 daysfrom receipt of this Bulletin.(b) Submit a written report required by Items 4 and 5 within 90 days fromreceipt 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.aand 7.b above, written statements of the above information, signed under oathor 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, Officeof Inspection and Enforcement, Division of Reactor Operations Inspection,Washington, D.C. 20555.

Enclosure 1Approved bygiven underIE Bulletin No. 79-01BDate: January 14, 1980 GAO, B180225 (R0072); clearance expires 7/31/80. Approval wasa blanket clearance specifically for identified generic problems.Attachments:1. List of SEP Plants2. Master List Standard Format, Typical3. System Component Evaluation Work Sheet4. Guidelines for Evaluating Environmental Qualification of ClassIE Electrical Equipment in Operating Reactors5. Interim Staff Position on Environmental Qualification of Safety-RelatedEquipment (To

Addressees

Only)

Attachment 1 to IE Bulletin 79-O1BSEP PlantsPlant RegionDresden 1 IIIYankee Rowe IBig Rock Point IIISan Onofre 1 VHaddam Neck ILaCrosse IIIOyster Creek IR. E. Ginna IDresden 2 IIIMillstone 1 IPalisades 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.EIE. Bull1et in. 79-OIBCOMPONENTSLocationPlant-Identification Inside Primary Outside PrimaryNumber Generic Name Containment ContainmentIPT 456 -PRESSURE TRANSMITTER xILT 594 LEVEL TRANSMITTER x.S 210 LIMIT SWITCH xII. SYSTEM: AUTOMATIC DEPRESSURIZATION SYSTEM (ADS)COMPONENTS..~Locatilon-.Plant Identifcation Inside Primary Outside Primary.Nuber Generic Name Containment ContainmentB21-ROOI VALVE MOTOR OPERATOR xB21-F003 -SOLENOID VALVE xB21-FOlO PRESSURE SWITCH .x II. SYSTEM. RHR EQUIPMENT/COMOI1NENTS(Typical) Attachment No.**COMPONENTS'.-2 to IE Bulletin 79-01Bl .k.__________________________________________________________________________ IPlant IdentificationNumber*416xP455O-RING GASKETx*EPA,- Clas~ E,Westinghouse: E OOC ELECTRICAL PENETRATION ASSEMBLY XKULKA No. ET35 TERMINAL BOARD xONKONITE, lOOOV, 3CBlack POWER CABLE x xX BRAND 10W-40 LUBRICATE OIL x15 KB69 (BostonWire & Cable) INSTRUMENTATION CABLE x xCutler Hamner TB TERMINAL BOX xN o .-6_ _ _ _ _ _ _ _ _ _ _RAYCHEM XYZ CABLE SPLICE x xScotch No. 54 INSULATING TAPE xT&B No. 10 INSULATE TERMINAL LUG xY Brand Epoxy No;. SEALANT x x.ll ._________________________* When a component ismanufacturer, model** Like components maynot identifiednumber, serialbe referenced.by plant identification number, use thenumber, etc.

' Facility:Unit:D ocket:SYSTEM COMPONENT EVALUATION WORK SHEET(Typical)Attachment No. 3 to IE Bulletin No. 79-OIBPage I of 3 t'EfIVI RONMENT DOCU1MENTATION'REF* QALFCTOOTTNDEQUIPMENT DESCRIPTION QUALIFICATION OUTSTANDIpec if- ua li- Specifi- ualiti- METHOD ITEMSPa -arameter iDra tnn -catin nn ._System: RHR Operating 15 min. 300 min. 5 Simultaneou! NonePlant ID No. IPT456 Time TestComponent Temperature SEE ACCIDENT AND 5 Simultaneou!PRESSURE TRANSMITTER. S EST PROFILESTAN( ) TEST PROFILES .Test NoneManufacture: PROVIDED :Fischer-Porter Co. Pressurao (PSIA) , 1 5 Simultaneou NoneModel Number: Test50-EN-1071-BCXN-NS RelativeFunctlon: Humidity(%) 100% 100% 1 5 Simultaneou NoneAccident Monitoringi. ii __- ' _ Test ,Chemical N3B03/Accuracy: Spec: 5% Spray NAOH 1 See Note 1Demon: 4% NOServi ce: RHR Pump lA 6Radiaton 4xl06rads l.2xlO8rad 2 6 SequentialDischarge Pressure Test NoneS/NiO7 11. Seq4entf NnLocation: Containment Aging yrs 40 yrs 3 7, 8 Test ysNoneFlood Level Elev: 620' Not Not NoneAbove Flood Level: Y Yes lSubmergence Required Required See Note 2N o x 'j_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _I'IG;C(-uocumentation References:,Nbtes:1.2.3.4.5.6.7..8.'tSAR Chapter 3, Paragraph 3.11FSAR Chapter 14, Paragraph 14.2.3.1Technical Specification 3.4.1, Paragraph ATechnical. Speciffcation 4.6.5, Paragraph BFIRL Test Report No. ?O00 dated November 2, 1972Fischer and Porter Co. Test Report No. 2500-1A. 0. DOD Engineering Evaluation Data.Report No. 6932Wylie Laboratbry Report.Ro. 4671. XYZ Letter No. 237-1, dated November 2, 1979,has been sent to MFG. requesting the qualificationinformation. If qualification not determinedacceptable by December 15, 19791, componentwill be replaced during refueling outage March 1980..,.I .2. In the FSAR submergence was not consideredan environmental parameter. ABC Laboratoryis to perform submergence test in April 1980..I

Attachment 3 to IE Bulletin 79-OIB SYSTEM COMPONENT EVALUATION WORK SHEETINSTRUCTIONS1. Equipment Description: Provide the specific information requested foreach Class IE electrical component. Provide component location, specificinformation such as the building, access floor elevations, and whetherthe component is above the flood level elevation. In addition, providethe specified and demonstrated accuracies of all instruments for theirtrip functions and/or post accident monitoring requirements. Cables,EPA's, terminal blocks, and other items shall be identified as part ofthe engineered safety features systems.2. Environment: List values for each environmental parameter indicated.List the specification values" obtained from postulated accident analysisin the "SPEC" column. List the "qualification values" obtained from testreports, engineering analysis data, etc. in the "Qual" column. Tempera-ture, pressure, etc., as a function of time shall be provided in profileor tabular form. Specify the time period that the component or equipmentis required to function and identify the document which provides thebasis for this time interval.It is expected that some listed parameters were not requested of thelicensee at the time of their license issuance: Address each parametercondition during this review. If it is determined that a parameter suchas submergence or a service condition such as aging was not previouslyconsidered, identify it as an "Outstanding Item."3. Documentation Reference: Reference the documents from which informationwas obtained in the "Spec" column. Identify the document, paragraph,etc., that contains the postulated accident environmental specificationdata. In the "Qual" column identify the document, paragraph, etc., thatcontains the environmental qualification data.4. Qualification Method: Identify the method of qualification. To describethe qualification method use words such as simultaneous test, comparisontest, sequential test, and/or engineering/mathematical analysis. Wordssuch as "test" and/or "analysis" when used alone do not adequately identifythe qualification method.5. Outstanding Items: Identify parameters for which no qualification datais 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 actionsplanned for determining qualification and the schedule for completingthese actions.

Attachment 3 of IE Bulletin 79-010 EQUIPMENTDESCRIPTIONNOTE 1POSTULATEDACCIDENTENVIRONMENTNOTE 2TYPICAL-2-SERVICE CONDITION PROFQUALIFICATIONTESTENVIRONMENTNOTE 3ACCURACY ACCURACYREQUIREMENTS DEMONSTRATEDNOTE 4 NOTE 5EXCEPTIONSORREMARKSNOTE 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 acharacteristic profile.3. Provide sufficient values of temperature and pressure as a function of time for which equipment was qualifiedto 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 tripfunctions and/or post accident monitoring as applicable.6. Identify any exception or deviation between specified service condition and qualification service condition andjustification to explain acceptance of deviation.

.Attachment No. 4 to6 3ulTetin 1--01B- GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATIONOF CLASS IE ELECTRICAL EQUIPMENTIN OPERATING REACTORS1.0 Introduction2.0 Discussion3.0 Identification of Class IE Equipment4.0 Service Conditions4.1 Service Conditions Inside Containment for a Loss ofCoolant Accident (LOCA)1. Temperature and Pressure Steam Conditions2. Radiation3. Submergence4. Chemical SDrays4.2 Service Conditions for a PWR Main Steam Line Break (MSLB)Inside Containment1. Temperature and Pressure Steam Conditions2. Radiation3. Submergence4. Chemical Sprays4.3 Service Conditions Outside Containment4.3.1 Areas Subject to a Severe Environment as a Resultof aHighEnergy Line Break (HELB)4.3.2 Areas Where Fluids are Recirculated From InsideC ainment to Accom'lish Lona. "er eCore Coolina Following a LOCA1. Temoerature, Pressure and Relative Humidity2. Radiation3. Submercence4. Chemical SDrays

.tAttachment No. 4 to IE Bulletin 79-01B'. -2-4.3.3 Areas Normally Mat--.talned at Room Conditions5.0 Qualification Methods5.1 Selection of Qualification Method5.2 Qualification by Type Testing-l. Simulated Service Conditions and Test Duration2. Test Specimen3. Test Sequence4. Test Specimen Aging5. Functional Testing and Failure Criteria6. Installation Interfaces5.3 Qualification by a Combination of Methods (Test, Evaluation,Analysis)* 6.0 Margin7.0 Acina8.0 DocumentationAppendix A -Typical Equipment/Functions Needed for Mitigation ofa LOCA or MSLB AccidentAppendix B -Guidelines for Evaluating Radiation Service ConditionsInside Containment for a LOCA and MSLB AccidentAppendix C -Thermal and Radiation Aging Degradation of SelectedMaterials Attachment No. 4 to IE Bulletin 79-01B GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATIONOF CLASS IE ELECTRICAL EQUIPMENTIN OPERATING REACTORS1.0 INTRODUCTIONOn February 8, 1979, the NRC Office of Inspection and Enforcement issuedIE Bulletin 79-01, entitled, "Environmental Qualification of Class IEEquipment." This bulletin requested that licensees for operating powerreactors complete within 120 days their reviews of equipment qualificationbegun earlier in connection with IE Circular 78-08. The objective ofIE Circular 78-08 was to initiate a review by the licensees to determinewhether proper documentation existed to verify that all Class IE electricalequipment would function as required in the hostile environment which couldresult from design basis events.The licensees' reviews are now essentially complete and the NRC staff hasbegun to evaluate the results. This document sets forth guidelines for theNRC staff to use in its evaluations of the licensees' responses to IEBulletin 79-01 and selected associated qualification documentation. Theobjective of the evaluations using these guidelines is to identify Class IEequipment whose documentation does not provide reasonable assurance of environ-mental qualification. All such equipment identified will then be subjectedto a plant application-specific evaluation to determine whether it should berequalified or replaced with a component whose qualification has been adequatelyverified.These guidelines are intended to be used by the NRC staff to evaluate thequalification methods used for existing equipment in a particular class ofplants, i.e., currently operating reactors including SEP plants.

Attachment No. 4 to IE Bulletin 79-01B2 Equipment in other classes of plants not yet licensed to operate, orreplacement equipment for operating reactors, may be subject to differentrequirements such as those set forth in NUREG-0588, Interim Staff Positionon Environmental Qualification of Safety-Related Electrical Equipment.In addition to its reviews in connection with IE Bulletin 79-01 the staffis engaged in other generic-reviews that include aspects of the equipmentqualification issue. TMI-2 lessons learned and the effects of failures ofnon-Class IE control and indication equipment are examples of these genericreviews. In some cases these guidelines may be applicable, however, thisdetermination will be made as part of that related generic review.2.0 DISCUSSIONIEEE Std. 323-19741 is the current industry standard for environmentalqualification of safety-related electrical equipment. This standard wasfirst issued as a trail use standard, IEEE Std. 323-1971, in 1971 and laterafter substantial revision, the current version was issued in 1974. Bothversions 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 inthe 1971 trial use standard.The intent of this document is not to provide guidelines for implementingeither version of IEEE Std. 323 for operating reactors. In fact most ofthe operating reactors are not committed to comply with any particularindustry standard for electrical equipment qualification. However, all ofthe operating reactors are required to comply with the General Design Criteria1IEEE Std. 323-1974, 'IEEE Standard for Qualifying Class IE Equipment forNuclear Power Generating Stations."

.'*. .. -Attachment No. 4 to IE Bulletin 79tO1B* specified in Appendix A of 10 CFR 50. General Design Criterion 4 statesin part that structures, systems and components important to safetS shallbe designed to accomodate the affects of and to be compatible with theenvironmental 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 requiredto confirm that operating reactors are in compliance with General DesignCriterion 4.3.0 IDENTIFICATION OF CLASS IE EQUIPMENTClass IE equipment includes all electrical equipment needed to achieveemergency reactor shutdown, containment isolation, reactor core cooling,containment and reactor heat removal, and prevention of significant releaseof radioactive material to the environment, Typical systems included inpressurized and boiling water reactor designs to perform these functionsfor the most severe postulated loss of coolant accident (LOCA) and mainsteanline break accident (MSLB) are listed in Appendix A.More detailed descriptions of the Class IE equipment installed at specificplants can be obtained from FSARs, Technical specifications, and emergencyprocedures. Although variation in nomenclature may exist at the various plants,environmental qualification of those systems which perform the functionsidentified in Appendix A should be evaluated against the appropriate serviceconditions CSection 4.0).The guidelines in this document are applicable to all components necessaryfor operation of the systems listed in Appendix A including but not limitedto valves, motors, cables, connectors, relays, switches, transmitters andvalve position indicators, Attachment No. 4 to IE Bulletin 79-O1B -4 -4.0 SERVICE CONDITIONSIn order to determine the adequacy of the qualification of equipment It Isnecessary to specify the environment the equipment is exposed to duringnormal 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 licenseesubmittals are acceptable, unless otherwise noted in the guidelines discussuedbelow.4,1 Service Conditions Inside Containment for a Loss of Coolant Accident (LOCA)1, Temperature and Pressure Steam Conditions q In general, the containmenttemperature and pressure conditions as a function of time should bebased on the analyses in the FSAR, In the specific case of pressuresuppression type containments, the following minimum high tempeatureconditions should be used: (l11BWR Drywells .3400F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; andC21 FWR 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.. ?adiation -When specifying radiation service conditions for equipmentexposed to radiation during normal operating and accident conditions,the normal operating dose should be added to the dose received duringthe course of an accident. Guidelines for evaluating beta and gammaradiation service conditions for general areas inside containment areprovided below, Radiation service conditions for equipment locateddirectly above the containment sump; in the vicinity of filters, or-submerced in contaminated liquids must be evaluated on a case by casebasis, Guidelines for these evaluations are not provided in thisdocument.,

, Attachment No. 4 to IE Bulletin 79-O1B Ganma Radiation Doses -A total gamma dose radiation service conditionof 2 x 107 RADS is acceptable for Class IE equipm..at located in generalareas inside containment for PWRs with dry type containments, Where adose less than this value has been specified, an application specificevaluation must be performed to determine If the dose specified isacceptable. Procedures for evaluating radiation service conditionsin such cases are provided In Appendix B, The procedures in AppendixB are based on the calculation for a typical PWR reported in AppendixD of XUREG-.05881Ga6nna dose radiation service conditions for BWRs and PWRs with icecondenser containments must be evaluated on a case by case basis.Since the procedures in Appendix B are based on a calculation for atypical PWR with a dry type containment, they are not directly applicableto BWRs and other containment types, However, doses for these otherplant configurations may be evaluated using similar procedures withconservative dose assumptions and adjustment factors developed on acase by case basIs,Bet.a Radiation Doses -Beta radiation doses generally are less significantthan gama radiation doses for equipment qualification, This is due tothe low penetrating power of beta particles in comparison to gamma raysof equivalent energy, Of the general classes of electrical equipmentin a plant (etg,, cables, instrument transmitters, valve operators,containment penetrations), electrical cable is considered the most1NUkE-0588, Interim Staff Position on Environmental Qualification ofSafetyRelated Electrical Equipment.

Attachment No. 4 to IE Bulletin 79-OIB -6-vulnerable to damage from beta radiation. Assuming a TID 14844source term, the average maximum beta energy and isotopic abundancewill vary as a function of time following an accident. If theseparameters are considered in a detailed calculation, the conservativebeta surface dose of 1.40 x x 108 RADS reported in Appendix 0 of NUREG0588 would be reduced by approximately a factor of ten within 30 milsof the sur face of electrical cable insulation of unit density. Anadditional 40 mils of insulation (total of 70 mils) results in anotheractor of 10 reduction in dose. Any structures or other equipment inthe vicinity of the equipment of interest would act as shielding tofurther reduce beta doses. If it can be shown, by assuming a conserva-tive unshielded surface beta dose of 2.0 x 108 RADS and consideringthe shielding factors discussed here, that the beta dose to radiationsensitive equipment internals would be less than or equal to 106 ofthe tota' garma dose to which an item of equipment has been qualified,then that equipment may be considered qualified for the total radiationenvironment (gamma plus beta). If this criterion is not satisfiedthe radiation service condition should be determined by the sum ofthe garma and beta doses.3. Submercence -The preferred method of protection against the effectsof submEergency is to locate equipment above the water flooding level.Specifying saturated steam as a service condition during type testingof equipment that will become flooded in service is not an acceptablealternative 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 bequalified for the most severe chemical environment (actdic orbasic) which could exist, Demineralized water sprays should notbe exempt from consideration as a potentially adverse servicecondition.,4.2 Service Conditions for a PWR Main Steam Line Break (MSLB) Inside ContainmentEquipment required to function in a steam line break environment mustbe qualified for the high temperature and pressure that could result.In some cases the environmental stress on exposed equipment may behigher than that resulting from a LOCA, in others it may be no moresevere than for a LOCA due to the automatic operation of a containmentspray system.1. Ter.Derature and Pressure Steam Conditions -Equipment qualified fora LOCA environment is considered qualified for a MSLB accident environ-rer.t in plants with automatic spray systems not subject to disablingsingle component failures. This position is based on the 'BestEstim.at'e calculation of a typical plant peak temperature and pressureand 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 SteamlineBreak Inside Containment.Class IE equipment installed in plants without automatic spraysystems or plants with Spray systems subject to disabling singlefailures or delayed initiation should be qualified for a MSLB accidentenvironment determined by a plant specific analysis. Acceptable methodsSee NUR E 0456, Short Term Safety Assessment on the EnvironmpntalQualification of Safety-Related Electrical Equipment of SEP OperatingReactors, 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 providedin Section 1.2 for Category II plants in NUREG-0588, Interim StaffPosition on Environmental Qualification of Safety-Related EletctricalEquipment.2. Radiation -Same as Section 4.1 above except that a conservativegamia 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 Containment4.3.1 Areas Subject to a Severe Environment as a Result of a High EnergyLine Break 'HELB)Service conditions for areas outside containment exposed to a HELB wereevaluated on a plant by plant basis as part of a program initiated bythe staff in Dece.mber, 1972 to evaluate the effects of a HELB. Theequipment required to mitigate the event was also Identified. Thisequipment should be qualified for the service conditions reviewed andapproved n tne i.-. Sa-ezy Evaluation Report. for each specific plant.4.3.2 Areas Where Fluids are Recirculated from Inside Containment to AccomplishLona-Temn Core Coolino Followina a LOCA1. Termerature and Relative Humidity -One hundred oercent relative humidityshouTd be established as a service condition in confined spaces. Thetemoerature and pressure as a function of time should be based on theplant unique analysis reported in the FSAR.

Attachment No. 4 to IE Bulletin 79-O1B 2. Radiation -Due to differences in equipment arrangement withinthese areas and the significant effect of this factor on doses,radiation service conditions must be evaluated on a case by casebasis. In general, a dose of at least 4 x 106 RADS would beexpected.3. Submergence -Not applicable.4. Chemical Sorays -Not applicable.4.3.3 Areas Normally Maintained at Room ConditionsClass IE equipment located in these areas does not experience significantstress due to a change in service conditions during a design basis event.This equipment was designed and installed using standard engineeringpractices and industry codes and standards (e.g., ANSI, NEMA, National:Electric Code). Based on these factors, failures of equipment in theseareas during a design basis event are expected to be random except tothe extent that they may be due to aging or failures of air conditioning orventilation systems. Therefore, no special consideration need be given tothe environmental qualification of Class IE equipment in these areas providedthe aging requirements discussed in Section 7.0 below are satisfied and theareas are maintained at room conditions by redundant air conditioning orventilation systemis served by the onsite emergency electrical power system.Equip.ent located irf areas not served by redundant systems powered fromonsite emergency sources should be qualified for the environmental extremeswhich could result from a failure of the systems as determined from a plantspecific analysis.5.0 QJALIFICATION METHODS

Attachment No. 4 to IE Bulletin 79-OB lo: V-10 -5.1 Selection of Qualification MethodThe choice of qualification method employed for a particular applicationof equipment is largely a matter of technical Judgement based on suchfactors as: (1) the severity of the service conditions; (2) the structuraland material complexity of the equipment; and (3) the degree of certaintyrequired in the qualification procedure (i.e., the safety importanceof the equipment function). Based on these considerations, type testingis the preferred method of qualification for electrical equipment locatedinside containment required to mitigate the consequences of design basisevents, i.e., Class IE equipment (see Section 3.0 above). As a minimum,the cualification for severe temperature, pressure, and steam serviceconditions for Class IE equipment should be based on type testing.:Qualification for other service conditions such as radiation and chemicalsprays may be by analysis (evaluation) supported by test data (see Section5.3 below). Exceptions to these general guidelines must be justified on acase by case basis.5.2 Oualification by Tyce TestinaThe evaluation of test plans and results should include consideration ofthe following factors:1. Simulated Service Conditions and Test Duration -The environment in thetest chamber should be established and maintained so that it envelopesthe service conditions defined in accordance with Section 4.0 above.The time duration of the test should be at least as long as the periodfrom the initiation of the accident until the temperature and pressureservice conditions return to essentially the same levels that existedbefore 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 materialsinvolved t 11 not experience significant accelerated thermal agingduring the period not tested.2. Test Soecimen -The test specimen should be the same model as theequipment being qualified. The type test should only be considered validfor equipment identical in design and material construction to the testspecimen. 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 asteam./air environment at elevated temperature, and pressure in thesequence defined for its service conditions. Where radiation is aservice condition which is to be considered as part of a type test, itmay-be applied at any time during the test sequence provided the componentdoes not contain any materials which are known to be susceptible tosignificant radiation damage at the service condition levels ormaterials whose susceptibility to radiation damage is not known (seeApn-endix C). If the component contains any such materials, the radiationdose should be applied prior to or concurrent with exposure to the elevatedtemperature and pressure steam/air environment. The same test specimenshould be used throughout the test sequence for all service conditionsthe equipment is to be qualified for by type testing. The type testshould only be considered valid for the service conditions applied tothe sare test specimen in the appropriate sequence.4. Test Soecimen Acing -Tests which were successful using test specimenswhich had not been preaged may be considered acceptable provided theco0cnent 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 Section7.0). If the component contains such materials a qualified life for thecomponent must be established on a case by case basis. Arrhenius techniquesare generally considered acceptable for thermal aging.S. Functional Testing and Failure Criteria -Operational modes testedshould be representative of the actual application requirements(e.g., components which operate normally energized in the plantshould be normally energized during the tests, motor and electricalcable loading during the test should be representative of actualoperating conditions). Failure criteria should include instrumentaccuracy requirements based on the maximum error assumed in theplant safety analyses. If a component fails at any time duringthe test, even in a so called "fail safe" mode, the test shouldbe considered inconclusive with regard to demonstrating the abilityof the component to function for the entire period prior to thefailure.6. Installation Interfaces -The equipment mounting and electrical ormechanical seals used during the type test should be representativeof the actual installation for the test to be considered conclusive.The equipment qualification program should include an as-builtinspection in the field to verify that equipment was installedas it was tested. Particular emphasis should be placed on commonproblems such as protective enclosures installed upside down withdrain holes at the top and penetrations in equipment housings forelectrical connections being left unsealed or susceptible tomoisture incursion through stranded conductors.

Attachment No. 4 to IE Bulletin 79-O1B* : -13 5.3 Oualification by a: Combination of Methods (Test, Evaluation,AnalysisAs discussed in Section 5.1 above, an item of Class IE equipment maybe shown to be qualified for a complete spectrum of service conditionseven though it was only type tested for high temperature, pressureand steam. The qualification for service conditions such as radiationand chemical sprays may be demonstrated by analysis (evaluation). Insuch cases the overall qualification is said to be by a combination ofmethods. Following are two specific examples of procedures that areconsidered acceptable. Other similar procedures may also be reviewedand fown: acceptable on a case by case basis.1. Radiation Oualiflcation -Some of the earlier tvop tests performedfor operating reactors did not include radiation as a servicecondition. In these cases the equipment may be shown to beradiation qualified by performing a calculation of the doseexpected, taking into account the time the equipment is requiredto remain functional and its location using the methods describedin Appendix B, and analyzing the effect of the calculated doseon the materials used in the equipment (see Appendix C). As ageneral rule, the time required to remain functional assumed for dosecalculations 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 incorrosion resistant cases (egg.1 stainless steel) may be shownto be qualified for a chemical environment by an analysis ofthe effects of the particular chemicals on the zarticular enclo-sure materials. The effects of chemical sprays on the pressureinmtegrity of any gaskets or seals present should be consideredin the analysis.

.Attachment No. 4 to IE Bulletin 79-O1B_14 6.0 MarcinIEEE Std. 323-1974 dC ines margin as the difference between the mostsevere specified service conditions of the plant and the conditions usedin type testing to account for normal variations in commercial productionof equipment and reasonable errors in defining satisfactory performance.Section 6.3.1.5 of the standard provides suggested-factors to be appliedto the service conditions to assure adequate margins. The factor appliedto the time equipment is required to remain functional is the mostsignificant in terms of the additional confidence in qualification thatis achieved by adding margins to service conditions when establishingtes: environments. For this reason, special consideration was given tothe time required to remain functional when the guidelines for FunctionalTesting and Failure Criteria in Section 5.2 above were established. Inaddition, all of the guidelines in Section 4.0 for establishing serviceconditions include conservatisms which assure margins between the serviceconditions specified and the actual conditions which could realisticallybe expected in a design basis event. Therefore, if the guidelines inSection 4.0 and 5.2 are satisfiedino separate margin factors are requiredto be added to the service conditions when specifying test conditions.7.0 AcinaInpiicit in the-staff position in Regulatory Guide 1.89 with regard tobackfitting IEEE Std. 323-1974 is the staff's conclusion that theincremental improvement in safety from arbitrarily requiring that aspecific qualified life be demonstrated for all Class IE equipment isnot sufficient to justify the expense for plants already constructedand 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 tosignificant degradation due to thermal and radiation aging. Componentmaintenance or replacement schedules should include considerations ofthe specific aging characteristics of the component materials. Ongoingprograms should exist at the plant to review surveillance and maintenancerecords to assure that equipment which is exhibiting age related degrada-tion will be identified and replaced as necessary. Appendix C contains alisting of materials which may be found in nuclear power plants along withan indication of the material susceptability to thermal and radiation aging.8.0 DocumentationCornplete and auditable records must be available for qualification byany of the methods described in Section 5.0 above to be considered valid.These records should describe the qualification method in sufficientdetail to verify that all of the guidelines have beensatisfied. A simple vendor certification of compliance with a designspecification should not be considered adequate.

Attachment No. 4 to IE Bulletin 79-OlB APPENDIX ATYPICAL EQUIPMENT/FUNCTIONS NEEDED FORMITIGATION OF A LOCA OR MSLB ACCIDENTEngineered Safeguards ActuationReactor ProtectionContainment IsolationSteanrline IsolationMain Feedwater Shutdown and IsolationEmergency PowerEmergency Core Cooling1Contairment Heat RenovalContainment Fission Product RemovalContainment Conbustible Gas ControlAuxiliary FeedwaterContainment VentilationContainment Radiation MonitoringControl Room Habitability Systems (e.g., HVAC, Radiation Filters)Ventilation for Areas Containing Safety EquipmentComponent CoolingService WaterEmergency Shutdown2Post Accident Sampling and MonitoringRadiation Monitoring3Safety Related Display Instrumentation3 Attachment No. 4 to IE Bulletin 79-O1B These systems will differ for PWRs and BWRs, and for older and newerplents. In each case the system features which allow fov transfer torecirculation cooling mode and establishment of long term coolingwith boron prec-ipitation control are to be considered as part ofthe system to be evaluated.Emergency shutdown systems include those systems used to bring theplant to a cold shutdown condition following accidents which do notresult in a breach of the reactor coolant pressure boundary togetherwith a rapid depressurization of the reactor coolant system. Examplesof such systems and equipment are the RHR system, PORVs, RCIC, pressurizersprays, chemical and volumse control system, and steam dump systems.3More specific identification of these types of equipment can be foundin the plant emergency procedures.

.* Attachment No. 4 to IE Bulletin 79-O1B-~~~~ El PEN~ vPROCEU?.ES FOR EVALUATING G6MfA RADIATION SERVICE CONDITWNSIntroduction and DiscussionThe adequacy of gamnma radiation servi-ce conditions specified for insidecontainment during a LOCA or FML3 accident can be verified by assuminga conservative dose at the contaTlment centerline and adjusting the doseaccording the plant specific parameters; The purpose of this appendixts to identify thase paraneters whose effect on the total gamma dose iseasy to quantify with a high degree of ccnfidence and describe procedureswhich may be used to take these effects into consideration.The bases for the procedures and restrictions for their use are asfollows:(l} A conservative dose at the containment centerline of 2 x 107 RADSfor 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 calculationdescribed In Appendix D of WP.EG-053, "Interim Staff Positionon Environrental Qualification of Safety-Related Electrical Equip-ment. " Therefore, all the llmitations listed in Appendix D ofNURES-.588 apply to these procedures.t2) The sample calculation In Appendix D of HLUREG-0588 is for a 4,000MWth pressurized water reactor housed in a 2.52 x 1O6 ft3 contain-ment wi.th an Iodine scrzbbing spray system. A similar calculationwithout Iodine scrubbint sprzys would increase the dose to equipmentapproxriately 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 toaccount for this factor. Therefore, the proc.edure is alsoapplicable to plants without an iodine scrubbing spray system.(3) Shielding calculations are based on an average gamma energy of1 MEY derived from TID 14844.(4) These procedures are not applicable to equipment located directlyabove the containment sump, submerged in contaminated liquids,or near filters. Doses specified for equipment located in theseareas must be evaluated on a case by case basis.(5) Since the dose adjustment factors used in these procedures arebased on a calculation for a typical pressurized water reactor witha dry type containment, they are not directly applicable toboiling water reactors or other containment types. However,doses for these other plant configurations may be evaluatedusing similar procedures with conservative dose assumptionsand adjustment factors developed on a case by case basis.ProcedureFigures I through 4 provide factors to be applied to the conservativedose 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 remainfunctional.

• , ..-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 aparticular item of equipment has been specified as 2 x 106 RADS. Theapplication specific parameters are:Reactor power level -3,000 MWthContainment volume -2.5 x 106 ft3Compartment Volume -8,000 ft3Thickness 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 equipmentcould be expected to receive in order to evaluate the adequacy of theradiation service condition specification.Step 1Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and2.5 x 10i ft3 containment volume and read a 30-day integrated dose of1.5 x 107 RADS.SteD 2Enter Figure 2 at a dose of 1.5 x 107 RADS and 24" of concrete shieldingfor 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 3Enter Figure 3 at 8,000 ft3 and read a correction factor of 0.13. Thedose 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-OlBPage-23 of 33-4-Step 4Enter 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. Applythis factor to the sum of the doses determined from steps 2 and 3 tocorrect the 30 day total dose to the equipment inside the compartmentto 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 xl1061 = 3 x 105 RADSIn this particular example the service condition of 2 x 106 RADSspecified is conservative with respect to the estimated dose of 3 x105 RADS calculated in steps 1 through 4 and is, therefore, acceptable.J

.; &nFIGURE 1K1tGAM FOR rONTAINMENT VOLUME AND REACTOR P'- mJA DOSE CORRECTIONS*CONTAINMENVOLUME (ft3xlC2x1CI Xi1o5 x 104x10V.3x16T3)~65-MWTH404o00_3000k_40DW _30 DAYINTEGRATEDYDOSE4 x 10oAttachment No. 4 to IE Bulletin 79-OlB K3 x 107_-1000II2 x107500%E2002x 10w1 x 107-I x 1O5 x 1061 _4 x 106 _3x1062.S x 1062.0 x 106 I-1 x 106_I*ISLB ACCIDENT DOSES SHOULD BE READ AS A FACTOR OF 10 LESS

DOSE CORRECTION FACTOR FOR CONCRETE SHIELDINGY( ONLY) Attachment NcoA to IE Bul108 page 25 of 33x1S x oS01 X 104I- -l10:3 fit ', Ids , Nit~to I t* 1oC 1O7 106 10I DOSE (RADS) WITHOUT SHIELDING (FROM FIGURE 1)letin 79-01BS

a106FIGURE 3 \DOSE CORRECTIN FACTOR FOR COMPARTMENT VOLUMBEAttachment No. 4 to IE Bulletin 79-O1B -*0I-zLuC0C;CDI106II I I I II ,II I0.2.4.6.81.0CORRECTION FACTOR

D URE 4DOSE CORRECTION FOR TIME hEQUIRED TO REMAIN FUNCTIONALc-V-.C.*1a,-wU.-.o Ur-.4Ju04J IAd O=: VI)al0n.C.0 .1.0.1-.I I II hIIII.01I a I II i fiI I A I I fia llI I I I I i ltI I I I l II , , I .. I ........... ....11.0101001000TIME REQUIRED TO REMAIN FUNCTIONAL MHRSP4

• - .Attachment No. 4 to IE Bulletin 79-O0Bt ' *Pale 28 of 33:APPENDI CThERMAL AND RADIATION AGING DEGRADATIONOF SELECTED MATERIALSTable C-1 is a partial list of materials which may be found in a nuclearpower plant along with an indication of the material susceptibility toradiation and thermal aging.Susceptibility to significant thermal aging in a 450C environment andnormal atmosphere for 10 or 40 years is indicated by an (*) in the appro-priate column. Significant aging degradation is defined as that amountof degradation that would place in substantial doubt the ability oftypical equipment using these materials to function in a hostileenvironment.*Susceptibility to radiation damage is indicated by the dose level andthe observed effect identified in the column headed BASIS. The meaningof the terms used to characterize the dose effect is as follows:# Threshold -Refers to damage threshold, which is the radiationexposure required to change at least one physical property ofthe material.* Percent Change of Property -Refers to the radiation exposurerequired to change the physical property noted by the percent.I Allowable -Refers to the radiation which can be absorbed beforeserious degradation occurs.The information in this appendix is based on a literature search of sourcesincluding the National Technical Information Service (NMIS), the NationalAeronautics and Space Administration's Scientific and Technical AerospaceReport (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 beconsidered all inclusive neither is it to be used as a basis forspecifying materials to be used for specific applications within anuclear plant. The list is solely intended for use by the NRC staffin making Judgements as to the possibility of a particular materialin a particular application being susceptible to significant degradationdue to radiation or thermal aging.The data base for thermal and radiation aging in engineering materialsis rapidly expanding at this time. As additional information becomesavailable Table C-1 will be updated accordingly.

11/14/79TABLE C-1THERMAL AND RADIATION AGING DEGRADATIONOF SELECTED MATERIALSr1 T IALSOASsmirnICANTAGING10 YILS 140 YRSRtADIATIONSUSCEPTIDI LITYTyPrs or rQtUiPI*N.tr (wrI~nll 10 WHIh iTEIAi, NAly lip tyoII))IA% y I C' "'hfl Ast'HKMATlt:I At.IIAS I SI- -1~ fi -A ! I I I -I I _ _Integrated Circuits JIC)Integrated Circuits IIC)C-tiSTransltorsDiodesSilicon-ControlledRectifiersIntegrated Circuits (IC)AnalogVulcanized FiberFish PaperPolyester (unfilled)NylonPolycarbondtePolywideChlorosulfonated Poly-ethylene8um-nIntegrated Circuit. (IC)104104l 14ThresholdaAllowableThresholdKKIKIxPolyamideItypalon'tSR/ti-triletubber)APikA105106610.105105106106KKKKIxxKKKKxxKIIKIKxKKIKIxxKIIKKKxKIxxxxCDP 0M SU0C* l0O CW =XTTLbiallyl PhthalateSilicone Rubbeta.I __________ I I L *Indicates that there is data available which shows a potential for significant thermal aging of the materialswhen exposed to normal operating conditions for either 10 or 40 years as indicated.

11/14/79

11/14/79I.9-v U r-MTreALALSOASrOTENTIALOR.tlCNIFICIWTAGINGi0 YM 40 YM8RRMArloNSuscePTInILITYI IsTYPES OF EQUIPAUrTr (WITHIN wiiiaC MATERIAL M"Y UK INwXI ./7 7 -4kv1a I I- I I IPolysultoneReaistora -Wire-oundResistors -CarbmrompositionCapacitors -CeramiaCapacitors -alas.Capacitora -RicaENA ThermosettingLamnatee, Oar X cHEA Thermos.ttin'Laminates, Grafe XXXP"EOA theuosettingLaminate.. Grafe XPXNm ThermosettingLaminates, Grade XPCWMR ThermoeettingLaminates, Grata XXHEt ThermoaettingLaOinate.. Grade XXPmHE ¶termosattinqLaminate., CGra XXXMhE TherrmoettingLaminate, Graft CeeOM ThermoaettingLa"nate. GCrade Cwrasde110710l19109109109109109109109l0g109109109109109fllowable24% Lossof Elonga-tionrhrerholdaAllowableUXXIKKIKIXXXIIIKIIXIIIIIINUSIIINKKI3.XKIK-0 :r-tC+IDOt0CD0@_hCf "Is-I40CI1-4IDs-O,UaXIaaUN.KI.1 L 1. .1 I ;i11/14/79* vI;_iTypes or rvQuirfl ("ITIN WIlc0 IIRTERIAI4.MAT IIe 1 09flU))109 Shre0lron5IE."IFICAPM SS~T1ILTAS 10 vp' 40 Tits GM BSI1L09 Threeholdt9103 19109109 N910109 *1010 1 It1 WENCLOSURE 2IE Bulletin No. 79-O0BDate: January 14, 1980 RECENTLY ISSUED IE BULLETINSBulletinNo.79-13(Rev. 2)SubjectCracking in FeedwaterSystem PipingDate Issued10/17/79Issued ToAll PWRs with an OLand Designated Ap-plicants (for Action),All Other PowerReactor Facilitieswith an OperatingLicense (OL) or Con-struction Permit (CP)(for Information)79-17(Rev. 1)79-2579-02(Rev. 2)79-2679-2779-28Pipe Cracks in StagnantBorated Water Systems10/29/79All PWRs with anOL (for Action). Allother Power ReactorFacilities with anOL or CP (for In-formation)All Power ReactorFacilities with anOL or CP (for Action)Failures of Westinghouse 11/2/79BFD Relays in Safety-Related SystemsPipe Base Plate DesignsUsing Concrete ExpansionBoltsBoron Loss From BWRControl BladesLoss of Non-Class-1-EInstrumentation and Con-trol Power System BusDuring OperationPossible Malfunctionof NAMCO Model EA180Limit Switches atElevated Temperatures11/8/79All Power ReactorFacilities with anOL or CP11/20/79All BWR Power ReactorFacilities with anOL11/30/79All Power ReactorFacilities with an OLand those nearingLicensing (for Action)All Power ReactorFacilities with a CP(for Information).12/7/79All Power ReactorFacilities with anOL or CP