ML19305B728
| ML19305B728 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point, Palisades File:Consumers Energy icon.png |
| Issue date: | 02/15/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-03-12, TASK-3-12, TASK-RR NUDOCS 8003200155 | |
| Download: ML19305B728 (40) | |
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NUCLEAR REGULATORY COMMISSION U
., E WASHINGTON, D. C. 20555
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February 15, 1980
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Docket No.
50-155/255 Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
Dear Mr. Hoffman:
RE: ELECTRICAL EQUIPMENT ENVIROMMENTAL QU/4.IFICATION PALISADES / BIG ROCK POINT PLANTS Reviews of the environmental qualification of safety-related electrical equipment have been initiated under Topic III-12 of the SEP program.
The reviews are being conducted in accordance with Enclosure 1,
" Guidelines for Evaluating Environmental Qualification of Class lE Electrical Equipment in Operating Reactors." The documents considered in the reviews include licensee submittals on SEP Topic III-12 and supporting material including equipment qualification records.
Preliminary results of our review indicate deficiencies affecting the scope of equipment addressed, the cefinition of its envircr. ment, and the completeness of the supporting documentaticn.
Accordingly, it is
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deemed necessary to increase the level of effor. applied to these reviews by the SEP licensees.
The SEP licensees are asked to perform the review of their safety-related equipment against the Guidelines of Enclosure 1.
This program redirection will accelerate the qualification review process by assuring that the site visit by the NRC staff, which heretofore has served largely to generate questions concerning oeficiences in documentation, can instead concentrate on (1) review of discrepancies previously identified by the licensee and (2) auditing the basis for accepting the qualification of other equiptent.
All SEP licensees are requested to attend a meeting at the NRC Headquarters in Bethesda, Maryland, at 10:00 a.m., on February 21, 1980 in Room P-llB of the Phillips Buliding.
The purpose of this meeting is to ensure full l
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Mr. David P. Hoffman understanding of the qualification review process. Necessary licensee submittals, site visits, and related schedules will be covered.
In order that you may begin your activity on this accelerated effort, we are providing Enclosure 2, " Guidelines for Identification of That Safety Equipment of SEP Operating Reactors for Which Environmental Qualification is to be Addressed."
Sincerely,
/
1.Yf g e Dennis L. Ziemann, Chief 4
Operating Reactors Branch #2 Division of Operating Reactors
Enclosures:
1.
Guidelines for Evaluating 2.
Guidelines for Identification a
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Mr. David P. Hoffman February 15, 1980 cc M. I. Miller, Esquire Charles Bechhoefer, Esq., Chaiman Isham, Lincoln & Beale Atomic Safety and Licensing Board Suite 4200 Panel One First National Plaza U. S. Nuclear Regulatory Commission Chicago, Illinois 60670 Washington, D. C.
20555 Mr. Paul A. Perry, Secretary Dr. George C. Anderson Consumers Power Company Department of Oceanography 212 West Michigan Avenue University of Washington Jackson, Michigan 49201 Seattle, Washington 98195 Judd L. Bacon, Esquire Dr. M. Stanley Livingston Consumers Power Company 1005 Calle Largo 212 West Michigan Avenue Santa Fe, New Mexico 87501 Jackson, Michigan 49201 Sheldon, Harmon, Roisman and Weiss Myron M. Cherry, Esquire 1725 I Street, N. W.
Suite 4501 Suite 506 One IBM Plaza Washington, D. C.
20006 Chicago, Illinois 60611 Mr. John O'Neill, II Kalamazoo Public Library Route 2, Box 44 315 South Rose Street Maple City, Michigan 49664 Kalamazoo, Michigan 49006 Herbert Grossman, Esq., Chaiman Hunton & Williams Atomic Safety and Licensing Board George C. Freeman, J r., Esquire U. S. Nuclear Regulatory Comission P. O. Box 1535 Washington, D. C.
20555 Richmond, Virginia 23212 Dr. Oscar H. Paris Peter W. Steketee, Esquire Atomic Safety and Licensing Board 505 Peoples Building U. S. Nuclear Regulatory Commission Grand Rapids, Michigan 49503 Washington, D. C.
20555 Charlevoix Public Library Mr. Frederick J. Shon 107 Clinton Street Atomic Safety and Licensing Board I
Charlevoix, Michigan 49720 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Ms. Mary P. Sinclair Great Lakes Energy Alliance Big Rock Point Nuclear Power Plant 5711 Summerset Drive ATTN: Mr. C. J. Hartman M'
and, Michigan 48640 Plant Superintendent Charlevoix, Michigan 49720 s
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-Mr. David P. Hoffman February 15, 1980 cc Chai rman County Board of Supervisors Charlevoix County Charlevoix, Michigan 49720 Office of the Governor (2)
Room 1 - Capitol Building Lansing, Michigan 48913 Director, Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Federal Activities Branch Region Y Office ATTN: EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604 Resident Inspector c/o U. S. NRC P. O. Box 87 South Haven, Michigan 49090 Palisades Plant ATTN: Mr. J. G. Lewis Plant Manager Covert, Michigaa 49043 Township Supervisor Covert Township Route 1, Box 10 Van Buren County, Michigan 49043 Chri sta-Maria Route 2, Box 108C
'Charlevoix, Michigan 49720 i
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ENCLOSURE 1 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 er 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 Sprays 4.2 Service Conditions for a PWR Main Stean Line Break (MSLB)
Inside Containment 1.
Temoerature 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 High Energy Line Break (HELB) 4.3.2 Areas Where Fluids are Recirculated From Inside Containment to Accomplish Long-Term Emergency Core Cooling Following a LOCA 1.
Tercerature, Pressure and Relative Humidity 2.
Radiation 3.
Submergence 4.
Chemical Sprays
4.3.3 Areas Normally Mai tained at Room Conditions
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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
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Test Sequence 4.
Test Specimen Aging 5.
Functional Tryting and Failure Criteria 6.
Installation Interfaces 5.3 Qualifica+ ion by a Combination of Methods (Test,- Evaluation, Analysis) 6.0 Margin 7.0 Aging 8.0 Documentation Appendix A - Typical Equipment / Functions Needed for Mitigation of a LOCA or MSLB Accident Appendix B - Guidelines for Evaluating Radiation Service Conditions Inside Containment for a LOCA and MSLB Accident Appendix C - Thermal and Radiation Aging Degradation of Selected 4
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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 is, sued 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 It 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 detemine whether it should be requalified or replaced with a component whose qualification has been adequately veri fied.
Tnese guicelines are intended to be used by tne 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.
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, Equipmmt 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.
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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 detemination will be made as part of that related generic review.
2.0 DISCUSSION IEEE Std. 323-1974I is the current industry standard for environmental qualification of safety-rehted 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 torth generic requirements for equiptent quali-fication but the 1Q74 standard includes specific requirements for aging, margins, and maintaining documentation records that were not included in the 1971 trial use standard.
Tne intent of this document is not to provide guidelines for implementing either version of IEEE Std. 323 fo operating reactors.
In fact most of the operating reactors are not committed to comply with any particular ind.:stry standard for electrical equipment qualification.
However, all of the operating reactors are required to comply with the General Design Criteria I
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 safets shall be designed to accomodate the affects of and to be compatible with the environmental conditions associated with normal operation, maintenance.
testing and postulated accidents, including loss-of-coolant accidents."
The intent of the' widelines is to provide a basis for judgements required to confirm that coerating r'eactors 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 cr be obtained from FSARs, Technical specifications, and emergency proceirez, 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).
Tne gu.ceiines in tnis document are ap;iitabic to all components necessary for operation of the systems listed in Appendix A includir; Dut not limited
- to valves, motors, cabies, connectors, relays, switches, transmitters and i
Valve-position indicators, p
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4.0 SERVICE CONDITIONS In order to detemine 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)
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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 - 3400F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and
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(2). PWR Ice Condenser Lower Compartments 3400F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, 2.
Radiation - When specifying radiation service conditions for equipment exposed to radiation during normal operating and accident conditions, the nomal 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 i
submerged in contaminated liquids must be evaluated on a case by case
- basis, Guidelines for these evaluations are not provided in this
- document,
+5-Gamma Radiation Doses - A total ganna dose radiation servi'ce condition 7
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 perfomed to detennine if the dose specified is acceptable.
Procedures for evaluating radiation service conditions in such cases are provided in Appendix B.
The procedures in Appendix B are based on the calculation for a typical PWR reported in Appendix I
D of NUREG-0588,
Gama 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 contai'nment, they are not directly applicable to BWRs and other containment types' However, doses for these other s
plant configurations may be evaluated using similar procedures with conservative dose assumptions and adjust.aent factors developed on a case by case basis..
Beta Radiation Doses - Beta radiation doses generally are less significant than gama radiation doses for equipment qualification, This is due to the low penetrating power cf beta particles in comparison to gamma rays of equivalent energy, Of the general classes of electrical equipment in a plant (e.g., cables, instrument transmitters, valve operators, containment penetrations), electrical cable is considered the most l NUREG-0588 Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment.
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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 cf 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 ty approximately a factor of ten within 30 mils of the surface of eiertrical cable insulation of unit density.
An additional 40 mils or 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 h?s 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 detennined by the sum of the gama and beta doses.
3.
Subnergence - 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.
7 4.
Containment Sprays - Equipment exposed to chemical sprays should be qualified for the most severe chemical environment {actdic or basic) which emsid exist.
Demineralized water sprays should not be exempt free: tansideration as a potentially adverse service condition.
4.2 Service Conditions for a PWR Main Steam Line Break (MSLB) Inside Containment Equipment required to function in a steam line break environment must be qualified for the high temperature and pressure that could result.
In some cases the environmental stress on exposed equipment may be higher than that resulting from a LOCA, in others it may be no more severe than for a LOCA due to the automatic operation of a containment spray system.
1.
Temperature 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 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 of this approach, i.e., use of tne "Best Estimate",
is pending the completion of Task Action Plan A-21, Main Steamline Breat Inside Containment.
Class IE equipment installed in plants without automatic spray systems or plants with gpray systems subject to disabling single failures or delayed initiation should be qualified for a MSLB accident environment determined by a plant specific analysis.
Acceptable methods I See NUREG 0458, Short Tern Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors, for a more cetailed discussion of the best estimate calculation.
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for-performing such an analysis for operating reac' ors are provided in Section 1.2 for Category II plants in NUREG-0588 Interim Staff Position on Environmental Qualification of Safety-Related Elitetrical Equipment.
2.
Radiation - Same as Section 4.1 above except that a conservative 6
gamma dose of 2 x 10 RADS is acceptable.
3.
Submergence - Same as Sect.on 4.1 above.
4.
Chemical Sprays - Same as Section 4.1 above.
4.3 Service Conditions Outside of Containment 4.3.1 Areas Sub_iect 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 requirec to mitigate the evert vcs also idcr.ti'ied.
Thir equipment should be qualified for the service conditions reviewed and approved in the H:;B Sa'ety Evaluation Report for each specific plant.
4.3.2 Areas Where Fluids are Retirculated fron Ins'.de Containment to Accomplish Lone-Term Core Coolino Followinc a LOCA 1.
Temperature and Relative Humid tr - One hundred percent relative humidity should be established as a serv ~ce condition in confined spaces.
The temperature and pressure as a function of time should be based on the
- 'u t aninae analysis reported in the FSAR.
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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.
Submeroence - Not applicable.
9 4.
_ Chemical Sprays - Not applicable.
4.3.3 Areas Normally Maintained at Room Conditions Class IE equipment located in these areas does rot experience sicnificant stress due to a change in service conditions during a design basis event.
Tnis 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 i
a eas during a design basis event are expected to be random except to tre exte.t that +Q' may be due t: agir.; or failures of air cordite-in or -
ventilation systems.
Therefore, no special consideration need be given to tne environmental qualification of Class IE equipment in these areas provided toe aging requirements discussed in Section 7.0 below are satisfied and the a-eas are maintained at roo. conditions by redundant air conditioning or ventilation systems served by the onsite emergency electrical power syste.
E:uipr.ent located ir. areas r.:t served by redunda".t systems powered from o site emproency sources should be qualified for the environmental extremes mi@ could result from a failure of the systems as detennined from a plant s:e:ific analysis.
5.C 3 *iIFICATION METHODS altdf$ L hN l
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5.1 Selection of Qualification Method The choice of qualification method employed for a particular application of kquipment is largely a matter of technical judgement based on such l
factors as: (1) the severily 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 vf 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, l
the qualification for severe temperature, pressure, and steam service conditions for Class It 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 l
- 5. 3 below).
Exceptions to these general guidelines must te justified on a case by case basis.
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5.2 Oualification by Type Testino 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 it g as
.t.? 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 l
- if specific analyses are provided to demonstrate that the. materials involved t 11 not experience significant accelerated thennal 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 also 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 susce::tibility 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 throagnout 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 Specir.er. Agin; - Tests which were successful using test specinens whicn hau r.ct bes
- resged may be considered acceptable prcvided the component does nct centain materials which are known to be susceptible IA 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 themal aging.
5_
Functional Testino 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 nomally 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 ecuipment 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.
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0 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 md steam.
The quali,fication 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.
Follov;inc 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 Qualification - 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 radiaticn qualif'ed by perfomirt a calc;1atir, of tha d:sc 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 i hour.
2.
Cherical Sprav Oaalification - Com::enents enclosed er.tirely in corrosion resistant cases (e.g., stainless steel) nay be shown to be quali#ied for a che-ic51 eV con.ent by ar a-elytis of the effects of the particular che-icals on tha par-i:J.ar encio-sure materials.
The effects of the:-ical sprays or the pressure integrity of any gaskets or seals present should be considered in the analysis.
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. 6.0 Marcin IEEE Std. 323-1974 & 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 comercial 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 conditio's to sssure adequate margins.
Th( factor applied n
to the time equipment is requited to remain functional is the most significant in terns of tne additional confidence in qualification that is achieved by adding margins to service conditions when establishing test environments.
For this reasor., 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 to be added to the service conditions wher, specifying test conditions.
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- 7. 0 Implicit in the staff positior. 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 demor.strated for all Class IE equipnelt is not sufficient to justify the expense for plants already constructec and operating. This position does not, however, exclude equipment using matetlals 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.
J.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 compittrce with a design specification should not be considered adequate.
APPENDIX A i
,T_YPICAL 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 Dnergency Core Coolingl 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)
Ver.tilation for Areas Containing Safety Equipment Component Cooling Service Water Emergency Shutdown 2 Post Accident Sampling and Monitoring 3 3
Radiation Monitoring 1
Safety Related Display Instru entation3 D "
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2-I These systems will differ for PWRs and BWRs, 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.
2 Emergency shutdown systems include those systems used to bring the plant to a cold shutdown condition following accidents which do not result in a breach of the reactor coolant pressure boundary together with a rapid depressurization of the reactor coolant system. Examples of such systems and equioment are the RHR system, PORVs, RCIC, pressurizer sprays, chemical and volums control system, and steam dump systems.
More specific identification of these types of egaipment can be found in the plant emergency procedures.
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APPENDIX B PROCEDURES FOR EVALUATING GAMMA RADIATION SERVICE CONDFTIONS Introduction and Discussion The adequacy of gama 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 its to identify those parameters whose effect on the total gama 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:
7 RADS (1) A ccnservative dose at the containment centerline of 2 x 10 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 hUREG-0588, " 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 6 ft3 contain-MWth pressurized water reactor housed in a 2.52 x 10 ment with an iodine scrubbing spray systen.
A similar calculation w iti out iodine scrubbing spreys i.:uiJ increase the dose to eqt.ipment 7
approximately 15L.
The conservati" dose of 2 x 10 RADS assumed D**]D T)
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t 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) Shield.'ng _ 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. Doses specified for equipment located in these areas must be evaluated on a case by case basis.
(5) Since the dose adjustment factors used in these procedures are based on a calculation for a typical pressurized water reactor with a dry type containment, they are not directly applicable to boiling water reactors or other containment types.
- However, doses for these other plant configurations mcy be evaluated using similar orocedures with conservative dose assun9tions 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.
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-The procedure for using the figuret 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 3
Containment volume - 2.5 x 106 ft Compartment Volume - 8,000 ft3 Thickness of compartment shield wall (concrete) - 24" i
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 l'
radiation service condition _ specification.
Step 1 l
Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and 6
2.5 x 10 ft3 containment volume and read a 30-day integrated dose of 7
1.5 x 10 RADS.
i i
Step 2 Enter Figure 2 at a dose of 1.5 x 107 RADS and 24" of concrete shielding i
for the compartment the equipnent 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 adced the dose from sources inside the compartment l,
(Step 3).
Steo 3 3
i.'
Enter Figure 3-at 8,000 ft and read a correction factor of 0.13.
The p
dose due to sources inside the compartment would then be 0.13 (1.5 x 10 )
7
= 1.95.x-106 RADS.
The sums of.the doses from steps 2 and 3 equals:
1 l
4.5 x 104 7
RA05 + 0.13 (1.5 x 10 ) RADS = 2.0 x 106 RADS
, Step 4 Enter Figure 4 at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and read a correction factor of 0.15.
Apply this factor to the sum of the doses detennined from steps 2 and 3 to correct the 30 day total dose to the equipment inside the compartment l
to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
x 10 ) = 3 x 105 6
RADS 0.15 (2.t 6
l In this particular example the service condition of 2 x 10 RADS specified is conservative with respect to the estimated dose of 3 x 5
10 RADS calculated in steps 1 through 4'and is, therefore, acceptable.
l l
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FIGURE 1 NOMOGRAM FOR CONTAINMENT VOLUME AND REACTOR POWER LOCA DOSE CORRECTIONS
- OONTAINMENT VOLUME (ft3) 3 x 106 2 x 106 30 DAY MWrH INTEGRATED 4000 yDOSE 1 x 106 7
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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 themal aging in a 45 C envircnment and nomal atmosphere for 10 or 40 years is indicated by an (*) in the appro-priate columr..
Significant aging degradation is defined as that amount of degradation that would place in substantial doubt the ability of ty.oical equipment using these materials to function in a hostile env ironmen 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 terms used to characterize the dose effect is as follows:
Threshold - Refers to damage threshold, which is the radiation e
exposure required to change at. cast one physical property of the material.
Percent Change of Property - Refers to the radiation exposure s
i requircJ +o change the physical property noted by the percent.
Allowable - Refers to the radiation which can be absorbed before e
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 i
Aeronautics anc Space Acrinistration's Scientific and Tetical Aerospace Report (STAR), NTIS Government Report Announcenents and Index (GRA), and 1
. various manufacturers data reports.
The materials list is not to be
- considered all intiusive 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.
7 The data base for thermal and radiation aging in engineering mat.erials is rapidly expanding at this time.
As additional information becomes available Table C ' will be updated accordingly.
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TilERMAL AND RADIATION AGING DEGRADATION OF, SEL.ECTED, MA,TE.RI,AL,5,
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ENCLOSURE 2 GUIDELINES FOR IDENTIFICAT!0N OF THAT SAFETY EQUIPMENT OF SEP OPERATING REACTORS FOR WHICH ENVIRONMENTAL QUALIFICATION 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 br 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 nutside containment, or FWLB inside or outside containment.
If a plant specific 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 equipment will probably be named in 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.
l (3)
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.
1 Equipment now indicated in emergency procedures in response to TMI-2 lessons learned should be listed.
Environmental Qualification of equipment to be installed for i
implementation of Regulatory Guide 1.97, " Instrumentation fcr Post Accident Monitoring,". will be addressed at the time of implementation.
(4)
To support the licensee's conclusions regarding the identification 4
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 attaining cold shutdown, radiation monitoring, sampling, etc. should be included.
(5)
The licensee should document anticipated service conditions in every portion of the plant where the environment could be influenced by the 1
accident or its conseq.ences, 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; however, the identi-fication of this equipment and the justification for excluding it from further review shall be documented.
APPENDIX A
, TYPICAL EQUIPMENT / FUNCTIONS NEEDED FOR
~
MITIGATION OF A LOCA OR_MSLB ACCIDENT Engineered Safeguards Actuation
[f Reactor Protection i
Containment Isolation
.6 L Steamline Isolation Main Feedwater Shutdown and Isolation E,mergency Power l
Energency Core Cooling Containment Heat Removal Containment Fission Product Renoval Containment Combustible Gas Contml Auxiliary Feedwater Containment Ventilation Containment Radiation Monitoring Control Doom Habitability Systems (e.g., HVAC, Radiation Filters)
Ventilation for Areas containing Safety Equipment Carmonent Cooling Service Water 2
E.mergency Shutdown 3
Post Accident Sampling and Monitoring 3
Radiation Monitoring 3
Safety Atlated Display Instrtsnentatien
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't hf h o Ju S
_ k inL:s ob 2-I These systems will differ for PWRs and FWRs, and for oldar and newer plants.
In each case the system features which allow fo 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 systees 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 depressuritation 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.
More specific identification of these types of equipment can be found in the plant energency procecures.
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