ML19254B982
| ML19254B982 | |
| Person / Time | |
|---|---|
| Issue date: | 09/23/1979 |
| From: | Butcher E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19254B981 | List: |
| References | |
| NUDOCS 7910090637 | |
| Download: ML19254B982 (29) | |
Text
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l Butcher / REDRAFT 9/23/79 GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS 1.0 Introduction 2.0 Identificati'on of Class IE Ecuipment 3.0 Service Conditions 3.1 Service Conditions Inside Containment for a Loss of Coolant Accident (LOCA) 1.
Temoerature and Pressure Steam Conditions 2.
Radiation 3.
Submergence 4.
Chemical Scrays 3.2 Service Conditions for a P'4R Main Steam Line Break (MSLB)
Inside Containment 1.
Temoerature and Pressure Steam Conditions 2.
Radiation 3.
Submergence 4.
Chemical Scrays 3.3 Service Conditions Outside Containment 3.3.1 Areas Subject to a Severe Environnent as a Result of a Hicn Energy Line Break (HELB) 3.3.2 Areas Where Fluids are Recirculated From Inside Containment to Accomolish Long-Term Emergency Core Cooling Following a LCCA 1.
Temcerature, Pressure and Relative Hunidi y 2.
Radiation 3.
Submergence 4
Chemical Scrays b
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4 3.3.3 Areas Maintained at Normal Room Conditions By Safety Related Air Conditioning Systems 4.0 Qualification Methods 4.1 Selection of Qualification Method 4.2 Qualification by Tyce Testing 1.
Sisulated Service Conditions and Test Duration 2.
Test Secuence 3.
Test Scecimen Aging 4.
Functional Testing and Failure Criteria 5.
Installation Interfaces 4.3 Oualification by a Combination of Methods (Test, Evaluation, Analysis) 5.0 Marcin 6.0 Aging 7.0 Documentation Appendix A - Typical Equipment / Functions Needed for Mitigation of a LOCA or MSLB Accident Appendix B - Guidelines for Evaluating Radiation Service Conditions Inside Containment for a LCCA and MSLB Accident Appendix C - Thermal and Radiation Aging Degradation of Selected Materials f
GUIDELINES FOR EVALUATING ENVIRONMENTAL OUALIFICATION OF CLASS IE ELECTRICAL EOUIPMENT 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 *w deter nine 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 licensee's reviews are now essentially complete and the NRC staff has begun to evaluate the results. This document sets forth a uniform set of generic guidelines for the NRC staff to use in its initial evaluation of the licensees' responses to IE Bulletin 79-01 and selected associated qualification documentation. The objective of the initial evaluation using these guidelines is to expeditiously identify Class IE equipment whose qualification is insignificant doubt. All such equipment identified will then be subjected to a plant application specific evaluation to determine whether it should be 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 SE? plants. The
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m 2-guidelines are basred on a value-impact assessment specifically for this class of plants. 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, based on a different value impact assessment.
The seismic qualification of Class IE equipment is being evaluated on a case-by-case basis by the Seismic Qualification Review Team (SORT) and is, therefore, outside the supe of this document.
2.0 IDENTIFICATION OF CLASS IE EQUIPMENT Class IE equipment includes all equipment needed to achieve emergency reactor shutdown, containment isolation, reactor core cooling, contain-ment and reactor heat removal, and prevention of significant release of radioactive material to the environment. Typical systems included in pressuri:ed and boiling water reactor designs to perform these functions for the most severe postulated loss of coolant accident (LOCA) and main steam?ine break accident (MSLB) are listed in Appendix A.
More detailed descriptions of the Class IE equipment installed at specific plants can be obtained frca FSARs, technical specifications, and emergency pr:cedures. Although variation in nonenclature may exist at the various plants, the environmental qualification of those systems wnich perform the functions identified in Accendix A should be evaluated against tne appropriate service conditions (Section 3.0).
6 216 i
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. 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.
3.0 SERVICE CONDITIONS '
In order to determine the adequacy of the qualification of equipment it is i
necessary to specify the environment the equipment is exposed to during nonnal and accident conditions. These environments are referred to as the
" service conditions."
The approved service conditions specified in the FSAR or other licensee sutmittals are acceptable, unless otherwise noted in the guidelines discussed below.
3.1 Service Conditions Inside Containment for a Loss of Coolant Accident (LOCA) 1.
Temeerature and Pressure Steam Conditions - The temperature and pressure steam conditions as a function of time should be based on the analyses in the FSAR.
2.
Radiation - When specifying radiation service conditions for equipment exposed to radiation during normal operation and accident conditions, the nornal operating dosa should be added to the dose received during the course of an accident. A total dose of 2 x 107 RADS is acceptable for Class IE equipment located inside containment.
In those cases where 7
the licensee has specified a dose less than 2 x 10 PADS it will be necessary for the NRC staff to perforn an application specific evaluation to deternine if the cose specified is acceptable. Guide-lines for evaluating radiation service conditions in such cases are provided in Appendix 3 6 217
. 3.
Submercence - The preferred method of protection against the effects of submergence 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 unless an application specific analysis has been perfomed to justify such an approach.
4 Chemical Sprays - Equipment exposed to chemical sprays should be qualified for the most severe chemical environment (acidic or basis) which could exist.
3.2 Service Conditions for a PWR Main Steam Line Break (MSLB)Inside Containment Equipment required to function in a steam line break environment must be qualified for the high temperature and pressure that could result.
In sane cases the environmental stress on exposed equipment may be higher than that resul*"
rm 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 - Equiptent qualified for a LOCA environment is considered qualified for a MSLB accident environ-ment in plants with automatic spray systems not subject to disabling f
single component failures. This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a themal analysis of typical components inside containment.3/
I See NUREG 0458, Short Tem Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more detailed discussien of the best estimate calculation.
6 218 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 MSL3 accident environment determined by a plant specific analysis. Acceptable methods for perforning such an analysis for operating reactors are provided in Section 1.2 for Category II plants in NUREG 0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
2.
Radiation - Same as Section 3.1 above except that an upper 6
bound dose of 2 x 10 RADS is acceptable.
3.
Submercence - Same as Section 3.1 above.
4.
Chemical Sorays - Same as Section 3.1 above.
3.3 Service Conditions Outside of Containment 3.3.1 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 HELS 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 equiprr.ent should be qualified for the service conditions reviewed and approved in the HELB Safety Evaluation Report for each specific plant.
3.3.2 Areas Where Fluids are Recirculated from Inside Containment to Accomolish Lonc-Term Core Cooling Following a LOCA 1.
Temeerature. Pressure and Relative Humidity - One hundred percent relative humidity should 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.
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Radiation - An upper bound dose of 4 x 10 RADS is acceptable except where noted differently in the FSAR.
Except for Figure 2 the dose reduction factors in Appendix B are not applicable.
3 3.
Submergence - Not applicable.
4.
Chemical Sorays'- Not applicable.
3.3.3 Areas Maintained at Normal Room Conditions by Safety Related Air Conditioning Systems 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, NEv.A, 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. 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 6.0 below are satisfied.
4.0 Qualification Methods i
4.1 Selection of Oualification 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 consecuences of design basis 6
220
. events, i.e., Class IE equiement (see Section 2.0 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 or evaluation. Exceptions to these general guidelines must be justified on a case by case basis.
4.2 Qualification by Tvoe Testing A simole vendor certification or statement by a licensee that an item of equipment was type tested does not in itself provide adequate assurance that the equipment is qualified to function in its design basis service conditions. The actual test plan and results must be reviewed to assure that the factors discussed below have been adequately accounted for.
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 3.0 above. The total 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 accident. A shorter test (uration may be considered acceptable 1
if an application specific analysis is provided to demonstrate that the specific materials involved will not experience significant accelerated themal aging during the pericd not testec.
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Test Sequence - The component being tested should be exposed to the high temperature, steam and pressure environment in the sequence defined for its service conditions. Where radiation is a service condition, 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 (see Section 6.0).
If the component contains any such materials, the radiation dose should be applied prior to or during exposure to the high temperature, steam and I
pressure environment. The same test specimen should be used throughout the test sequence for all service conditions (i.e.,
separate effects tests generally are not acceptable).
3.
Test Soecimen Aging - Tests which were successful using test specimens which had not been preaged may be considered valid provided the component does not contain any materials which are known to be susceptible to significant degradation due to themal and radiation aging (see Section 6.0).
If the component contains any such materials a qualified life for the component must be established on a case by case basis. Arrhenius techniques are generally considered acceptable for thennal aging.2.l 4.
Functional Testing and Failure Criteria - Operational modes tested should be representative of the actual application requirements (e.g., components which operate normally energi:ed in the plant should be nornally energi:ed during the tests).
If a comconent 2/- (Reference to a description of Ar-henuis techniques to be inserted here) 6 222
. fails at any time during the tr.st, even in a so called " fail safe" mode, the test should ce considered inconclusive with regard to demonstrating tne ability of the compencnt to function for the entire: period prior to the f ailues.
5.
Installation Interfaces - The ecuigent moonting and electrical or mechanical seals used during tra type test should be representative of the actual installition for the test to be considered car.ciusive.
The equipment qu&lification 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 comon problems such as procective enclosures installed upside down with drain holes at the top and penetrations in equipment housings for electrical connections being left unsealed or susceptit.le to moisture incursion tn;tugh stranded c'ondi ctors.
4.3 Cualification by a Comb' nation of MetMds (Test, Evaluation, Analysis As discussed in Seccion 4.1 above, an item of Class IE equipment may be shown to be qualified for a complete spectrum of service conditions even though it was cnly type tested for high temperature, pressore and steam. The qualification for service condition; such as radiation and chemical sprays may be demonstrated by analysis or evaluation..
In such cases the overall quiification is said to be ty 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.
6 22b 1.
Radiation Qualification - Some of the earlier type tests perfomed 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 time and location dependent calculation of the dose expected using the methods described in Appendix B and analyzing the effect of the calculated dose on the materials used in the equipment. As a general rule, the minimum dose that should be assumed is that which would result from at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of exposure.
2.
Chemical Spray Cualification - 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 materials. Any gaskets or seals present should be considered in the analysis.
5.0 Ma rgin IEEE Std. 323-1974 defines margin as the difference between the most severe specified service conditions of the plant and the conditions used in type testing to account for nomal variations in commercial production of equipment and reasonable errors in defining satisfactory perfomance.
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 6 224 the time requircd to remain functional when the guidelines for Functional Testing and railure Criteria in Section 4.2 above were established.
In addition, all of the guidelines in Section 3.0 for establishing service cor.diticas 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 3.0 and 4.2 are satisfied no separate margin factors are required to be added to the service conditions when specifying test conditions.
7.0 Ach Implicit in the staff position in Regulatory 1.89 with regard to backfitting IEEE Std. 323-1974 is the staff's cenclusion that the incr mental improvement in safety frcm arbitrarily reaniring that a e
specific qualified life be demonstrated for all Class IE equipment is not sufficient to justify the expense for plants already constructed and operating. This position does not, however, exclude equipment using materials known to exhibit significant degradation due to thermal and radiation aging from a requiretrent to have a maintenance or replace-ment schedule' based en the specific acing characteristics of the material. The staff position is that a qualified life should be demonstrated for all such equipment and that a formal program should l
exist to review surveillance testing and maintenance records at the plant to assure that equipment which is uhibiting age related degridation is identified. Appendix C is a listing of materials which may be found in nuclear power plants along with an indication of the material susceptibility to thermal and radiation aging.
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! l 8.0 Documentation
' Complete and auditable records must be available for qualification by any of the methods described in Section 4.0 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 unless the certification document identifies the basis for thecertification (i.e., a test report or analysis document).
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APPENDIX A
, TYPICAL EQUIPMENT / FUNCTIONS NEEDED FOR M_ITIGATION OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power l
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 Water 2
Emergency Shutdown l
3 Posc Accident Sampling and Monitoring 3
Radiation Monitoring 3
Safety Related Display Instrumentation b b!
I i
2-I These systems will differ for PWRs and BWRs, and for older and newer plants.
In each case the system features which allow for transfer to recirculation cooling made and establishment of long term cooling with baron precipitation control are to be considered as part of the system to be evaluated.
Emergency shutdewn 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, standby liquid control system, and steam dump systems.
1"More specific identification of these types of equipment can be found in the plant emergency procedures.
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APPENDIX B GUIDELINES FOR EVALUATING RADIATION SERVICE CONDITIONS Introduction and Discussion The adequacy of 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 whase effect on the total 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 guidelines and restrictions for their use are as follows:
7 (l) A conservative dose at the containment centerline of 2 x 10 RADS 6
for a LOCA and 4 x 10 RADS for a MSLB accident has been assumed.
This assumption and all the dose rates used in the procedure outlined below are based on the methods and sample calculation described in Appendix D of NUREG 0588, Interim Staff Position i
on Environmental Qualification of Safety-Related Electrical Equip-ment. The sample calculation in NUREG 0588 is for a 4,000 MWth 3
reactor in a 2.52 x 106 ft containment.
(2) Shielding calculations are based on an average gama energy of 1MEV derived from TID 148 4.
(3) These procedures are not applicable to equipment located directly above the containment sump or near filters. A ccnser-7 vative dose of 2 x 10 RADS may be assumed directly above the containment sump but the dose in the vicinity of filters must be calculated on a case by case basis.
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b (4) Only gama doses are considered. The beta energies involved are assumed to be too low to produce radiation damage of major consequence to the general classes of equipment of interest.
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.
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 x 106 fg3 3
Compartment Volume - 8000 ft Thick. ness 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 j
i service condition specification.
6 230 i
I Step 1 Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and 2 x 106 ft3 containment volume and read a dose of 1.5 x 107 RADS.
Steo 2 7
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 6 x 104 RADS.
This is the dose the equipment receives from sources outside the compa rtment. To this must be added the dose from sources inside the compartment (Step 3).
Step 3 3
Enter Figure 3 at 8,000 ft and read a correction factor of 0.13.
The 7
dose due to sources inside the compartment would then be 0.13 (2 x 10 )
6
= 2.6 x 10 RADS. The sum of the doses from steps 2 and 3 equals:
4 RADS + 0.13 (2x10 ) RADS = 2.66 x 106 7
RADS 6 x 10 Step 4 Enter Figure 4 at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and read a correction fa.or of 0.15.
Apply this factor to the sum of the doses determined from steps 2 and 3 to get the total dose to the equipment inside the compartment for I hour.
0.15 (2.66 x 10 ) = 4 x 103 RADS 6
6 In this particular example the service condition of 2 x 10 RADS specified 5
is conservative with respect to the estimated dose of 4 x 10 RADS calcu-lated in steps 1 through 4 and is, therefore, acceptable.
6 2M
FIGURE 1 NOMOGRAM FOR CONTAINMENT VOLUME AND REACTOR POWER DOSE CORRECTIONS CONTAINMENT VOLUME (ft3) 3 x 106 l
2 x 108 30 DAY INTEGRATED yDOSE 1 x 108 4000 3 x 107 -
3000 2000 5 x 105 1000 2 x 107 4 x 108 RAD 3 x 105 200 1 x 107 2 x 10s 1 x 105 5 x 108 4 x 108 3 x 108 2.5 x 106 2.0 x 108 1 x 106
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FIGURE 2 DOSE CORRECTION FACTOR FOR CONCRETE SHIELDlNG
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l APPENDIX C THERMAL AND RADIATION AGING DEGRADATION OF SELECTED MATERIALS Table B-1 is a partial list of materials which may be found in a nuclear power plant along with an indication of the material susceptibility to radiation and themal aging.
0 Susceptibility to significant themal aging in a 45 C environment and nomal 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 environmen t.
Susceptibility to radiation damage is indicated by the dose level and the observed effect identified in the column headed BASIS. The meaning of the tems used to characterize the dose effect is as follows:
e Threshold - Refers to damage threshold, which is the radiation exposure required to change at least one physical property of the material.
e Fercent Change of Property - Refers to the radiation exposure required to change the physical property noted by the percent.
e Allowable - Refers to the radiation which can be absorbed before serious degradation occurs.
The infomation in this appendix is based on a literature search of sources including the National Technical Infomation Service (NTIS), the National Aeronautics and Space Administration's Scientific and Technical Aerospace Report (STAR), NTIS Government Report Announcements and Index (GRA), and 6
236
. 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.
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