ML19256F655
| ML19256F655 | |
| Person / Time | |
|---|---|
| Issue date: | 11/13/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| TASK-A-24, TASK-OR IEB-79-01, IEB-79-1, NUDOCS 7912190733 | |
| Download: ML19256F655 (35) | |
Text
AR FOR nnow f[e k
UNITED STATES y
j s-n ' g NUCLEAR REGULATORY COMMISSION E
WASHINGTON D. C. 20555 E
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NOV 131979 MEMORAt!DUM FOR:
V. Stello, Director, Office of Inspection and Enforcement FROM:
H.R. Denton, Director, Office of Nuclear Reactor Regulation
SUBJECT:
GUIDELINES FOR EVALUATING QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS Enclosed is a copy of the subject guidelines. These guidelines were prepared by NRR, D0R to satisfy its comitment to IE to provide guidelines and criteria for IE to use in its reviews of the licensee responses to IE Bulletin 79-01.
As stated in Section 1.0, Introduction, the objective in preparing the guidelines was to set forth guidelines that should be used to identify Class IE equipment installed in operating reactors whose documentation does not provide reasonable assurance of environmental qualification.
Once IE has identified any such equipment it is anticipated that IE would transfer the lead responsibility for the final resolution to NRR, DOR. This is consistent with our plan as outlined at the July 11, 1979, Comission Briefing on IE Bulletin 79-01 and equipment oualification.
Your particular attention is directed to Appendix C, Themal and Radiation Aging Degradation of Selected Materials.
This appendix is provided to support implementation of the staff position stated in Section 7.0, Aging.
In summary, the staff position for existing equipment in operating reactors is that a specific qualified life based on thermal and radiation age degradation need only be established for equipment using materials known to exhibit significant degradation from these aging effects.
Appendix C is a partial listing of materials which may be found in nuclear power plants along with an indication of the material susceptibility to aging. This listing is based on input from only one of several DOR consultants.
Reports from the other D0R consultants are under review.
We expect to complete the review by the end of December 1979, and we will supplement Appendix C with additional informa ion at that time.
e.,
H.R. Denton, Director Office of Nuclear Reactor Regulation
Contact:
E. Butcher X-27900 1626 152 7 912100 733
Enclosure:
As stated cc w/ enclosure:
E. Jordan H. Denton F. Schroeder R. Mattson D. Eisenhut R. Vollmer J. Miller W. Gammill B. Grimes S. Hanauer G. Knighton G. Lainas D. Tondi R. Satterfield D. Crutchfield B. Snyder F. Rosa W. Russell T. Dunning W. Morris D. Mcdonald (IE Region IV)
S. Block T. Quay R. Feit A. Szukiewicz M. Chiramal J.E. Knight E. Butcher 1626 153
Butcher / REDRAFT 9/23/79 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 Ecuionent 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 Sorays 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 Sorays 4.3 Service Conditions Outside Containment 4.3.1 Areas Subiect to a Severe Environment as a Result of a High Enercy Line Break (HEL3) 4.3.2 Areas Where Fluids are Recirculated From Inside Containment to Accomolisn Long-Tern Emergency Core Coolino following a LOCA 1.
Tem::erature, Pressure and Relative Humidity 2.
Radiation 3.
Submergence 4
Chemical Sorays 1626 154
. 4.3.3 Areas Normally Mat-tained at Room Conditions 5.0 Qualification Methods 5.1 Selection of Qualification Method 5.2 Qualification by Tyoe Testing 1.
Simulated Service Conditions and Test Duration 2.
Test Scecimen 3.
Test Sequence 4.
Test Specimen Aaing 5.
Functional Testing and Failure Criteria 6.
Installation Interfaces 5.3 Oualification by a combination of Methods (Test, Evaluation, Analysis) 6.0 Margin 7.0 Aoina 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 Materials 1626 155
GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION 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 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 cualification documentation. The objective of the evaluations using these guidelines is to identify Class IE eauipment whose documentation does not provide reasonable assurance of environ-mental qualification. All such equipment identified will then be subjected to a plant acolicaticn scecific evaluation to determine whether it should be requalified or reolaced with a component whose cualification has been adecuately veri fi ed.
These guicelines are intended to be used by the NRC staff to evaluate the qua'ification re: nods used for existing equipment in a particular class of
- plants, i.e., :urrently operating reactors including SEP plants.
1626 156 Equipmant in other classes of plants not yet licensed to operate, or replacement equipment for operating reactors, may be subject to different requirements such as those set forth in NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
In addition to its reviews in connection with IE Bulletin 79-01 the staff is engaged in other generic reviews that include aspects of the equipment qualification issue. TMI-2 lessons learned and the effects of failures of non-Class IE control and indication equipment are examples of these generic reviews.
In some cases these guidelines may be applicable, however, this detemination will be made as part of that related generic review.
2.0 DISCUSSION l
IEEE Std. 323-1974 is the current industry standard for environmental qualification of safety-related electrical equipment. This standard was first issued as a trail use standard, IEEE Std. 323-1971, in 1971 and later after substantial revision, the current version was issued in 1974 Both versions of the standard set forth generic requirements for equipment quali-fication but the 1974 standard includes specific requirements for aging, margins, and maintaining documentation records that were not included in the 1971 trial use standard.
The intent of this document is not to provide guidelines for implementing either version of IEEE Std. 323 for operating reactors.
In fact most of the operating reactors are not committed to comply with any particular industry standard for electrical eauipment qualificatien. However, all of the operating reactors are required to comply with the General Design Criteria I IEEE Std. 323-1972, "!EEE Standard for Qualifying Class IE Equipment for Nuclear Power Generning Stations. "
1626 157 specified in Appendix A of 10 CFR 50.
General Design Criterien 4 states in part that " structures, systems and components important to safet; shall be designed to accomodate the affects of and to be compatible with the environmental conditions associated with nomal operation, maintenance, testing and postulated accidents, including loss-of-coolant cccidents."
The intent of these guidelines is to provide a basis for judgements required to confirm that operating r'eactors are in compliance with General Design Criterion 4.
3.0 IDENTIFICATION OF CLASS IE EOUIPMENT Class IE equipment includes all electrical equipment needed to achieve emergency reactor shutdown, containment isolation, reactor core cooling, containment and reactor heat removal, and prevention of significant release of radioactive material to the environment. Typical systems included in pressurized and boiling water reactor designs to perform these functions for the most severe postulated loss of coolant accident (LOCA) and main steamline break accident (MSLB) are listed in Appendix A.
More detailed descriptions of the Class IE equipment installed at specific plants can be obtained from FSARs, Technical specifications, and emergency procedures. Although variation in nomenclature may exist at the various plants, environmental qualification of those systems which perform the functions identified in Appendix A should be evaluated against the appropriate service conditions (Section 4.0).
The guidelines in this document are applicable to all components necessary for operation of the systems listed in Accendix A including but not limited to valves, motors, cables, connectors, relays, switches, transmitters and valve position indicators.
1626 158 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) 1 Temoerature 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 nigh 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 (21 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 radia*. ion 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 submerged in contaminated liquids must be evaluated on a case by case basis, Guidelines for these evaluations are not provided in this
- document, 1626 159
,5-Gamma Radiation Doses - A total gama dose radiation service 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 s
dose less than this value has been specified, an application specific evaluation must be perfomed to determine 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.
Gamma dose radiation service conditions for SWRs 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 wi.th a dry type containment, they are not directly applicable to BWRs and other containment types, However, doses /or these other plant configurations may be evaluated using similar procedures with conservative dose assumptions and adjustment factors develooed 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 of beta particles in comparison to gamma rays nf 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 NUREG-0588, Interim Staf# Position on Environmental Qualification of Safety-Related Electrical Equipment.
1626 160 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 c.alculation, the conservative 8
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 108 RADS and considering the shielding factors discussed here, that the beta dose to radiation sensitive equipment internals would be less than or equal to 10", of the total gamma dose to which an item of equipment has been qualified, then that equipment may be considered qualified for the total radiation environment (gamma plus beta).
If this criterion is not satisfied the radiation service condition should be determined by the sum of the garm:a and beta doses.
3.
Submergence - The preferred method of protection against the effects of submergency is to locate equipment above the water ficoding 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.
1626 161 4.
Containment Sorays - Equipment exposed to chemical sprays should be qualified for the most severe chemical environment (acidic 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 PWR Main Steam Line Break (MSLB) Inside Containment Equipment required to function in a steam line break environment must be qualified for the high temperature and pressure that could result.
In some cases the environmental stress on exposed equipment may be higher than that resulting from a LOCA, in others it may be no more severe than for a LOCA due to the automatic operation of a containment spray system.
1.
Temcerature and Pressure Steam Conditions - Equipment qualified for a LOCA environment is considered qualified for a MSLB accident en.riron-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.E The final acceptability of this approach, i.e., use of the "Best Estimate",
is pending the completion of Task Action Plan A-21, Main Steamline Break Inside Containment.
Class IE equipment installed in plants without automatic spray systems or plants with spray systems subject to disabling single failurer shouic be qualified for a MSL3 accident environment detemined by a p:3nt soecific analysis. Acceptable methods 1626 162 1 See NUREG 0a55, Short Tem Safety Assessirent on the Environmenta' Qualification of Safety Re:ated Electrical Equipmen of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation.
for performing such an analysis for operating reactors are provided in Section 1.2 for Category II plants in NUREG-0588, Interim Staff position on Environmental Qualification of Safety-Related Electrical Equipment.
2.
Radiation - Same as Section 4.1 above except that a conservative 6
gamma dose of 2 x 10 RADS is acceptable.
3.
Submercence - Same as Section 4.1 above.
4.
Chemical Sorays - Same as Section 4.1 above.
4.3 Ser/ ice Conditions Outside of Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a Hich Enercy Line Break (HELB)
Service conditions for areas outside containment exposed to a HELB were evaluated on a plant by plant basis as part of a program initiated by the staff in December,1972 to evaluate the effects of a HELB. The equipment required to mitigate the event was also identified. This equipment should be qualified for the service conditions reviewed and approvec 'n :ne 'i: B Sa'ety Evaluation Report for each specific plant.
4.3.2 Areas Where Fluids are Recirculated from Inside Containment to Accomolish Lonc-Ter n Core Cooline Followino a LOCA 1.
Te-cerature and Relative Humidity - One hundred cercent relative humidity should be established as a service condition in confined spaces. The tem:erature and pressure as a function of time should be based on the plant unicue analysis reported in the FSAR.
h $ ffg[
1626 163
9_
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 6 RADS would be basis.
In general, a dose of at least 4 x 10 expected.
3.
Submergence - Not applicable.
4.
Chemical Sorays - Not applicable.
4.3.3 Areas Normally Maintained at Room Conditions Class IE equiprent located in these areas does not experience 51gnificant stress due to a change in service conditions during a design basis event.
This equipment was designed and installed using standard engineering practices and industry codes and standards (e.g., ANSI, NEMA, National
. Electric Code). Based on these factors, failures of equipment in these areas during a design basis event are expected to be random except to the extent that they may be due to aging or failures of air conditioning or ventilation systems. Therefore, no special consideration need be given to the environmental qualification of Class IE equipment in these areas provided the aging recuirements discussed in Section 7.0 below are satisfied and the areas are maintained at room conditions by redundant air conditioning or ventilation systems served by the onsite emergency electrical pcwer system.
Eauipment locatec in areas not served by redundant systems powered from onsite emercency sources should be qualified for the envirenrental extremes which could result from a failure of the systems as determined from a plant specific analysis.
5.0 OUAu!FICATION METHCCS 1626 164
. 5.1 Selection of Qualification Method The choice of qualification method employed for a particular application of equipment is largely a matter of technical judgement based on such factors as: (1) the severity of the service conditions; (2) the structural and material complexity of the equipment; and (3) the degree of certainty required in the qualification procedure (i.e., the safety importance of the equipment function).
Based on these considerations, type testing is the preferred method of qualification for electrical equipment located inside containment required to mitigate the consequences of design basis events, i.e., Class IE equipment (see Section 3.0 above). As a minimum, the qualification for severe temperature, pressure, and steam service conditions for Class IE equipment should be based on type testing.
Qualification for other service conditions such as radiation and chemical sprays may be by analysis (evaluation) supported by test data (see Section 5.3 below). Exceptions to these general guidelines must be justified on a case by case basis.
5.2 Oualification by Tyce 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 long as the period from the initiation of the accident until the temoerature and pressure service conditions return to essentially the same levels that existed before the postulated accident.
A shorter test duration may te acceptable 1626 165
. if specific analyses are provided to demonstrate that the materials involved v 11 not experience significant accelerated themal aging during the period not tested.
2.
Test Soecimen - 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 Secuence - The component being tested should be exposed to a steam / air environment at elevated temcerature, Jnd pressure in the sequence defined for its service conditions. Where radiation is a service condition which is to be considered as part of a type test, it may be applied at any time during the test sequence provided the component does not contain any materials which are known to be susceptible to significant radiation damage at the service condition levels or materials whose susceptibility to radiation damage is not known (see Appendix C).
If the component contains any such materials, the radiation dose should be applied prior to or concurrent with exposure to the elevated temperature and pressure steam / air environnent. The same test specimen snould be used througnout the test secuence for all service concitions the ecui ment is to be cualified for by type testing. The type test should only be considered valid for the service conditicns applied to the same test specimen in the appropriate secuence.
4 Test Sceciren Acing - Tests which were successful using test specimens wnicn hac not been creaged may be considered accectable Orovided.ne
- cenant dces not contain materials wnich are known to be susce;;ib e 1626 166 to significant degradation due to thermal and radiation agir (see Section
- 7. 0 ).
If the component contains such materials a qualified life for the component must be established on a case by case basis. Arrhenius techniques are generally considered acceptable for thermal aging.
5.
Functional Testina 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 nonnally energized during the tests, motor and electrical cable loading during the test should be representative of actual operating conditions).
Failure criteria should include instrumer,t accuracy requirements based on the maximum error assumed in the plant safety analyses.
If a component fails at any time during the test, aven 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 peried prior to the failure.
6.
Installation Interfaces - The ocuipment mounting and electrical or mecnanical seals used during the type test should be representative of the actual installation for the t? t 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 oroblems sucn as protective enclosures installed upside down with drain holes at the top and penetrations in equipment housings for electrical c:nnections being left unsealed or susceptible to moisture incursion through stranded conductors.
1626 167 5.3 Qualification by a Combination of Methods (Test, Evaluation, Analysis As discussed in Section 5.1 above, an item of Class IE equipment may be shown to be qualified for a complete spectrum of service conditions even though it was only type tested for high temperature, pressure and steam. The 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.
Following are two specific examples of procedures that are consicered acceptable. Other similar procedures may also be revie.4ed anc founc acceptable on a case by case basis.
1.
Radiation Qualification - Some of the earlier tvoa tasts 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 perfoming a calculation of the dose expected, taking into account the time the equipment is required to remain functional and its location using the methods described in Appendix B, and analyzing the effect of the calculated dose on the materials used in the equipment (see Aopendix C). As a general rule, the time requi.ed to remain functional assumed for dose calculations should be at least I hour.
2.
Chemical Soray Oualification - 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 tne carticular cnemicals cn the particular enclo-sure materiais. The ef#ects of chemical sprays on the cressure integrity of any gaskets or seals present should be considerec in the analysis.
1626 168 6.0 Margin 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 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 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 speciried 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 when specifying test conditions.
ag_i.ng i
- 7. 0 Implicit in the staff position in Regulatory Guide 1.89 with regard to backfitting IEEE Std. 3?3-1974 is the staff's conclusion that the incremental improvement in safety from arbitrarily requiring that a specific qualified life be demonstrated for all Class IE ecuipment is not sufficient to justify the expense for plants alread/ constructed ar.d coerating. This position does not, however, exclude equipment 1626 i?9 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 1',dication of the material susceptability to themal and radiation aging.
8.0 Docu entation Complete and auditable records must be available for qualification by any of the methods described in Section 5.0 above to be considered valid.
These records should describe the qualification method in sufficient detail to verify that all of the guidelines have been satisfied. A simple vendor certification of compliance with a design specification should not be considered adequate.
1626 170
APPENDIX A
, TYPICAL EQUIPMENT / FUNCTIONS NEEDED FOR MITIGATION OF A LOCA OR MSLB ACCIDENT Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power 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 3
Post Accident Sampling and Monitoring 3
Radiation Monitoring 3
Safety Related Display Insterentation 1626 171 I These systems will differ for PWRs and CWRs, and for older 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. 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 emergency procedures.
1626 172
APPENDIX B PROCEDURES FOR EVALUATING GAMMA RADIATION SERVICE CONDITIONS Introduction and Discussion The adequacy of gamma radiation service conditions specified for inside containment during a LOCA or MSLB accident can be verified by assuning a conservative dose at the containment centerline and adjusting the dose according the plant specific parameters, The purpose of this appendix is to identify those parameters whose effect on the total gamma dose is easy to quantify with a high degree of confidence and describe procedures which may be used to take these effects into consideration.
The bases for the procedures and restrictions for their use are as follows:
7 (1) A conservative dose at the containment centerline of 2 x 10 RADS for a LOCA and 2 x 106 RADS for a MSL3 accident has been assumed.
This assumotion and all the dose rates used in the procedure out-lined below are based on the methods and sample calculation described in Appendix 0 of NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equip-ment."
Therefore, all the limitations listed in Acpendix 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 MWth pressurized water reactor housed in a 2.52 x 10 contain-ment with an iodine scrubbing spray system. A similar calculation without iodine scrubbing sprays would increase the dose to equipment 7
aporoximately 15".
The conservative dose of 2 x 10 RADS assumeo 1626 173 in the procedure below includes sufficient conservatism to account for this factor. Therefore, the pro wdure is also applicable to plants without an iodine scrubbing spray system.
(3) Shielding calculations are based on an average gama energy of 1 MEV derived from TID M844 (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 may be evaluated using similar procedures with conservative dose assumptions and adjustment factors developed on a case by case basis.
Procedure Figures 1 through 4 provide factors to be applied to the conservative dose to correct the dose for the following plant specific paramete s:
(1) reactor power level; (2) containment volume; (3) shielding; (4) compartment volune; and (5) time equipment is required to remain functional.
1626 174 The procedure for using the figures is best illustrated by an example.
Consider the following case.
The radiation service condition for a 6
particular item of equipment has been specified as 2 x 10 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" 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.
Steo 1 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.
Step 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 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 compartment (Step 3).
Steo 3 3
Enter Figure 3 at 3,000 ft and read a correction factor of 0.13.
The 7
dose due to sources inside the compartment would then be 0.13 (1.5 x 13 )
6
= 1.95 x 10 RADS. The sums of the doses from steps 2 and 3 equals:
4.5 x 10 RADS - 0.13 (1.5 x 10 ) RADS = 2.0 x 106 RADS 4
7 1626 175 Steo 4 Enter Figure 4 at I hour and read a correction factor of 0.15.
Apply this factor to the sum of the doses determined from steps 2 and 3 to correct the 30 day total dose to the equipment inside the compartment to 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 10 ) = 3 x 105 6
PADS 6
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 a and is, therefore, acceptable.
1626 176
FIGURE 1 NOMOGRAM FOR CONTAINMENT VOLUME AND REACTOR POWER LOCA DOSE CORRECTIONS
- CONTAINMENT VOLUME (ft3) 3 x 106 2 x 108 30 DAY M wTH INTEGRATED 4000 yDOSE 1 x 108 7
3000 4 x 10 2000 3 x 107 1000 5 x 108 4 x 105 7
2 x 10 3 x 106 200 2 x 105 1 x 107 1 x 105 5 x 108 4 x 106 3 x 106 2.5 x 106 2.0 x 106 1626 177 "iSL3 ACCIDENT DOSES SHOULD BE READ AS A FACTCR OF 10 LESS
FIGURE 2 DOSE CORRECTION FACTOR FOR CONCRETE SHIELDlNG
( y ONLY) 10s _
1 x 107 O
E S
$ 1 x 106 -
m Ib 3
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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 themal aging.
0 Susceptibility to significant themal 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 envi ronmen 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 characterire the dose effect is as follows:
s 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.
e Allowable - Refers to tne 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 Recor (STAR), NTIS Government Recort Announcements and Index (GRA), and 1626 181 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 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 infomation becomes available Table C-1 will be updated accordingly.
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