ML17249A983

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

ML17249A983
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Site:	Millstone
Issue date:	02/15/1980
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TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8003130395
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ENCLOSURE I GUIDELINES FOR EVALUATING ENVIRONMENTAL UALIFICATION OF CLASS IE ELECTRICAL E UIPMENT IN OPERATING REACTORS

1. 0 Introducti on 2.0 Discussion 3.0 Identification of Class IE E ui ment 4.0 Service Conditions 4.1 Service Conditions Inside Containment for a Loss of Coolant Accident LO 1.

Tem erature and Pressure Steam Conditions 2.

Radiation 3.

~33 4.

~CS 1

1S 4.2 Service Conditions for a PWR Main Steam Line Break MSLB Insi e Containment 1.

Tem erature'and Pressure Steam Con"itions 2.

Radiation 3.

~33 4.

~bh 4.3 Service Conditions Outside Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a Hi h Ener Line Break HELB 4.3.2 Areas Where Fluids are Recirculated From Inside Containment to Accom lish Lon -Term Emeraenc Core Coolin Followin a

LOCA 1.'emperature, Pressure and Relative Humidit 2 ~

Radiation

~b 4.

~bh.,i 1S

4.3.3 Areas Normall Mai.tained at Room Conditions 5.0 uglification Methods 5.1 Selection of uglification Method 5.2 ualification b T

e Testin l.

Simulated Service Conditions

'and. Test Duration

2. T~S

3. i~S 4.

Test S ecimen A in 5.

Functional Testin and Failure Criteria 6.

Installation Interfaces 5.3 gualification b a Combination of Methods Test Evaluation nal sis 6.0

~ar<ain

7. 0

~A~in 8.0 Documentation Appendix A - Typical Equipment/Functions Needed for Miiigaz'.on of a

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

GUIDELINES FOR EVALUATING ENVIRONMENTAL UALIFICATION OF CLASS IE ELECTRICAL E IPMENT IN OPERATING REACTORS

1.0 INTRODUCTION

On February 8,

1979, the NRC Office of Inspection and Enforcement issued IE Bulletin 79-01, entitled, "Environmental gualification 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'eviews 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'esponses 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'hese guid lines are intended to be used by the NPC staff to evaluate the qualification methods used for existing equipment in a particular class of plants, i.e., currently operating reactors including SEP plants.

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Equipm~~t 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 gualification of Safety-Related Electrical Equipment.

In addition to its reviews in connection with IE Bulletin 79-01 the staff is engaged in'ther 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. 323-1974 is the current industry standard for environmental qualification of safety-relat'ed 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

'th'e 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 Criter'a IEEE Std. 323-1974, "IEEE Standard for gualifying Cl'ass IE Equipment for Nuclear Power Generating Stations."

~E 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 compati'ble 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 E UIPYiENT 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 wat'er reactor designs to perform these funct'.ons for the most severe postulated loss of coolant accident (LOCA) and main steamline break accident (MSLB) are listed in Appendix A.

gore detailed descri pti ons of the Class IE equipment installed at specific plants can be obtained from FSARs, Technical specifications, and emergency procedures, Although variation in nomenclature may,exist at the various plants, environmental qualification of those systems which perform the functions identified in Appendix A should be evaluated against the appropriate service C

condi-ions (Section 4.0),

Tne g" celines in this document are ap..'iicabie to all components necessary for operation of the systems listed 'in Appendix A includin" but noi limited to valves, motors, cables, connectors, relays, switches, transmitters and valve position indicators,

- 4

4. 0 SERV ICE CONDITIONS In order to determine the adequacy of the qualification of equipment't is necessary to specify the environment the equipment i's 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 condi'tions specie'ied in the FSAR or other licensee submittals are acceptable, unless other'wi'se noted in the guidelines discussued

below, 4.1 Service Conditions Inside Containment for a Loss of Coolant Accident LOCA 1,

Tem erature and Pressure

'Steam Conditions - In general, the containment temperature and pressure conditions as a function of time should be based on the analyses jn the

FSAQ, In the specific case of pressure suppression type containments, the following minimum high tempeature conditions should be used:

(1)

BWR Drywells 340oF for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and (2).

PHR 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,;

Radiation - When speci'fying radiation service conditions for equipment exposed to radiation during normal operating and accident conditions, the nominal operating dose should. be added to the dose received during the course of an accident.

Gui'deljnes for evaluating beta 'and gamma radiation service conditions for general areas inside containment are provided below, Radiation service condlti,ons for equipment located directly above the containment sump, in the vicinity of filters, or submerged i.n contaminated liquids must be evaluated on a case by case basi,s, Guidelines for these evaluations are not provided in this

document,

Gamma Radiation Doses - A total gamiia dose radiation service condttion of 2 x 10 RADS is acceptable for Class IE equipm.tt located in general V

areas inside containment for PWRs with dry type contaimments, Nhere a dose less than this value has been specified, an application specific evaluation must be performed to determine if the dose specified

$ s acceptable.

Procedures for eValuatirig radiation service conditions in such cases are provided in Appendix B, The procedures in Appendix B are based on the calculation for a typical PWR reported iri Appendix D of NUREG 0588 Gamma dose radiation ser vice conditions for BWRs and PHRs with ice condenser containments must be evaluated on a case by case

basis, Since the procedures ia Appendix B are based on a calculation for a typicatl PWg with a dry type contai'nment, they are not directly applicable to BWRs and other. containment types,

However, doses for these other plant configurations'may be evaluated using similar procedures with conservati've dose assumptions and adjustment factors developed on a

case by case bpsi's, 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 parti'cles in comparison to gamma rays of equivalent

energy, Of the general classes of electrical equipment in a plant (e,g,,

cables, i'nstrument transmitters, valve operators, C

containment penetrations),

electrical cable is considered the most NUREG-0588, Interim Staff Position on Environmental gualification of Sa fety Rel a ted Electri ca 1 Equi pmen t, 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 10 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 10 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 i'tern 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 gama and beta doses.

3.3~h.f3p f

d thd fp t tf gf ttt ff t 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.

4.

Containment S ra s - Equ(pment exposed to chemical spr ys should be qualified for the most severe chemical envtronment factdic or basic) which could exist, Demineralized water sprays should not be exempt from consideration as a potentially adverse service condition, 4.2 Service Conditions for a PMR 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 sane 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.

Tem erature and Pressure Steam Conditions - Equipment qualified for a

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

This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a thermal analysis of typical components inside containment.1/

The final acceptability

o. this approach, i.e.,

use of the "Best Estimate",

is pending the completion of Task Action Plan A-21, Hain Steamline Break Inside Containment, Class IE equipment installed in plants without automatic spray systems or plants witJ spray systems subject to disabling single failures or delayed initiation should be qualified for a HSLB accident environment determined by a plant specific analysis, Acceptable methods

'See NUREG p458, Shor:

Term Sa ety Assessment on the Environmr<<a

~

gualification of Safe'.~-Related Electrical Equipment of SEP Opera.ing Reac:ors, for a more de:ailed discussion of the best estimate calculation.

I for performing such an analysis for operating reac':.nrs are provided in Section 1.2 for Category II plants in NUREG-0588, Interim Staff Position on Environmental gualification of Safety-Related E li!ctrical Equipment.

2"; 'Radiation - Same as Section 4.1 above except that a conservative gamma dose of 2 x 10 RADS is acceptable.

3.

Submer ence - Same as Section 4.1 above.

4.

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-S S

i 4.1 4.3 Service Conditions Outside of Containment 4.3. 1 Areas Sub 'ect to a Severe Environment as a Result of a Mi h Ener Line Break MELB Service'conditions for areas outside containment exposed to a

MELB 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 evert

~:as is> -'.dent!..'.ed.

This equipment should be qualified for the service conditions reviewed and approved

,n the M.'

Sa<e

.y Evaluation Report for each specific plant.

4.3.2 Areas Where Fluids are Recirculated from Inside Containment to Accom lish Lono-Term Core Coolin Followin a

LOCA 1.

Tem erature and Relative Humidit One hundred percent relative humidity should be established as a service condition in confined spaces.

The t

temperature and pressure as a function of time should be based on the

.':t.'. 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 10 RADS would be expected.

N.

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ppl 1 bl N.

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1 1

N ppll bl 4.3.3 Areas Normall Maintained at Room Conditions h

V ~

NNP Class IE equipmen.

iocated in these areas does no', experience significant stress due to a change in service conditions during a design basis event.

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

ANSI, NEMA, National

.Electric Code).

Based on these factors, failures of equipment in these a -eas during a design basis event are expected to be random except to

="=- extent that th.".'ay be due tc aging or failures of ai~ car."-'ti"."i~", or ventilation systems.

Therefore, no special consideration need be given to tne environmental qualification of Class IE equipment in these areas provided tne aging requi rements discussed in Section 7.0 below are satisfied and the areas are maintained at room conditions by redundaiit air conditioning or-ventilation systems served by the onsite emergency electrical power system.

E".ipment located in areas rot served by redunCant systems power'ed from onsite emergency sources should be qualified for the environmental extremes w",ict'ould result from a failure of the systems as determined from a plant s===i i: analysis, 0'w'F ICATION M THODS 5.1 Selection of uglification 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

's the preferred method of qualification for electrical equipment located inside containment required to mitigate the consequences of design basis l

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.

gualification for other service conditions suck as radiation and chemical sprays may be by analysis (evaluation) supported by.est data (see Section 5.3 below).

Exceptions to these general guidelines must be,iustified on a

case by case basis.

5.2 Oualification b T

e Testin The evaluation of test plans and results should include consideration of the following factors:

1.'imulated Service Conditions and Test Duration - The envi ronment 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

".c"g as

he period from the initiation of the accident until the tempera ure and pressure service conditions return to essentially the same levels that existed before the postulated accident.

A shorter test duration may be acceptable

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

2. T~Si

-Th <<p 1

h 1db th d1 th equipment being qualified.

The type test should only be considered valid for equipment identical in design and material construction to the test specimen.,

Any deviations should be evaluated as part of the qualifica-tion documentation (see also Section 8.0 below).

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-1 p

tb1g d

h 1db p

d 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 du'ring 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'ervice 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 Aoinc - Tests which were successful using test specimens which had net bee..=r=aged mav be considered accep.able provided the component does not contain materials which are known to be susceptible to significant degradation due to thermal and radiation agir (see Section 7.0).

If the component contains such materials a qualified life for the

'omponent must be established on a case by case basis.

Arrhenius techniques are generally considered acceptable for thermal aging.

6.

Functional Testin and Failure Criteria - Operational modes t'ested

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 instal;lation 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 penetr'ations in equipment housings for electrical connections being left unsealed or susceptible to

. moisture incursion through stranded conductors.

5.3 ualification b

a Combination of Methods

Test, Evaluation

~Anal sis 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 foun."'cceptable on a case by case basis.

1.

Radiation Oualification -

Some of the earlier tvoe tests performed for operating reactors did not include radiation as a service condition.

In these cases the equipment may be shown to be radiation qual is'.ed by perfo~ine a calcalati",n of the d"se

expected, taking into account the time the equipment is required to remain functional and its location using the methods described in Appendix 8, and analyzing the effect of the calculated dose on the materials used in the equipment (see Appendix C).

As a

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

2.

Chemical S rav uglification - Components enclosed ertirely in or-. fusion resistant cases

(.g., s.ainless steel) may be shown to be qua',ified for a chemical env'onment by ar a~alvsis of the ef ec.s of the particu ar che;icals on the par=i=i,'ar enclo-sure materials, The effec.s of che-.ical sprays or the pressure integrity of any gaskets or seals present should be considered i~ the analysis.

6.0

~Mar in 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 conditions to assure adequate margins.

The factor applied to the time equipment is required to remain functional is the most significant in terms of the additional confidence in-.qualification that is achieved by adding margins to service conditions when establishing 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 satisfied,no separate margin factors are required W

to be added to the service conditions when specifying test conditions.

7,0

~Ann Implicit in the staff position in Regulatory Guide 'I.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

spec Ific qualified life be demorstrated for all Class IE equi pme it is not sufficient to justify the exponse for plants already constructed ard operating.

This position does not, however, exclude equipment

4I 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 susceptabi lity to thermal and radiation aging.

S.O 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 comp:',=nce with a de"ign specification should no.

be considered adequate.

t

APPENDIX A TYPICAL E UIPMENT FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR NSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power Emergency Core Cooling 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 Mater Emergency Shutdown Pos Accident Sampling and Monitoring Radiation Monitoring Safe.y Related Display Instrumentation

1 These systems kill differ for PMRs and BNRs, and for old"r and newer plants.

In each case the system features which allow f.

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.'xamples of such systems and equipment are the RHR system,

PORVs, RCIC, pressurizer

sprays, chemical and volume control system, and steam dump systems.

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

APPENDIX B PROCEDURES FOR EVALUATING GAMMA RADIATION SERVICE CONDITIONS Introduction and Dis cus si on The adequacy of gamna 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 gambia 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 10 RADS for a LOCA and

2. x lb6 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 descri'bed ia Appendix D of NUREG-0588,

" Interim Staff Position on Environmental gualification 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 MMth pressurized water reactor housed in a 2.52 x 10 ft contain-ment with an iodine scrubbing spray system, A similar calculation v<<thou'. iodine scrubbing sprays v:"u'i-'ncrease the dose to equipmen approximately 15%.

The conservative dose of 2 x 10 R.-".3S assumed

2 in the procedure below includes sufficient conservatism to account for this factor.

Therefore, the pro ~dure is also applicable to plants without an iodine scrubbing spray system.

(3)

Shielding calculations are based on an average gamma energy of..

1 NEV derived from TID 14844,

'(4)

These procedures are not applicable to equipment located directly above the containment

sump, submerged in contaminated liquids, or near filters.

Doses specified for equipment located in these areas must be evaluated on a case by case basis.

(5)

Since the dose adjustment factors used in these procedures are based on a calculation for a typical pressurized water reactor with a dry type containment, they are not directly applicable to boiling water reactors or other containment

types, However, doses for these o.h'er plant configurations may be evalua+

d using similar procedures w.'th conserva+ive dose assumotions and adjustment factors developed on a case by case

basis, If 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.

k Consider the following case.

The radiation service condition for a h

ll particular item of equipment has been specified as 2 x 10 RADS.

The application specific parameters are:

Reactor power level - 3,000 MWth Containment volume - 2.5 x 10 ft Compartment Volume - 8,000 ft3 Thickness of compartment shield wall (concrete)

- 24" Time equipment is required to remain functional -

1 hr.

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

~Ste Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and

~.5 x 10 ft containment volume and read a 30-day integrated dose of 6

3 1.5 x 107 RADS

~Ste 2

Enter Figure 2 at a dose of 1,5 x 10 RADS and 24" of concrete shielding for the compartment the equipment is located in and read 4.5 x 10 RADS.

This is the dose the equipment receives from sources outside the compart-ment.

To this must be added the dose from sources inside the compartmen.

(Step 3).

~Ste 3

Enter Figure 3 at 8,000 f:

and read a correction factor of 0.13.

The dose due to sources inside the compartment would then be 0.13 (1.5 x 137)

= 1.95 x 10 RAOS.

The sums of the doses from steps 2 and 3 equals:

4.5 x 104 RAOS + 0.13 (1.5 x 10

)

RAOS

= 2.0 x 10 RAOS

~Ste 4

'I Enter Figure 4 at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and read a

orrection factor of 0.15.,Apply this factor to the sum of the doses determined from steps 2 and 3 to I ~

correct the 30 day total dose to the equipment inside the compartment to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

~

0.15 (2.0 x 106)

= 3 x 10 RADS In this particular example the service condition of 2 x 10 RADS specified is conservative with respect to the estimated dose of 3 x 10 RADS calculated in steps 1 through 4 and is, therefore, acceptable.

FIGURE 1

NOMOGRAM R CONTAINMENT VOLUME AND REACTOR OWER LOCA DOSE CORRECTIONS*

ONTAIN NIENT VOLUME (ft3) 3x 106 2x 106 1 x106 MW 4000 30 DAY INTEGRATED yDOSE 4x10 5x105 4x105 3x105 3 x107 2 xl0 2 x 1(P 1 x 107 1 x105 5x 106 4 x106 3 x 106 2.5 x 106 2.0 x 106 1 x106

• MSLB ACC 0".':

DPS"=S S~~V' E'"-

RE-'..".

<'. -'<.T3f

':,=

1C L.:SS

DOSE ORRECTION FACTOR FOR G NCRETE SMIELD(Nci

( v ONLY)

I 1 x 107 Uz Q

LLI 1 x 106 MxI-M Q

cL'

'1 x10>

0Q 1 x104 1 x103 107

FIGURE 3 DOSE CORRECTION FACTOR FOR COMPARTMENTVOLUME D

00 105 z

I-Q 0

.6

.8

".0

FiGURE 4 DOSE CORRECTION FOR TrSIE nEQUIRED TO REMAIN FUNCTIONAL

'3

":.0 Pl v)

~ g i3 lpga V)0D D

.01 1:0 10 100 TIIVIEREQUIREf. TO REIVIAIN'UNCTIONAL(HRS)

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

Susceptibility to significant thermal aging in a 45 C environment and normal atmosphere for 10 or 40 years is indicated by an (*) in the appro-priate column.

Significant aging degradation is defined as that amount of degradation that would place in substantial doubt the ability of typical equipment using these materials to function in a hostile environment.

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

The meaning of the terms used to characterize the dose effect is as fol;ows:

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

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

9 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 Na.ional Technical Information Service (NTIS), the National Aeronautics and Space Ad"..inistration's Scientific and Technical Aerospace Report (STAR), NTIS Government Report Announcements and Index (GRA), and

various manufacturers data reports.

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

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

'I due to radiation or thermal aging.

The data base for thermal and radiation aging in engineering materials is rapidly expanding at this time.

As additional information becomes available Table C-1 will be updated accordingly.

TABLE C-1 TNERMAL AND RADIATION AGING DEGRADATION I

1 /1(1/ /9 OF SELECTED NTERIALS tlhTI:Illhl ~

hl ':I) rtt(MI h!~

In vt".

Zn VII.,

Ilr!1Srrlhf, Ivm Oi I (SI II'IC'hIIT h(IIth)

I(hi!9 (PINNA llhsiI6 Ilh Ill hTt(III OII!ICI'.I'TJIIItilTV tlhT!'!!Clht. Nhv Ill! Ivu>>sl) c(C

~)

()I Crn cC pc y

Ce un?

c/Pc 4'y Cc j> cy 1

Cci Cc I

c'l p

cc" ', !i )

4c (act I

P<~n.'

~~

I(

1

(+y CI y cr V Cc 'I q; A C,

rg Cl Tvl'I:8 ol':C)till'ltl!IrrIHI'I'IIIIIIIIII(st l>>tc i)rut(xi Circ>>Its IIC)

II I(I).'I I>>lc cl ~ >>toil Circ:ults IIC)

~

tlc OI

'r!.>>I ~Il>>I I>> o Illcxlil'I

I I lc:i>>l-Co>>t Iol loci I:~ c CIIIrr!C tilt(qihte I cllcults Ifc) h>>.llcrl 10 10 IO 10 IO 10 Threshold Vole:1 Ilhocl Fllx!r

~ ish Pstior I'iilyc!zt.ol

(>>uftllcct)

Ilylc>>l I'oIyilIrtxulilt,c I'illyI xl l etc I lllcllosulfo>>1toct Poly "Ihy I v>>cl Itic>>u 'I I>>Ic ~lrhtccl Clrcultcl IIC)

'I I'I~

'o1yor!lc)e lypslo>>

IIIIIPII-

.rJlo tilhhor 10 10 lo'0 IO IO IO'0 hllou(!bia X

Threshold X

X X

X X

X X

X X

X X

X III.IIIyl Phthslnto

'llI Ic:o>>Q Itulllx!c

)hP IO 10

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

TYI'E. Ok':gcttl'tctttfr (WITIIIIIWtlf<XI IIATk!Icthl, Nhy f<1: Vlktttllt)

IIATNICIAir SBIC Bobber At<SO Btl<nttI AS FOTEIITIhlr FOR Si I<BII F I I >NT h< ~ 111< ~

IO Ylt

>10 YB.

BAD.>

c~

BASIS 10 Bhl)I ATION SilISCt:I'Tl0 II ITY Op Qo Qr>

i i

g

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(+ )

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n

~

'g

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fi 0 W tr c'>Is>el tore - Tn<>tal<s<>

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I 'f<c I I

I

) ~:. I'. I'l>anni lo 9-4050 I >< S I I Iso>><r

< y<<.>l >o lhl<S)

I>>1> 0<at<el Clteults {IC)

I:<'I ~

I

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> <>lynnter Olnns

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> ~ > I yc I by Ieno

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> ro

~ '> ~ >nn l,l>>k<>d

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>'nli.>i I I<>< n - Nylnr

tbylene-

'> opy lr<>e

<<d>I><'r 10 10 10 10 10 10 10 10 10 10 10 10 10 10 IOY 10 hIlcn<nb Io Tl>resbold IG% Loss of Elon<)n t1on brest>old A11ovnbl o Tbresl>old 11oc<ablo X

X X

rent>el<'I 11ounble X

X X

X X

X

11/14/79 Ith'(7'!IIIAt.

AI~<<)

F.0( >at(I AS I(t YRS 40 YRS 17rfl tfflhli FOR S I(:NIFICANT AGIIIG Itht>S<

(sat(HA nhgtS Rhl >I ATION SUSCCI'TI BII<ITY TYI'I'Si (>F I:QttlI'Half It(ITIIINIIIII(7!HATI!Rlhh HAY ltt! 17>(tttt>)

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~.

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()

AP A n, t(t>

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I

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~ >>tS>>>! II Ion II'l>>><t tors - C<>rn<tlc

>!> >c> tc>rs Glnss

">I>>><"ters Hlca tit:Hh Thor><>onotttng I.>><>li>i>ton, (:an<In XXXPC Itl:tth '7>or<<><>sc>ttla><t t.>>sin>tos, 0m<to XXXP I>t:tth ",lwr><insetting I >><>In ates, Grado XPX FFJIA 'fhoa><insetting 17<<clnatns, Grado XPC lll:.Nh TI>stt>tnsottltaq I

><<>In alen>

(i>a<le xx ttt(HA fh< ~ a t>>>sot ting I.w<<>In ates, Ga n<te XXP

>I> tth fhe>><>>>setting I !.<nl ~

> ~ C n, 0>'nit<! XXX rn><>Ido 107 10 10 IO 10 10 10 10 lng 10 10 10 10 10 I Iowat >16 24s Il>ns

>E F.longn.

tlon hrcshol>able X

X X

X X

x Itilt(h Ie>><>I >>

>.t:tah I I ><ill I'hn>><>naut tI ng

ates, Grn<ln CH fhrt><conn\\ I lng

aton, G>nate!

C 10 10

TYPFS Ill VQ>ill'III'.NT lt>f11>IN IIIII(11 Hh'll'.Rlhl, HAY IIN VIP>>HI>)

Nh'I'I.'llI AI<

hleSI)

KI>I>>IN hge 14m!NTIAI~

I OR SIGNIFIC:htrr h> < IN>i IO YBni 40 Yl>S I>AI)S r~lh BhgilS I>hi

>IATION SUSOFI'Tl 8 IlilTY OO'

)3 a()

<s n.

)

~

4 <<'

4> cr n'y 4

O

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II:Ilh l1>elmosrttl>lg I '<<<< Inttn>l~

I ~lie'de Ie Nl:tth 'I'hi >mes<tnt ting I >ml>sst e!I ~

I l<ld M

>It)HA rhett s<lttlng I )min ttOS, I;rade FR-2 l>1:HA 'rhr>mo:I<sttls>g

.'min)ton, I;railo VB-3 t>>A><> rhnsm<>I><!ttlng l>>min>I<'n,

>e>ad< ~ VR-4 III:t>A rh<!>me>set tlg I>>min>tee, (;rado Flt-10 tlu>A rhe! em<<selt>.lng I.>min.tins, I:sado h t>t!Ith 'fhosmosett lng i~>min >les, Iirndo AA i ~ <<el>A rl>otmns<.ttlng I,I',>mls>st.os, I;rado I:-3 VIAL fhnrt<>osettltsg I't>>i>>>I><S, Ctade C-ll

>i>l:1>A 'rhetmosettlt>g I <min>>.ns, Grade VR-5 I < ~I>ls'os >I - Film 10 Threshold 9

10 10 10'0 10 10 10 10 9

10 10 1010

ENCLOSURE 2

GUIDELINES FOR IDENTIFICATION OF THAT SAFETY E UIPMENT OF SEP OPERATING REACTORS F

R WHICH ENYIRONMENTAL UALIFICATION IS TO BE ADDRESSED For operating reactors, all electrical equipment needed to mitigate high energy line breaks (LOCA, MSLB, FWLB) inside or outside containment should be qualified to perform the required safety function in the accident environment.

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

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

(1)

Safety functions typically performed by plant safety systems are listed in Appendix A.

For each safety function identified in Appendix A, list the systems, sub-systems, or components assumed available in the plant FSAR or emergency procedures to perform that

.function during a

LOCA, MSLB inside or outside containment, or FWLB inside or outside containment.

If a plant specifi c safety function not listed in Appendix A is identified, that function and the corresponding systems or equipment.to perform the function should be added to the licensee's list.

(2)

All the systems or equipment implied in Item (1) should be identified regardless of the original classification of the equipment when the plant received its operating license; i.e.,

some control grade equi pment will probably be named i n emergency procedures,

However, if plant emergency procedures specify a preferred mode of accident mitigation involving equipment recognized by the licensee as unlikely to meet environmental qualification criteria, an alternate mode of performing the safety function with potentially qualifiable equipment may be identified.

In such cases, the emergency procedures must clearly indicate how the operator is to use environmentally qualified safety related display instrumentation to diagnose failure to perform safety functions and how the alternate equipment can be used to perform such safety functions.

Plant emergency procedures typically include provisions for the operator to sample or monitor radioactivity levels or combustible gas levels, to confirm that valves are in the correct position, to monitor flow or temperature, etc.

Some of these functions are essential for correct operator action, to mitigate accidents, and prevent radioactive releases.

When this is the case, the radiation

sensors, valve position indicators, pressure transmitters, thermo-

couples, etc. should be qualified to function in the relevant accident environment.

Licensees should, therefore, review their emergency procedures to determine the electrical components needed to perform the functions of Safety Related Display Information, Post Accident Sampling and Monitoring, and Radiation Monitoring.

When equipment implied by the emergency procedures is not listed, justification must be provided that failure of such equipment would not-prevent accident mitigation

~

or radioactivity release.

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

Environmental gualification of equipment to be installed for implementation of Regulatory Guide 1.97,

" Instrumentation for Post Accident Monitoring,". will be addressed at the time of implementation.

To support the. licensee's conclusions regarding the identification of equipment for which environmental qualification should be addressed, the plant emergency procedures for mitigation of LOCAs, MSLBs, or

'FWLBs should be submitted.

The submittals should include all secondary procedures referenced in '.the emergency procedures.

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

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

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

Whenever an item of safety-related equip-ment may be located in an environment outside the range of normal conditions, place it on the list of equipment in a potentially hostile environment.

Equipment subject to service conditions that are at no time more severe than the normal environment for which the equipment has been designed and under which its proper operation has been verified, can be regarded as qualified by experience and need not be considered further except with respect to the aging requirements of the qualification guidelines; how'ever, the i denti-fication of this equipment.and the justification for excluding it from further review shall be documented.

APPENDIS A

TYPICAL E UIPMENT FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR MSLB ACCIDENT Engineered Safeguar ds Actuation Reactor Protection Con ainment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power Emergency Core Cooling 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 Ccaqonent Cooling Service Mater Emergency Shutdown Post Accident Sampling and Monitoring Radiation Monitoring Safety Related Display Instrumenta.ion 3

These systems will differ for l%Rs and

FWRs, and for old"r.and newer plants.

In each case the system features which allow fo'ransfer 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.

Emergency shutdown systems include those systems used to bring the plant to a cold shutdown condition following accidents which do not result in a breach of.he reac.or 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, Nore specific identification of these types of equipment can be found in the plant emergency procedures.