ML14170A572

Forwards IE Bulletin 79-01B, Environ Qualification of Class IE Equipment W/Attachments.Action Required

ML14170A572
Person / Time
Site:	Robinson, Brunswick
Issue date:	01/14/1980
From:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Jackie Jones CAROLINA POWER & LIGHT CO.
References
NUDOCS 8001290418
Download: ML14170A572 (45)

Text

UNITED STATES X

NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 In Reply Refer To:

JAN 14 1980 RII:JPO 50-325 50-324 Carolina Power and Light Company ATTN:

J. A. Jones Senior Executive Vice President and Chief Operating Officer 411 Fayetteville Street Raleigh, North Carolina 27602 Gentlemen:

Enclosed is IE Bulletin No.79-01B which requires action by you with regard to your power reactor facility(ies) with an operating license. to IE Bulletin 79-01B, entitled "Interim Staff Position on Environmental Qualifications of Safety-Related Electrical Equipment",

will be forwarded at a later date.

Should you have questions regarding this Bulletin or the actions required of you, please contact this office.

Sincerely, James P. O'Reilly Director

Enclosures:

1.

IE Bulletin No.79-01B with Enclosures

2.

List of Recently Issued IE Bulletins 800 12 90 kz

JAN 14 1980 Carolina Power and Light Company

-2 cc w/encl:

A. C. Tollison, Jr.

Plant Manager Box 458 Southport, North Carolina 28461 R. B. Starkey, Jr., Plant Manager Post Office Box 790 Hartsville, South Carolina 29550

UNITED STATES SSINS No.:

6820 NUCLEAR REGULATORY COMMISSION Accessions No.:

OFFICE OF INSPECTION AND ENFORCEMENT 7910250528 WASHINGTON, D.C. 20555 January 14, 1980 IE Bulletin No.79-01B ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT Description of Circumstances:

IE Bulletin No. 79-01 required the licensee to perform a 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 containment, aging, and submergence., "GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS", provides the guidelines and criteria the staff will use in evaluating the adequacy of the licensee's Class IE equipment evaluation in response to this Bulletin.

In 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 contain ment, 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.

IE Bulletin No. 79-0IB January 14, 1980 Page 2 of 3

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 Enclosure 1)

1.

Provide a "master list" of all Engineered Safety Feature Systems (Plant Pro tection Systems) required to function under postulated accident conditions.

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

Pages 1 and 2 of Enclosure 2 are standard formats to be used for the "master list" with typical information included.

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

IE Bulletin No.79-01B January 14, 1980 Page 3 of 3

4.

Evaluate the qualification of your Class IE electrical equipment against the guidelines provided in Enclosure 4. Enclosure 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 Enclosure 3, provide a detailed "Equipment Qualification Plan." Include in this plan-specific actions which will be taken to determine equipment qualification and the schedule for completing the actions.

5.

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

6.

Submit a "Licensee Event Report" (LER) for any Class IE electrical equipment item which has been determined as not being capable of meeting environmental qualification requirements for service intended. Send the LER to the approp riate NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification. If plant opera tion is to continue following identification, provide justification for such operation in the LER. Provide a detailed written report within 14 days of identification to the appropriate NRC Regional Office. Those items which were previously reported to the NRC as not being qualified per IEB-79-01 do 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.

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

Enclosures:

1.

11 SEP Plants

2.

Master List

3.

System Component Evaluation Work Sheet Instructions

4.

Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors

5.

Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment (NUREG-0588)

Note: The above enclosures are to be sent to the corporate offices only.

SEP Plants Plant Dresden 1 Region Yankee Rowe Big Rock Point I

San Onofre 1 V

Haddam Neck I

LaCrosse Oyster Creek R. E. Ginna Dresden 2 Millstone Palisades

Facility: XYZ Docket No.:

50-XXX MASTER LIST (Typical)

(Class IE Electrical Equipment Required to Function Under Postulated Accident Conditions)

1. SYSTEM: RESIDUAL HEAT REMOVAL (RHR)

COMPONENTS Location

'Plant Identification Inside Primary Outside Primary Number Generic Name Containment Containment IPT 456 PRESSURE TRANSMITTER x

ILT 594 LEVEL TRANSMITTER x

1LS 210 LIMIT SWITCH x

M:

AUTOMATIC DEPRESS!!RIZATION SYSTEM (ADS)

I COMPONENTS I 'V V J ocation Identification Tnside Primary Outside Primary Number

.Generic Name Containment Containment B21-ROol VALVE MOTOR OPEPATOR x

821-FOO3 SOLENOID VALVE x

B21-FO1O PRESSURE SWITCH x

1~1 pill

2-III.

SYSTEM:

RHR EQUIPMENT/COMPONENTS (Typical)

• • COMPONENTS Location Plant Identification Inside Primary Outside Primary; Number*

Generic Name Containment Containment 16xP455 0-RING GASKET x

EPA, Class E, Westinghouse, lOOC ELECTRICAL PENETRATION ASSEMBLY X KULKA No. ET35 TERMINAL BOARD x

ONKONITE, 100V,, 3C 8lack POWER CABLE x

x X BRAND 1OW-40 LUBRICATE OIL 15 K569 (Boston

.Wire & Cable)

INSTRUMENTATION CABLE x

x Cutler Hamer TB rNo.

6 TEP1UNAL BOX RAYCHEM XYZ CABLE SPLICE Scotch No. 54 INSULATING TAPE x

T&B No.

0.

INSULAT )

TErtFINAL LUG

.Y Brand Epoxy No..-

SEAL-ANT x

x

• 4hen a component is not identified by plant identification numbrr, use the anufacturer, model number, serial number, etc.

• • ike components may be referenced.

AX IiI I

2X F!X iii,.

X

~X jj~X 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 qualification analysis such as submergence, qualified life (aging), or HELB. Identify in the "Notes" section on page 1 of this enclosure the actions planned for determining qualification and the schedule for completing these actions.

Page 2 of Enclosure 3 TYPI CAL

-2 SERVICE CONDITION PROFILES POSTULATED QUALIFICATION EXCEPTIONS EQUIPMENT ACCIDENT TEST ACCURACY ACCURACY OR DESCRIPTION ENVIRONMENT ENVIRONMENT REQUIREMENTS DEMONSTRATED REMARKS NOTE 1 NOTE 2 NOTE 3 NOTE 4 NOTE 5 NOTE 6 NOTES:

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

2.

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

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 tabluar form.

4.

Provide tha accuracy requirements for sensors and transmitters for trip functions and/or post accident monitoring 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.

Unit:

SYSTEM 1COMPONENT NVAWORK NSHEET Pae3o Ec-~t3 ENV IRONMENT

)DOCUMENTATION REF*

ILFCAIOOUSAI EUPETParameter fSpecifi-lQualmf-Spc uail METHOD ITEMS Syste.:

RHR Ccrating 15 !ain.

2C0 min.

15

~ S multaneou~

None Ph..it rD Nlo.

IPT456 Time IITest Component:

i____________

PtiFSSUTEST PROFILEST~f ~~

Test None Mnufacture:

-IPROVIDED t_____I______

Fischow-Porter Co.

Pressure 1

~ mltnol Nn Model Number:

5.IA I__ __

_T e!

50-En-1071-BCXN-NS

.1 u

10O

-- Test IAccidnnt Mlonitoring 1

1 1 PiuTeOL Non C' ici1

• 3 33 Accuracy:

Spec:

5%

J Spray NO Damon 4% NOH 1See Notel D~i

Pr.

Ufiip IA

adiaticn, WO 4x rads 11.2xI108d 2

6 Sequential S /~ r.

Pi ' s u e J0 7_

_ _ _ _ _ f a j T e s t N o n e S/N O7I.

Squent I1 LoainInj.iitnt A

I ng 10ys 0Is37 Test None

[oc~ir~: Crnti~j~nt A~ing j1Oys ~ O YS 37,8 2._Eng. Anapsis

!20 177 Sa bot No Nn r~~ w _________ a uuf red 10 I,

Se

_VNote

.. ti ;

Notes:

1. 1..

.tpr 3,1 Paragraph 14 23.7

7. XYZ Letter No. 237-1, dated November 2,199, tee hn sent to MFG. requesting the qualification.

4.Rajieaph Ai~i); !:A'1 I'i~

Ar I

-aPcn If qt!Thlfication not determined

4.

to QC dtCdNoeiie 2 17 dtabl a by C-c&-xbcr 15, 1979,,component 5.FLnL fo~oi Roori JtdNvme

,17 will be replaced during refuialing outage March 1980.

6. Fischcr and "oi'~

C 7as't Raport No. 2500-1

.8. Wylie Laboratory Report no. 467 an environrrnental parameter.

A~BC Laboratory is to perform submergence test in April 1980.

(Total 33 pages)

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. TemDerature and Pressure Steam Conditions

2. Radiation

3. Submergence

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

1. Temperature, Pressure and Relative Humidity

2.

Radiation

3.

Submergence

4.

Chemical Sorays

-(1)

(1)

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

1. Simulated Service Conditions and Test Duration

2. Test Specimen

3. Test Sequence

4. Test Specimen Aqing

5. Functional Testing and Failure Criteria

6.

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

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

1.0 INTRODUCTION

On February 8, 1979, the NRC Office of Inspection and Enforcement issued IE Bulletin 79-01, entitled, "Environmental Qualification of Class IE Equipment." This bulletin requested that licensees for operating power reactors complete within 120 dAys their reviews of equipment qualification begun earlier in connection with IE Circular 78-08. The objective of IE Circular 78-08 was to initiate a review by the licensees to determine whether proper documentation existed to verify that all Class IE electrical equipment would function as required in the hostile environment which could result from design basis events.

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

These 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.

-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 IEEE Std. 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."

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

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

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

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

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,

-4 4.0 SERVICE CONDITIONS In order to determine the adequacy of the qualification of equipment it is necessary to specify the environment the equipment is exposed to-during normal and accident conditions with a requirement to remain functional.

These environments are referred to as the *service conditions."

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

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

1.

Temperature and Pressure Steam Conditions

• In general, the containment temperature and pressure conditions as a function of time should be based on the analyses in the FSAR, In the specific case of pressure suppression type containments, the following minimum high tempeature conditions should be used: (1).BWR Drywells. 340 0F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and (21 PWR Ice Condenser Lower Compartments 340oF for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />,.

2.

Radiatton - 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 submerged in contaminated liquids must be evaluated on a case by case basis., Guidelines for these evaluations are not provided in this document.,

Gamma Radiation Doses - A total gamma dose radiation servce condition of 2 x.10 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 8, The procedures in Appendix B are based on the calculation for a typical PWR reported in Appendix D of NUREG-05881.

Gamma dose radiation service conditions for BWRs and PWRs with ice condenser containments must be evaluated on a case by case basis, Since the procedures in Appendix B are based on a calculation for a typical PWR with a dry type 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, Beta Radiation Doses - Beta radiation doses generally are less significant than gamma radiation doses for equipment qualification. This is due to the low penetrating power of beta partiles in comparison to gamma rays of equivalent energy, Of the general classes of electrical equipment in a plant Ceg., cables, instrument transmitters, valve operators, containment penetrations), electrical cable is considered the most 1NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.

• 6 vulnerable to damage from beta radiation.

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

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

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

7q.

4.

Containment Sarays - 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, 4i2 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. Temcerature and Pressure Steam Conditions - Equipment qualified for a LOCA environment is considered qualified for a MSLB accident environ ment in-plants with automatic spray systems not subject to disabling sincle component failures. This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a thermal analysis of typical components inside containment."/

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

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

Class IE equipment installed in plants without automatic spray systems or plants with spray systems subject to disabling single failures should be qualified for a MSLB accident environment dete-mined by a plant specific analysis.

Acceptable methods 1See NUREG 0458, Short Term Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation.

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 E1ectrical Equipment.

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

3.

Submeroence - Same as Section 4.1 above.

4.

Chemical Sprays -

Same as Section 4.1 above.

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

Service conditions for areas outside containment exposed to a HELB were evaluated on a plant by plant basis as part of a program initiated by the staff in December, 1972 to evaluate the effects of a HELB. The equipment required to mitigate the event was also identified. This equipment should be qualified for the service conditions reviewed and approved :n the dikLB Safety Evaluation Report for each specific plant.

4.3.2 Areas Where Fluids are Recirculated from Inside Containment to Accomolish Lono-Term Core Cooling Followino a LOCA

1. 7er-Derature and Relative Humidity - One hundred percent relative humidity shouTd be established as a service condition in confined spaces. The temperature and pressure as a function of time should be based on the plant unique analysis reported in the FSAR.

2. Radiation -

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

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

3.

Submergence - Not applicable.

4. Chemical Sprays - Not applicable.

4.3.3 Areas Normally Maintained at Room Conditions Class IE equipment located in these areas does 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 systems served by the onsite emergency electrical power system.

Equipment located in areas not served by redundant systems powered from onsite energency 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 QUALIFICATION METHODS

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

As a minimum, the qualification for severe temperature, pressure, and steam service conditions for Class IE equipment should be based on type testing.

Qualification for other service conditions such as radiation and chemical sprays may be by analysis (evaluation) supported by test data (see Section 5.3 below). Exceptions to these general guidelines must be justified on a case by case basis.

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

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

The time duration of the test should be at least as 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

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

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

3.

Test Sequence - The component being tested should be exposed to a steam/air environment at elevated temperature, and pressure in the sequence defined for its service conditions. Where radiation is a service condition which is to be considered as part of a type test, it may be applied at any time during the test sequence provided the component does not contain any materials which are known to be susceptible to significant radiation damage at the service condition levels or materials whose susceptibility to radiation damage is not known (see Appendix C). If the component contains any such materials, the radiation dose should be applied prior to or concurrent with exposure to the elevated temperature and pressure steam/air environment. The same test specimen should be used throughout the test sequence for all service conditions the equipment is to be qualified for by type testing. The type test should only be considered valid for the service conditions applied to the same test specimen in the appropriate sequence.

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

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

If the component contains such materials a qualified life for the component must be established on a case by case basis.

Arrhenius techniques are generally considered acceptable for thermal aging.

5.

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

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

6. Installation Interfaces - The equipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conclusive.

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

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

In such cases the overall qualification is said to be by a combination of methods. Following are two specific examples of procedures that are considered acceptable. Other similar procedures may also be reviewed and found acceptable on a case by case basis.

1. Radiation Oualification - Some of the earlier tvoe tosts 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.

Cher-ical Spray Qualification - Components enclosed entirely in corrosion resistant cases (e.g., stainless steel) may be shown to be qualified for a chemical environment by an analysis of the effects of the particular chemicals on the particular enclo sure materials., The effects of chemical sprays on the pressure integrity of any gaskets or seals present should be considered in the analysis.

14 6.0 Margin IEEE Std. 323-1974 d-:ines margin as the difference between the most severe specified service conditions of the plant and the conditions used in type testing to account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance.

Section 6.3.1.5 of the standard provides suggested factors to be applied to the service conditiohs to assure adequate margins. The factor applied to the time equipment is required to remain functional is the most significant in terms of the additional confidence in qualification that is achieved by adding margins to service conditions when establishing test environments. For this reason, special consideration was given to the time required to remain functional when the guidelines for Functional Testing and Failure Criteria in Section 5.2 above were established. In addition, all of the guidelines in Section 4.0 for establishing service conditions include conservatisms which assure margins between the service conditions specified and the actual conditions which could realistically be expected in a design basis event. Therefore, if the guidelines in Section 4.0 and 5.2 are satisfiedno separate margin factors are required to be added to the service conditions when specifying test conditions.

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

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

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

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

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

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

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

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

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

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

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

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

This assumption and all the dose rates used in the procedure out lined below are based on the methods and sample calculation described in Appendix D of NUREGT0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equip ment." Therefore, all the limitations listed in Appendix D of NUREG-0588 apply to these procedures.

(2) The sample calculation in Appendix D of NUREG-0588 is for a 4,000 MWth pressurized water reactor housed in a 2.52 x 106 ft3 contain ment with an iodine scrubbing spray system. A similar calculation without iodine scrubbing sprays would increase the dose to equipment approximately 15%.

The conservative dose of 2 x 107 RADS assumed

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

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

(4) These procedures are not applicable to equipment located directly above the containment sump, submerged in contaminated liquids, or near filters. Dbses 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 1 through 4 provide factors to be applied to the conservative dose to correct the dose for the following plant specific parameters:

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

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

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

Reactor power level - 3,000 MWth Containment volume -

2.5 x 106 ft3 Compartment Volume -

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

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

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

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

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

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

Step 3 Enter Figure 3 at 8,000 ft3 and read a correction factor of 0.13. The dose due to sources inside the compartment would then be 0.13 (1.5 x 107)

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

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

--4 SteD 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 I hour.

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

FIGURE I NOMOGRAM FOR NTAINMENT VOLUME AND REACTOR POW A DOSE CORRECTIONS*

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

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

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

.Susceptibility to radiation damage is indicated by the dose level and the observed effect identified in the column headed BASIS.

The meaning of the terms used to characterize the dose effect is as follows:

• 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.

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

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

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

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IE Bulletin No.79-01B Enclosure January 14, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.

80-01 Operability of Ads Valve 1/11/80 All BWR power reactor Pneumatic Supply facilities with an operating license 79-0IB Environmental Qualifica-1/14/80 All power reactor tion of Class IE Equipment facilities with an operating license 79-28 Possible Malfunction of 12/7/79 All power reactor Namco Model EA 180 Limit facilities with an Switches at Elevated OL or a CP Temperatures 79-27 Loss Of Non-Class-1-E 11/30/79 All power reactor Instrumentation and facilities holding Control Power System Bus OLs and to those During Operation nearing licensing 79-26 Boron Loss From BWR 11/20/79 All BWR power reactor Control Blades facilities with an OL 79-25 Failures of Westinghouse 11/2/79 All power reactor BFD Relays In Safety-Related facilities with an Systems OL or CP 79-17 Pipe Cracks In Stagnant 10/29/79 All PWR's with an (Rev. 1)

Borated Water System At OL and for information PWR Plants to other power reactors 79-24 Frozen Lines 9/27/79 All power reactor facilities which have either OLs or CPs and are in the late stage of construction 79-23 Potential Failure of 9/12/79 All Power Reactor Emergency Diesel Facilities with an Generator Field Operating License or Exciter Transformer a construction permit 79-14 Seismic Analyses For 9/7/79 All Power Reactor (Supplement 2) As-Built Safety-Related Facilities with an Piping Systems OL or a CP 79-22 Possible Leakage of Tubes 9/5/79 To Each Licensee of Tritium Gas in Time-who Receives Tubes pieces for Luminosity of Tritium Gas Used in Timepieces for Luminosity