Regulatory Guide 1.89
| ML031480399 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1984 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| References | |
| EE 042-2 RG-1.089, Rev. 1 | |
| Download: ML031480399 (18) | |
Revision 1*
U.S. NUCLEAR REGULATORY COMMISSION June 1984 REGULATORY GUIDE
7 ,OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.89 (Task EE 042-2)
ENVIRONMENTAL QUALIFICATION OF CERTAIN ELECTRIC EQUIPMENT IMPORTANT TO
SAFETY FOR NUCLEAR POWER PLANTS
A. INTRODUCTION
Section 50.49 does not include requirements for seismic and dynamic qualification, protection of electric The Commission's regulations in 10 CFR Part 50, equipment against other natural phenomena and external
"Domestic Licensing of Production and Utilization events, and equipment located in a mild environment.
Facilities," require that structures, systems, and com- ponents important to safety in a nuclear power plant This regulatory guide describes a method acceptable be designed to accommodate the effects of environ- to the NRC staff for complying with § 50.49 of mental conditions (ie., remain functional under postu- 10 CFR Part 50 with regard to qualification of electric lated accident conditions) and that design control equipment important to safety for service in nuclear measures such as testing be used to check the adequacy power plants to ensure that the equipment can perform of design. These general requirements are contained in its safety function during and after a design basis General Design Criteria 1, 2, 4, and 23 of Appendix accident.
A, "General Design Criteria for Nuclear Power Plants,"
to Part 50; in Criterion I, "Design Control," Criterion The Advisory Committee on Reactor Safeguards has XI, "Test Control," and Criterion XVII, "Quality been consulted concerning this guide and has con- Assurance Records," of Appendix B, "Quality Assurance curred in the regulatory position. Any guidance in Criteria for Nuclear Power Plants and Fuel Reprocessing this document related to information collection activities Plants," to Part 50; and in § 50.55a. has been cleared under OMB Clearance No. 3150-0011.
- Specific requirements pertaining to qualification of
B. DISCUSSION
certain electric equipment important to safety are contained in § 50.49, "Environmental Qualification of IEEE Std 323-1974, "IEEE Standard for Qualifying Electric Equipment Important to Safety for Nuclear Class IE Equipment for Nuclear Power Generating Power Plants," of 10 CFR Part 50. Section 50.49 Stations,"' published February 28, 1974, was prepared requires that three categories of electric equipment by Subcommittee 2, Equipment Qualification, of the important to safety be qualified for their application Nuclear Power Engineering Committee of the Institute and specified performance and provides requirements of Electrical and Electronics Engineers (IEEE) and was for establishing environmental qualification methods approved by the IEEE Standards Board on Decem- and qualification parameters. These three categories are ber 13, 1973. The standard describes basic procedures
(1) safety-related electric equipment (Class E), (2) for qualifying Class 1E equipment and interfaces that non-safety-related electric equipment (non-Class I E) are to be used in nuclear power plants, including com- whose failure under postulated environmental conditions ponents or equipment of any interface whose failure could prevent satisfactory accomplishment of safety could adversely affect any Class IE equipment.
functions by safety-related equipment, and (3) certain postaccident monitoring equipment. This regulatory For the purposes of this guide, "qualification" is a guide applies only to these three categories of electric verification of design limited to demonstrating that the equipment important to safety. electric equipment is capable of performing its safety iCopies may be obtained from the Institute of Electrical and
'The substantial number of changes in this revision has made Electronics Engineers, Inc., 345 East 47th Street, New York, it impractical to indicate the changes with lines in the margin. New York 10017.
USNRC REGULATORY GUIDES Comments should be sent to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Regulatory Guides are issued to describe and make available to the Attention: Docketing and Service Branch.
public methods acceptable to the NRC staff of Implementing specific parts of the Commission's regulations, to delineate tech- The guides are Issued in the following ten broad divislons:
niques used by the staff In evaluating specific problems or postu- lated accidents or to provide guidance to applicants. Regulatory 1. Power Reactors 6. Products Guides are nor substitutes for regulations, and compliance with 2. Research and Test Reactors 7. Transportation them is not required. Methods and solutions different from those set 3. Fuels and Materials Facilities 8. Occupational Health out In the guides will be acceptable If they provide a basis for the 4. Environmental and Siting 9. Antitrust and Financial Review findings requisite to the Issuance or continuance of a permit or 5. Materials and Plant Protection 10. General license by the Commission.
Copies of Issued guides may be purchased at thecurrent Government This guide was Issued after consideration of comments received from Printing Office price. A subscription service for future guides In spe- the public. Comments and suggestions for Improvements In these cific divisions Is available through the Government Printing Office.
guides are encouraged at all times, and guides will be revised, as Information on the subscription service and current GPO prices may appropriate, to accommodate comments and to reflect new Informa- be obtained by writing the U.S. Nuclear Regulatory Commission, tion or experience. Washington, D.C. 20555. Attention: Publications Sales Manager.
function under significant environmental stresses resulting Electric equipment to be qualified in a nuclear from design basis accidents in order to avoid common- radiation environment should be exposed to radiation
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cause failures. Paragraph 50.49(e)(5) calls for equipment that simulates the calculated integrated dose (normal qualified by test to be preconditioned by natural or and accident) that the equipment must withstand prior artificial (accelerated) aging to its end-of-installed-life to completion of its intended safety function. Regulatory condition and further specifies that consideration must Position C.2.c proposes the use of source terms that are be given to all significant types of degradation that can consistent with previous guidance in the original edition have an effect on the functional capability of the of this guide, NUREG-0588, "Interim Staff Position on equipment. There are considerable uncertainties regarding Environmental Qualification of Safety-Related Electrical the processes and environmental factors that could result Equipment,2 and the DOR Guidelines, "Guidelines for in such degradation. Oxygen diffusion, humidity, and Evaluating Environmental Qualification of Class IE
accumulation of deposits are examples of such effects. Electrical Equipment in Operating Reactors." 3 Because of these uncertainties, state-of-the-art precondi- tioning techniques are not capable of simulating all Item (8) of Regulatory Position C.2.c addresses significant types of degradation, and natural pre-aging is qualification of equipment exposed to low-level radiation difficult and costly. As the state of the art advances doses. Numerous studies that have compiled radiation and uncertainties are resolved, preconditioning techniques effects data on all classes of organic compounds show may become more effective. Experience suggests that that compounds with the least radiation resistance have consideration should be given, for example, to a combi- damage thresholds greater than 104 rads and would nation of (1) preconditioning of test samples employing remain functional with exposures somewhat above the the Arrhenius theory and (2) surveillance, testing, and threshold value. Thus, for organic materials, radiation maintenance of selected equipment specifically directed qualification may be readily justified by existing test toward detecting those degradation processes that, based data or operating experience for radiation exposures on experience, are not amenable to preconditioning and below 104 rads. However, for electronic components, that could result in common-cause functional failure of studies have shown failures in metal oxide semiconductor the equipment during design basis accidents. devices at somewhat lower doses. Therefore, radia- tion qualification for electronic components may have a It is essential that safety-related electric equipment be lower exposure threshold.
qualified to demonstrate that it can perform its safety function under the environmental service conditions in The regulatory positions delineated in this guide
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which it will be required to function and for the length of reflect the state of the art. Research programs currently time its function is required and that non-safety-related in progress are investigating such concerns as the effects electric equipment covered by paragraph 50.49(b)(2) of oxygen in a LOCA environment, the validity of be able to withstand environmental stresses caused sequential versus simultaneous applications of steam and by design basis accidents 'under which its failure could radiation environments, and fission product releases prevent the satisfactory accomplishment of safety func- following accidents. The staff recognizes that the results tions by safety-related equipment. This concept applies of research programs may lead to revisions of the throughout this guide. The specific environment for regulatory positions.
which individual electric equipment must be qualified will depend on the installed location and the conditions
C. REGULATORY POSITION
under which it is required to perform its safety function.
The procedures described by IEEE Std 323-1974, The following are examples of considerations to be "IEEE Standard for Qualifying Class IE Equipment for taken into account when determining the environment Nuclear Power Generating Stations,"1 are acceptable to for which the equipment is to be qualified: (1) equip- the NRC staff for satisfying the Commission's regulations ment outside containment would generally see a less pertaining to the qualification of electric equipment for severe environment than equipment inside containment; service in nuclear power plants to ensure that the
(2) equipment whose location is shielded from a radia- equipment can perform its safety functions subject to tion source would generally receive a smaller radia- the following:
tion dose than equipment at the same distance from the source but exposed to its direct radiation; (3) equip- 1. Section 50.49, "Environmental Qualification of ment required to initiate protective action would generally Electric Equipment Important to Safety for Nuclear be required for a shorter period of time than instrumen- tation required to follow the course of an accident; and
(4) analyses taking into account arrangements of equip- ment and radiation sources may be necessary to deter-
2 mine whether equipment needed for mitigation of design Copies may be obtained from the NRCIGPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555.
basis accidents other than loss-of-coolant accidents .I
3 (LOCA) or high-energy line breaks (HELB) could be Available for inspection or copying at the U.S Nuclear exposed to a more severe environment than the LOCA Regulatory Commission Public Document Room, 1717 H Street NW., Washington, D.C., as Enclosure 4 to IE Buiietin No. 79-OIB,
or HELB environments delineated in this guide. January 14, 1980.
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Power Plants," of 10 CFR Part 50 requires that safety- b. Effects of Sprays and Chemicals. The effects of related electric equipment (Class 1E) as defined in containment spray system operation should be considered.
paragraph 50.49(b)(1) be qualified to perform its intended This consideration should include, as appropriate, the safety functions. Typical safety-related equipment and effects of demineralized water spray or chemical spray systems are listed in Appendix A to this guide. Paragraph systems.
50.49(b)(2) requires that non-safety-related electric equip- ment be environmentally qualified if its failure under c. Radiation Conditions Inside and Outside Contain- postulated environmental conditions could prevent satis- ment. The - radiation environment for qualification of
,factory accomplishment of the safety functions by electric equipment should be based on the radiation safety-related equipment. Typical examples of non-safety- environment normally expected over the installed life of related electric equipment are included in Appendix B the equipment plus that associated with the niost severe to this guide. Paragraph 50.49(b)(3) requires that certain design basis accident during or following which the postaccident monitoring equipment also be environmen- equipment must remain functionaL The accident-related tally qualified. These are specified as "Categories 1 environmental conditions should be assumed to occur at and 2" in Revision 2 of Regulatory Guide 1.97, "Instru- the end of the installed life of the equipment. -Methods mentation for Light-Water-Cooled Nuclear Power Plants acceptable to the NRC staff for establishing radiation to Assess Plant and Environs Conditions During and doses for the qualification of equipment for BWRs and Following an Accident." PWRs are provided in Appendix D and the following:
(1) The source term to be used in determining
2. Paragraph 50.49(d) and Section 6.2 of IEEE Std the radiation environment associated with a design basis
323-1974 require equipment specifications to include LOCA should be taken as an instantaneous release to performance and environmental conditions. For the the containment of 100% of the noble gas activity, 50%
requirements called for in item (7) of Section 6.2 of of the halogen activity, and 1% of the remaining fission IEEE 323-1974 and paragraph 50.49(d)(3), the following product activity. The fission product solids should be should be included: assumed to remain in the primary coolant and to be carried by the coolant to th' containment sump(s).
a. Temperature and Pressure Conditions Inside Containment for LOCA and Main Steam Line Break (2) For all other design basis accidents (e.g.,
(MSLB). The following methods are acceptable to the non-LOCA high-energy line breaks or rod ejection or NRC staff for calculating and establishing the contain- rod drop accidents), the qualification source term ment pressure and temperature envelopes to which calculations should use the percentage of fuel damage equipment should be qualified: assumed in the plant-specific analysis (provided in the Final Safety Analysis Report (FSAR)). The nuclide
(1) Methods for calculating mass and energy inventory of the breached fuel elements should be release rates for LOCAs and MSLBs are referenced in calculated at the end of core life assuming continuous Appendix C to this guide. The calculations should full-power operation. The inventory of the fuel rod gap account for the time dependence and spatial distribution should be assumed to be 10% of the total rod activity of these variables. For example, superheated steam inventory of iodine and 10% of the total activity inven- followed by saturated steam may be a limiting condition tory of noble gases (except for krypton-85, for which a and should be considered. release of 30% should be assumed). All the gaseous constituents in the gaps of the breached fuel rods
(2) For pressurized water reactors (PWRs) with should be assumed to be instantaneously released to the a dry containment, calculate LOCA or MSLB contain- primary system. When substantial fuel damage is postu- ment environment using CONTEMPT-LT or equivalent lated, 100% of the noble gases, 50% of the halogens, industry codes. and 1% of the remaining fission product solids in the affected fuel rods should be assumed to be instantane-
(3) For PWRs with an ice condenser contain- ously released to the primary system.
ment, calculate LOCA or MSLB containment environ- ment using LOTIC or equivalent industry codes. (3) For a limited number of accident-monitoring instrumentation channels with instrument ranges that
(4) For boiling water reactors (BWRs) with a extend well beyond the values the selected variables can Mark I, II, or III containment, calculate LOCA or attain under limiting conditions as specified in Regulatory MSLB environment using CONTEMPT-LT or equivalent Guide 1.97, Revision 2, the environmental qualification industry codes. should be consistent with Regulatory Positions C.1.3.1.a and C.1.3.2.a of Regulatory Guide 1.97, Revision 2.
Since the test profiles included in Appendix A to IEEE Std 323-1974 are only representative, they should (4) The calculation of the radiation'environment not be considered an acceptable altemative to using associated with design basis accidents should take into plant-specific containment temperature and pressure account the time-dependent transport of released fission design profiles unless plant-specific analysis is provided products within various regions of the containment and to verify the applicability of those profiles. auxiliary structures.
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(5) Electric equipment that could be exposed a. Electric equipment that could be submerged to radiation should be environmentally qualified to a should be identified and qualified by testing in a sub- radiation dose that simulates the calculated radiation merged condition to demonstrate operability for the environment (normal and accident) that the equipment should withstand prior to completion of its required safety functions. Such qualification should consider that duration required. Analytical extrapolation of results for test periods shorter than the required duration should be justified.
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equipment damage is a function of total integrated dose and can be influenced by dose rate, energy spectrum, b; Electric equipment located in an area where and particle type. The radiation qualification should rapid pressure changes are postulated simultaneously factor in doses from all potential radiation sources at with the most adverse relative humidity should be the equipment location. Plant-specific analysis should be qualified to demonstrate that the equipment seals and used to justify any reductions in dse or dose rate vapor barriers will prevent moisture from penetrating resulting from component location or shielding. The into the equipment to the degree necessary to maintain qualification environment at the equipment location equipment functionability.
should be established using an analysis similar in nature and scope to that included in Appendix D to this guide c. The parameters to which electric equipment is and incorporating appropriate factors pertinent to the being qualified (e.g., temperature, pressure, radiation) by actual plant design (e.g., reactor type, containment exposure to a simulated environment in a test chamber design). should be measured sufficiently close to the equipment to ensure that actual test conditions accurately represent
(6) Shielded components need be qualified only the environment characterized by the test.
to the gamma radiation environment provided it can be demonstrated that the sensitive portions of the compo- d. Performance characteristics that demonstrate the nent or equipment are not exposed to significant beta operability of equipment should be verified before, radiation dose rates or that the effects of beta radiation, after, and periodically during testing throughout its including heating and secondary radiation, have no range of required operability. Variables indicative of deleterious effects on component performance. If, after momentary failure that prevent the equipment from considering the appropriate shielding factors, the total performing its safety function, e.g., momentary opening beta radiation dose contribution to the equipmerit or of a relay contact, should be monitored continuously to component is calculated to be less than 10% of the ensure that momentary failures (if any) have been total gamma radiation dose to which the equipment or accounted for during testing. For long-term testing, component has been qualified, the equipment or compo- however, monitoring during periodic intervals may be nent is considered qualified for the beta and gamma used if justified. l radiation environment.
e. Chemical spray or demineralized water spray
(7) Electric equipment located outside contain- that is representative of service conditions should be ment that is exposed to the radiation from a recirculat- incorporated during simulated event testing at pressure ing fluid should be qualified to withstand the radiation and temperature conditions that would occur when the penetrating the containment plus the radiation from the spray systems actuate.
recirculating fluid.
f. Cobalt-60 or cesium-137 would be acceptable
(8) Electric equipment that may be exposed to gamma radiation sources for environmental qualification.
low-level radiation doses should not generally be consid- ered exempt from radiation qualification testing. Excep- 4. The suggested values in Section 6.3.1.5, "Margin,"
tions may be based on qualification by analysis supported of IEEE Std 323-1974, except time margins, are accept- by test data or operating experience that verifies that able for meeting the requirements of paragraph 5Q49(e)(8).
the dose and dose rates will not degrade the operability Altematively, quantified margins should be applied to of the equipment below acceptable values. the environmental parameters discussed in Regulatory Position C.2 to ensure that the postulated accident d. Environmental Conditions for Equipnient Outside conditions have been enveloped during testing. These Containment. Electric equipment that is subjected to the margins should be applied in addition to any conserva- effects of pipe breaks and is required to mitigate the tism applied during the derivation of local environmental consequences of the breaks or to bring the plant to conditions of the equipment unless these conservatisms safe shutdown should be qualified for the expected can be quantified and shown to contain appropriate environmental conditions. The techniques to calculate the margins. The margins should account for variations in environmental conditions should employ a plant-specific commercial production of the equipment and the inac- modeL curacies in the test equipment.
3. Section 6.3, "Type Test Procedures," of IEEE Std Some electric equipment may be required by the
323-1974 should be supplemented with the following: design to perform its safety function only within the
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first ten hours of the event. This equipment should a. The item of equipment to be replaced is a remain functional in the accident environment for a component of equipment that is routinely replaced as period of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of the time assumed part of normal equipment maintenance, e.g., gaskets, in the accident analysis unless a time margin of less o-rings, coils, these may be replaced with identical than one hour can be justified. This justification must components.
include, for each piece of equipment, (1) consideration of a spectrum of breaks, (2) the potential need for the b. The item to be replaced is a component that is equipment later in an event or during recovery opera- part of an item of equipment qualified as an assembly;
tions, (3) .a determination that failure of the equipment these may be replaced with identical components.
after performance of its safety function. will not be detrimental to plant safety or mislead the. operator, c. Identical equipment to be used as a replacement and (4) a determination that the margin applied to the was on hand as a part of the utility's stock prior to minimum operability time, when combined with the February 22, 1983.
other test margins, will account for the uncertainties associated with the use of analytical techniques in the d. Replacement equipment qualified in accordance derivation of environmental parameters, the number of with the provisions of § 50.49 does not exist.
units tested, production tolerances, and test equipment imaccuracies. For all other equipment (e.g., postaccident e. Replacement equipment qualified in accordance monitoring, recombiners), the 10% time margin identified with the provisions of § 50.49 is not available to meet in Section 6.3.1.5 of IEEE Std 323-1974 should be installation and operation schedules. However, in such used. case, the replacement equipment may be used only until upgraded equipment can be obtained and an outage of
5. Section 6.3.3, "Aging," -of IEEE Std 323-1974 sufficient duration is available for replacement.
and paragraph 50.49(e)(5) should be supplemented with the following: f. Replacement equipment qualified in accordance with § 50.49 would require significant plant modifica- a. If synergistic effects have been identified prior tions to accommodate its use.
to the initiation of qualification, they should be accounted for in the qualification program. Synergistic effects g. The use of replacement equipment qualified in known at this time are dose rate effects and effects accordance with § 50.49 has a significant probability of resulting from the different sequence of applying radia- creating human factor problems that would negatively tion and (elevated) temperature. affect plant safety and performance, for example:
b. The expected operating temperature of the (1) Knowledge, skils, and ability of existing equipment under service conditions should be accounted plant staff would require significant upgrading to operate for in thermal aging. The Arrhenius methodology is or maintain the specific replacement equipment;
considered an acceptable method of addressing accelerated thermal aging within the limitation of state-of-the-art (2) The use of the replacement equipment technology. Other aging methods will be evaluated on a would create a one-of-a-kind application; or case-by-case basis.
(3) Maintenance, surveillance, or calibration activ- c. The aging acceleration rate and activation ities would be unnecessarily complex.
energies used during qualification testing and the basis upon which the rate and activation energy were estab- 7. In addition to the requirements of paragraph lished should be defined, justified, and documented. 50.490) of 10 CFR Part 50 and Section 8, "Documen- tation," of IEEE Std 323-1974, documentation should d. Periodic surveillance and testing programs are address the information identified in Appendix E to this acceptable to account for uncertainties regarding age- guide. A record of the qualification should be maintained related degradation that could affect the functional in an auditable file to permit verification that each item capability of equipment. Results of such programs will of electric equipment is qualified to perform its safety be acceptable as ongoing qualification to modify desig- function under its postulated environmental conditions nated life (or qualified life) of equipment and should be throughout its installed life.
incorporated into the maintenance and refurbishment/
replacement schedule
s.
D. IMPLEMENTATION
6. Replacement electric equipment installed subse- The purpose of this section is to provide information quent to February 22, 1983, must be qualified in accor- to applicants and licensees regarding the NRC staff's dance with the provisions of § 50.49 unless there are plans for using this regulatory guide.
sound reasons to the contrary. The NRC staff considers the following to be sound reasons for the use of replace- Except in those cases in which the applicant or ment equipment previously qualified in accordance with licensee proposes an acceptable alternative method for the DOR Guidelines or NUREG-0588 in lieu of upgrading: complying with specified portions of the Commission's
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regulations, the methods described herein will be used in the NRC has previously required qualification of that the evaluation of the qualification of electric equipment equipment in accordance with "Guidelines for Evaluating for all operating plants and plants that have not received Environmental Qualification of Class IE Electrical Equip- an operating license subject to the following: ment in Operating Reactors" (DOR Guidelines), or NUREG-0588, "Interim Staff Position on Environmental In accordance with paragraph 50.49(k), applicants for Qualification of Safety-Related Electrical Equipment."
and holders of operating licenses are not required to These applicants and licensees may continue to use the requalify electric equipment important to safety (replace- criteria in these documents for qualifying electric equip- ment equipment excepted) in accordance with the pro- ment important to safety in the affected plants, with visions of § 50.49 and in accordance with this guide if the exception of replacement equipment.
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APPENDIX A
TYPICAL SAFETY-RELATED ELECTRIC EQUIPMENT OR SYSTEMS*
Engineered Safety Feature Actuation Emergency Core Cooling Reactor Protection Containment Heat Removal Containment Isolation Containment Fission Product Removal Steamline Isolation Containment Combustible Gas Control Main Feedwater Shutdown and Isolation Auxiliary Feedwater Emergency Power Containment Ventilation Containment Radiation Monitoring
- Paragraph 50.49(b)(1) identifies safety-related electric equip- Control Room Habitability System (e.g., HVAC, Radiation ment as a subset of electric equipment important to safety and defines it as the equipment that is relied upon to remain func- Filters)
tional during and following design basis events to ensure (1) the Ventilation for Areas Containing Safety Equipment integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe Component Cooling shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential Service Water offsite exposurs comparable to the 10 CFR Part 100 guidelines. Emergency Systems to Achieve Safe Shutdown
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APPENDIX B I
TYPICAL EXAMPLES OF NON-SAFETY-RELATED EQUIPMENT
Associated circuits, as defined in Regulatory Guide core cooling system pump) will include termination
1.75, "Physical Independence of Electric Systems," need commands on loss, of lubrication oil pressure or low only be qualified to ensure that they will not fail under suction pressure. These features are provided for equip- postulated environmental conditions in a manner that ment protection. Failure of these features, however, could prevent satisfactory accomplishment of safety would defeat the safety-related function. They must functions by safety-related equipment. therefore be environmentally qualified.
The equipment identified in Examples 1, 2, and 3 Example 3 has typically been classified as safety-related on recently licensed plants. However, some operating plants were A safety-related fluid system may have non-safety- licensed using less definitive safety classification criteria related portions of the system that are isolated from the than those applied to recent designs, and they may safety-related portions of the system upon the generation contain non-safety-related equipment such as that in of a safety feature actuation signaL Isolation may be Examples 1, 2, and 3. The provisions of § 50.49 performed by motor-operated valves. These valve opera- require that the licensee provide appropriate environ- tors must be environmentally qualified.
mental qualification for equipment described in these examples regardless of the safety classification of that Example 4 equipment.
Harsh environments associated with HELBs could Example 4 applies to some plants, depending on the cause control system malfunctions resulting in conse- specific location of control system components. quences more severe than those for the HELBs analyzed in the FSAR (Chapter 15) or beyond the capability of operators or safety systems. In these cases, the control Example 1 system failures could prevent satisfactory accomplish- ment of the safety functions required for the HELBs.
The injection of emergency feedwater (EFW) for Typical examples of control systems that could fail as a PWRs and high-pressure coolant iection (HPCI) for result of an HELB and whose consequential failure may BWRs are safety-related functions. The EFW system not be bounded by HELBs analyzed in the FSAR are:
and the HPCI system are initiated upon detection of low water leveL Automatic termination of these systems 1. The automatic rod control system, upon detection of high water level may also be provided.
The high-level trip in some cases has been considered an 2. The pressurizer power-operated relief valve control equipment protection device; however, the inadvertent system, termination of EFW or HPCI due to misoperation of the level sensing equipment when subjected to a harsh 3. The main feedwater control system, environment could defeat the safety-related injection function. Thus the electric equipment associated with 4. The steam generator power-operated relief valve automatic termination of the injection must be envi- control system, and ronmentally qualified.
5. The turbine generator control system.
Example 2 Based on the above, it may be necessary to environ- In some cases, the electrical control system for a mentally qualify components associated with various pump (for example, a charging pump or an emergency control systems.
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APPENDIX C
METHODS FOR CALCULATING MASS AND ENERGY RELEASE
LOSS-OF-COOLANT ACCIDENT MAIN STEAM LINE BREAK
Acceptable methods for calculating the mass and Acceptable methods for calculating the mass and energy release to determine the loss-of-coolant accident energy release to determine the main steam line break environment for PWR and BWR plants are described in environment are described in the following:
the follovCing:
1. Topical Report WCAP-8822 (MARVEL/TRANSFLA)
1. Topical Report WCAP-8312A for Westinghouse for Westinghouse plants. Use of this method is accept- plants. able for all Westinghouse plants with the exception that a plant-specific containment temperature analysis
2. Section 6.2.1 of CESSAR System 80 PSAR for will be required for ice condenser containments.
Combustion Engineering plants.
2. Appendix 6B of CESSAR System 80 PSAR for
3. Appendix 6A of B-SAR-205 for Babcock & Combustion Engineering plants.
Wilcox plants.
3. Section 15.1.14 of B-SAR-205 for Babcock &
4. NEDO-10320 and Supplements 1 and 2 for General Wilcox plants.
Electric plants. NEDO-20533 dated June 1974 and Supplement 1 dated August 1975 for GE Mark IIL 4. Same as item 4 above for General Electric plants.
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APPENDIX D 4
METHODOLOGY
AND SAMPLE CALCULATION
FOR QUALIFICATION RADIATION DOSE
This appendix illustrates the staff model for calculat- 1. BASIC ASSUMPTIONS USED IN THE ANALYSIS
ing dose rates and integrated doses for equipment qualification purposes. The doses shown in Figure b- Gamma and beta doses and dose rates should be include contributions from airborne and plateout radia- determined for three types of radioactive source distri- tion sources in the containment and cover a period of butions: () activity suspended in the containment one year following the postulated fission product release. atmosphere, (2) activity plated out on containment The dose values shown are provided for illustration surfaces, and (3) activity mixed in the containment only and may not be appropriate for plant-specific sump water. A given piece of equipment may receive a application for equipment qualification levels. The dose dose contribution from any or all of these sources. The levels intended for qualification purposes should be amount of dose contributed by each of these sources is determined using the maximum time the equipment is determined by the location of the equipment, the intended to function. It should be noted, however, that time-dependent and location-dependent distribution of for equipment that must be qualified for more than the source, and the effects of shielding.
thirty days, a source term that incorporates considerable quantities of cesium as suggested by the accident at Following the TMI-2 accident, the staff concluded Three Mile Island Unit 2 (TMI-2) may produce doses that a thorough examination of the source term assump- greater than those estimated by the present source tions for equipment qualification was warranted. It is term. recognized, however, that the TMI-2 accident represents only one of a number of possible accident sequences
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leading to a release of fission products and that the mix X Beta Dose Rads) of fission products released under various core conditions
0 Gamma DoseRntgensl could vary substantially.
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Research under way may lead to modifications in LO source term assumptions. The research will consider the
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experience from the TMI-2 accident of 1979, con- temporary fission product release phenomenology, the transport and attenuation of fission products in primary I-'1 coolant systems and containments, and distinctions o,
between design basis accidents and events beyond the design basis. This research may result in revision of this guide.
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2. ASSUMPTIONS USED IN CALCULATING FISSION
TIM EIHOURS)
PRODUCT CONCENTRATIONS
Figure D-1 Sample Airborne and Plateout Ooses for a Dose Point on the containment Centertine This section discusses the assumptions used to simulate the PWR and BWR containments for determining the time-dependent and location-dependent distribution of The beta and gamma integrated doses presented in the airborne noble gas and iodine activity within the Tables b-1 and b-2 and Figure D-1 have been determined containment atmosphere, the activity plated out on using models and assumptions contained in this appendix. containment surfaces, and the activity in the sump This analysis incorporates the important time-dependent water.
phenomena related to the action of engineered safety features (ESFs) and such natural phenomena as iodine The staff used a computer program, TACT, to model plateout, as in previous staff analyses. the time-dependent behavior of iodine and noble gases within a nuclear power plant. The TACT code or other Doses were calculated for a point inside the contain- equivalent industry codes would provide an acceptable ment (at the midpoint of the containment) taking method for modeling the transfer of activity from one sprays and plateout mechanisms into account The containment region to another and for modeling the doses presented in Figure D-1 are values for a PWR reduction of activity due to the action of ESFs. Another -
plant having a containment free volume of 2.5 million staff code, SPIRT (Ref. 1), is used to calculate the cubic feet and a power rating of 4100 MWt. removal rates of elemental iodine by plateout and
1.89-10
O sprays. These codes were used to develop the source term estimates. The assumptions in the. following sectionss were used to calculate the distribution of radioactivity of iodine. Further, this model assumes that during the recirculation phases, the pH of the sump water is maintained above 8.5.
within the containment following a design basis LOCA.
8. The spray removal rate constant () was calculated
2.1 PWR Dry Containments using the staff's SPIRT program, conservatively assuming the operation of only one spray train and an instanta- The following methods and assumptions were used by neous partition coefficient (H) for elemental iodine of the staff for calculating the radiation environment in 5000. The calculated value of the spray removal constant PWR dry containments: for elemental iodine was 27.2 hi".
l. In the analysis of the accident radiation environ-- 9. Natural deposition (ie., plateout) of airborne ment, the staff assumed that 50% of the iodine core activity should be determined using a mechanistic model activity inventory and 100% of the core noble gas (see Reference 1). In the staff's example, plateout of activity inventory were released instantaneously to the iodine on containment internal surfaces was modeled as containment atmosphere. One percent of the remaininE a first-order rate removal process, and best estimates for
"solids" activity inventory was assumed released from model parameters were assumed. Based on an assumed the core and carried with the primary coolant directly total surface area within containment of approximately to the containment sump. 5.0 x 105 ft 2 , the calculated value for the overal plateout constant for elemental iodine was 1.23 hI'.
2. The containment free volume was taken as 2.52 x The assumption that 50% of the activity is instanta-
106 ft3 . Of this volume, 74% or 1.86 x 106 ft3 was neously plated out should not be used.
assumed to be directly covered by the containmenl sprays, leaving 6.6 x 05 ft3 of the containment free 10. The spray removal and plateout processes were volume unsprayed. The latter includes regions within the modeled as competing iodine removal mechanisms.
main containment space under the containment dome Removal of iodine from surfaces by the flow of con- and compartments below the operating floor level densed steam or by washoff by the containment spray (Plants with different containment free volumes shouleI may be assumed if such effects can be verified and use plant-specific values.) quantified by analysis or experiment.
- 3. The initial distribution of activity within the 11. A spray removal rate constant (X) for particulate containment should be based on realistic assumptions ;. iodine concentration was calculated using the staffs The staffs examples assumed a relatively open (non SPIRT program (Ref. 1). The staff calculated a value of compartmented) containment with a large release uni A = 0.43 hi" and allowed the removal of particulate formly distributed in the containment. This is a reason - iodine to continue until the airborne concentration was able simplification for dose assessment in a large dryr reduced by a factor of 104. The organic iodine concen- PWR containment and it is realistic in terms of specify - tration in the containment atmosphere is assumed ing the time-dependent radiation environment in mosi t not to be affected by either the containment spray or areas of the containment. plateout removal mechanisms.
4. The ESF fans were assumed to have a design flowv i2. The sprays were assumed to remove elemental rate of 220,000 cfm in the post-LOCA environment iodine until the instantaneous concentration in the Mixing between all major unsprayed regions and compart - sprayed region was reduced by a factor of 200. This is ments and the main sprayed region was assumed. necessary to achieve an equilibrium airborne iodine concentration consistent with previous LOCA analyses.
f- - 5. Effects of the ESF systems that remove airborm activity or redistribute activity within containment (e.g. 13. The analysis assumed that more than one species containment spray and containment ventilation systems' of radioactive iodine is present in a design basis LOCA.
should be evaluated using assumptions consistent wit] I The calculation of the post-LOCA environment assumed previous licensing practice. For example, the air exchange that, of the 50% of the core inventory of iodine released, between the sprayed and unsprayed regions was assume( 1 5% is associated with airborne particulate materials, 4%
to be one-half of the design flow rate of the ESF fans forms organic compounds, and 91% remains as elemental Good mixing of the containment activity between th e iodine. For conservatism, this composition was assumed sprayed and unsprayed regions is ensured by naturaI present at time t = 0. (These assumptions concerning convection currents and ESF fans. the iodine form are obtained from Regulatory Guides
1.3, "Assumptions Used for Evaluating the Potential
6. The containment spray system was assumed ti3 Radiological Consequences of a Loss-of-Coolant Accident m
- have two equal-capacity trains each designed to irect
3000 gpm of boric acid solution into the containment
7. Trace levels of hydrazine were assumed to b e for Boiling Water Reactors," and 1.4, "Assumptions Used for, Evaluating the Potential Radiological Conse- quences of a Loss-of-Coolant Accident for Pressurized Water Reactors," when a plateout factor of 2 is assumed added during the injection phase to enhance the removaIl for the elemental form.)
L
- 'r .1 1.89-11
14. The staff analysis conservatively assumed that no For the Mark I and Mark II designs, all of the activity leakage from the contaimnent. building to the environ- should be assumed initially released to the drywell area ment occurred. and the transfer of activity from these regions via containment leakage to the surrounding reactor building
15. Removal of airborne activity by engineered safety volume should be used to predict the qualification levels features may be assumed when calculating the radiation within the reactor building (secondary containment).
environment following other non-LOCA design basis accidents provided the safety features systems are 3. Removal of airborne iodine in the drywell or automatically activated as a result .of the accident. reactor building by the action of both plateout and'
spray processes may be assumed provided the effeo-
16. The radiation environment resulting from normal tiveness of these competing iodine removal processes are operation should be based on the conservative source evaluated using conditions and assumptions consistent term estimates reported in the plant's Safety Analysis with items 6 through 12 in Section 2.1 and plant-specific Report or should be consistent with the primary coolant parameters.
specific activity limits contained in the plant's technical specifications. The use of equilibrium primary coolant 4. The removal of airborne activity from the reactor concentrations based on 1% fuel cladding failures would building by operation of the standby gas treatment be one acceptable method. system (SGTS) may be assumed.
2.2 PWR Ice Condenser Containments 3. MODEL FOR CALCULATING THE DOSE RATE OF
AIRBORNE AND PLATEOUT FISSION PRODUCTS
The assumptions and methods presented for calculating the radiation environment in PWR dry containments are The beta and gamma dose rates and integrated doses appropriate foi use in calculating the radiation environ- from the airborne activity within the containment ment for ice condenser containments following a design atmosphere were calculated for the midpoint in the basis LOCA with the following modifications: containment. The containment was modeled as a cylinder with the height and diameter equaL Containment
1. The source should be assumed to be initially shielding and internal structures were neglected because released to the lower containment compartment. The they would involve a degree of complexity beyond the distribution of the activity should be based on the scope, of the present work. The calculations of Refer- forced recirculation fan flow rates and the transfer rates ence 2 indicate that the specific internal shielding and through the ice beds as functions of time. structure would be expected to reduce the gamma doses .. I
and dose rates by factors of two or more depending on
2. Credit may be taken for iodine removal via the the specific location and geometry.
operation of the ice beds and the spray system. A
time-dependent removal efficiency consistent with the Because of the short range of the betas in air, the steam/air mixture for elemental iodine may be assumed. airborne beta doses presented in Tables D1 and D-2 were calculated using an infinite medium approximation.
3. Removal of airborne iodine in the upper compart- This is shown in Reference 3 to result in only a small ment of the containment by the action of both plateout error. Beta doses for equipment located on the contain- and spray processes may be assumed provided these ment walls or on large internal structures may be removal processes are evaluated using conditions and calculated using the serniinfinite beta dose model.
assumptions consistent with items 6 through 12 in Section 2.1 and plant-specific parameters. The staff recognizes that this approach is conservative and that, for most plant-specific calculations, a semi-
2.3 BWR Containments infinite beta dose model may be more appropriate. The use of the semiinfinite model is acceptable provided The assumptions and methods presented for calculating there is sufficient justification for its use (such as the radiation environment in PWR dry containments are location, shielding, minimal thickness). Further, the staff appropriate for use in calculating the radiation environ- recognizes that for some equipment the use of a finite- ment for BWRs following a design basis LOCA with the cloud beta dose model may be warranted. Because the following modifications: use of the finite-cloud model would result in beta doses much smaller than the values presented in Table D-2, a
1. A decontamination factor (DF) of 10 may be case-by-case justification for use of the finite-cloud assumed for bath elemental and particulate iodine as the model will be required.
iodine activity passes through the suppression pooL No credit should be taken for the removal of organic iodine The gamma dose rate contribution from the plated- or noble gases in the suppression pooL out -iodine on containment surfaces to the point on the
2. For Mark III designs, all of the activity passing through the suppression pool should be assumed instanta- neously and uniformly distributed within the containment.
centerline was also included. The model calculated the plateout activity in the containment assuming only one spray train and one ventilation system were operat- ing. It should be noted that washoff of the plated-out U
1.89-12
iodine activity by the sprays was not addressed in this calculated on the assumption of a time-dependent sump evaluation. iodine buildup is not significant.)
Finally, all gamma doses were multiplied by a correc- The "solid" fission products should be assumed to be tion factor of 1.3 as suggested in Reference 3 to account instantaneously carried by the coolant to the sump and for the omission of the contribution from the decay uniformly distributed in the sump water. The gamma chains of the isotopes. and beta dose rates and the integrated doses should be computed for a center point located at the surface of the large pool of sump water, and the dose rate
4. MODEL FOR CALCULATING THE DOSE RATE calculation should include an estimate of the effects OF SUMP FISSION PRODUCTS of buildup.
The staff model assumed the washout of airborne
5. CONCLUSION
iodine from the containment atmosphere to the contain- ment sump. For a PWR containment with sprays and The values given in Tables D-1 and D-2 and Figure good mixing between the sprayed and unsprayed regions, D-1 for the various locations in the containment provide the elemental iodine (assumed to constitute 91% of the an estimate of expected radiation qualification values for released iodine) is very rapidly washed out of the a 4100 MWt PWR design.
atmosphere to the containment sump (typically 90% of the airborne iodine in less than 15 minutes). The NRC Office of Nuclear Regulatory Research is continuing its research efforts in the area of source The dose calculations may assume a time-dependent terms for equipment qualification following design basis iodine source. (The difference between the integrated accidents. As more information in this area becomes dose calculated on the assumption of 50% of the core available, the source terms and staff models may change iodine immediately available in the sump and that to reflect the new information.
1.89-13
Table D-1 ESTIMATES FOR TOTAL AIRBORNE GAMMA DOSE
CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER I
Time Airborne Iodine Airborne Noble Plateout Iodine Total Dose (Hr) Dose (R) Gas Dose (R) ' Dose (R) (R)
0.00
0.03 4.82E+4 7.42E+4 1.69E+3 1.24E+5
0.06 8.57E+4 1.39E+5 3.98E+3 2.29E+5
0.09 1.09E+5 1.98E+5 7.22E+3 3.14E+5
0.12 1.25E+5 2.51E+5 l.lOE+4 3.87E+5
0.15 1.38E+5 3.01E+5 1.52E+4 4.54E+5
0.18 1.47E+5 3.48E+5 1.96E+4 5.1SE+5
0.21 1.55E+5 3.92E+5 2.41E+4 5.71E+5
0.25 1.64E+5 4.49E+5 3.03E+4 6.43E+5
0.38 1.87E+5 6.19E+5 5.05E+4 8.57E+5
0.50 2.03E+5 7.6 1E+5 6.90E+4 1.03E+6
0.75 2.36E+5 1.03E+6 1.06E+5 1.37E+6
1.00 2.66E+5 1.26E+6 1.40E+5 1.67E+6
2.00 3.62E+5 2.04E+6 2.61E+5 2.66E+6
5.00 5.SOE+5 3.56E+6 5.40E+5 4.65E+6
8.00 6.63E+5 4.38E+6 7.47E+5 5.79E+6
24.0 1.O1E+6 6.26E+6 1.45E+6 8.72E+6
60.0 1.31E+6 7.16E+6 2.10E+6 1.06E+7
96.0 1.45E+6 7.56E+6 2.39E+6 1.14E+7
192 1.68E+6 8.29E+6 2.86E+6 1.28E+7
298 1.85E+6 8.76E+6 3.19E+6 1.38E+7
394 1.95E+6 8.85E+6 3.41E+6 1.42E+7
560 2.07E+6 9.06E+6 3.64E+6 1.48E+7
720 2.13E+6 9.15E+6 3.76E+6 1.SOE+7
888 2.16E+6 9.19E+6 3.83E+6 1.52E+7
1060 2.18E+6 9.21EE+6 3.87E+6 1.53E+7
1220 2.19E+6 9.21E+6 3.89E+6 1.53E+7
1390 2.20E+6 9.21E+6 3.90E+6 1.53E+7
1560 2.20E+6 9.22E+6 3.91E+6 1.53E+7
1730 2.20E+6 9.22E+6 3.91E+6 1.53E+7
1900 2.20E+6 9.22E+6 3.92E+6 1.53E+7
2060 2.20E+6 9.22E+6 3.92E+6 1.53E+7
2230 2.20E+6 9.22E+6 3.92E+6 1.53E+7
2950 2.20E+6 9.23E+6 3.92E+6 1.54E+7
3670 2.20E+6 9.24E+6 3.92E+6 1.54E+7
4390 2.20E+6 9.24E+6 3.92E+6 1.54E+7
5110 2.20E+6 9.25E+6 3.92E+6 1.54E+7
5830 2.20E+6 9.25E+6 3.92E+6 1.54E+7
6550 2.20E+6 9.26E+6 3.92E+6 1.54E+7
7270 2.20E+6 9.27E+6 3.92E+6 1.54E+7
8000 2.20E+6 9.27E+6 3.92E+6 1.54E+7
8710 2.20E+6 9.28E+6 3.92E+6 1.54E+7 Total 1.54E+7
1.89-14
Table D-2 ESTIMATES FOR TOTAL AIRBORNE BETA DOSE
CONTRIBUTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER
Time Airborne Iodine Airborne Noble Total Dose (Hr) Dose (rads)* Gas Dose (rads)* (rads)*
0.00
0.03 1.47E+5 5.48E+5 6.95E+S
0.06 2.62E+5 9.86E+S 1.25E+6
0.09 3.33E+5 1.35E+5 1.68E+6
0.12 3.83E+S 1.65E+6 2.03E+6
0.15 4.20E+S l.91E+6 2.33E+6
0.18 4.49E+5 2.14E+6 2.59E+6
0.21 4.73E+5 2.35E+6 2.82E+6
0.25 5.OOE+5 2.60E+6 3.10E+6
0.38 5.67E+5 3.30E+6 3.87E+6
0.50 6.15E+5 3.86E+6 4.48E+6
0.75 7.13E+5 4.89E+6 5.60E+6
1.00 8.OOE+5 5.81 E+6 6.61 E+6
2.00 1.07E+6 9.02E+6 1.OIE+7
5.00 1.58E+6 1.65E+7 1.81E+7
8.00 1.88E+6 2.20E+7 2.39E+7
24.0 2.87E+6 4.08E+7 4.37E+7
60.0 3.89E+6 6.15E+7 6.54E+7
96.0 4.37E+6 7.48E+7 7.92E+7
192 5.14E+6 1.OOE+8 1.05E+8
298 5.64E+6 1.17E+8 1.23E+8
394 5.99E+6 1.25E+8 1.3 1E+8
560 6.34E+6 1.34E+8 1.40E+8
720 6.53E+6 1.39E+8 1.46E+8
888 6.63E+6 1.42E+8 1.49E+8
1060 6.69E+6 1.44E+8 1.51 E+8
1220 6.73E+6 1.45E+8 1.52E+8
1390 6.75E+6 1.47E+8 1.54E+8
1560 6.76E+6 1.49E+8 1.56E+8
1730 6.76E+6 1.51 E+8 1.58E+8
1900 6.76E+6 1.52E+8 1.59E+8
2060 6.76E+6 1.54E+8 1.61E+8
2230 6.77E+6 1.55E+8 1.62E+8
2950 6.77E+6 1.62E+8 1.69E+8
3670 6.77E+6 1.69E+8 1.76E+8
4390 6.77E+6 1.76E+8 1.83E+8 SllO 6.77E+6 1.83E+8 1.90E+8
5830 6.77E+6 1.89E+8 1.96E+8
6550 6.77E+6 1.96E+8 2.03E+8
7270 6.77E+6 2.03E+8 2.1OE+8
8000 6.77E+6 2.09E+8 2.16E+8
8710 6.77E+6 2.16E+8 2.23E+8 Total 2.23E+8
- Dose conversion factor is based on absorption by tissue.
Li>. ~~~~~~~~~~~~~~1.89-15
APPENDIX D I
REFERENCES
1 A. K. Postma, R. R. Sherry, and P. Tam, "Techno- Containments," in Transactions of the American logical Bases for Models of Spray Washout and Nuclear Society, Vol. 23, pp. 604-605, 1976.
Airbome Contaminants in Containment Vessels,"
U.S. Nuclear Regulatory Commission, NUREG/CR-
0009, November 1978.* 3. M. J. Kolar and N. C. Olson, "Calculation of Acci- dent Doses to Equipment Inside Containment of
2. E. A. Warman and E. T. Boulette, "Engineering Power Reactors," in 7ransactions of the American Evaluation of Radiation Environment in LWR Nuclear Society, VoL 22, pp. 808-809, 1975.
BIBLIOGRAPHY
Kocher, D. C., Ed., "Nuclear Decay Data for Radio- the American Nuclear Society, VoL 18, pp. 358-359, nuclides Occurring in Routine Releases from Nuclear 1974.
Fuel Cycle Facilities," ORNL/NUREG/TM-102, August
1977.* Postma, A. K., and R Zavadoski, "Review of Organic Iodide Formation Under Accident Conditions in Water Lorenz, R. A., J. L Collins, and A. P. Malinauskas, Cooled Reactors," U.S. Nuclear Regulatory Commission,
"Fission Product Source Terms for the LWR Loss-of- WASH-1233, pp. 62-64, October 1972.*
Coolant Accident: Summary Report," U.S. Nuclear Regulatory Commission, NUREGICR-0091, May 1978.* Rogovin, M., et a, "Three Mile Island-A Report to the Commissioners and to the Public," NUREG/CR-1250,
Normand, E., and W. R. Determan, "A Simple Algorithm Volume II, Part 2, April 5, 1979.*
to Calculate the Immersion Dose," in Transactions of-
- Copies are available from the National Technical Informa- U.S. Nuclear Regulatory Commission, "Technical Basis for Estimating Fission Product Behavior During LWR
4 tion Service, Springfield, Virginia 22161. Accidents," NUREG-0772, June 1981.*
I
1.89-16
APPENDIX E
QUALIFICATION DOCUMENTATION FOR ELECTRIC EQUIPMENT
In order to ensure that an environmental qualification accident environment for the time during which it program conforms to General Design Criteria 1, 2, 4, must not fail with safety margin to failure.
and 23 of Appendix A; Sections III, XI, and XVII of Appendix B; and § 50.49 of 10 CFR Part 50, the c. Equipment that will experience environmental following information on the qualification program conditions of design basis accidents through which it should be submitted to NRC for electric equipment need not function for mitigation of such accidents and within the scope of this guide: whose failure (in any mode) is deemed not detrimental to plant safety or accident mitigation; it need not be
1. Provide a list of all electric equipment within the qualified for any accident environment.
scope of this guide such as the following:
d. Equipment that has performed its safety func- a. Switchgear tion prior to the exposure to an accident environment b. Motor control centers and whose failure (in any mode) is deemed not detri- c. Valve operators and solenoid valves mental to plant safety and will not mislead the opera- d. Motors tor; it need not be qualified for an accident environment.
e. Logic equipment f. Cable 4. For each item of equipment in the categories of g. Connectors equipment listed in item 3, provide the following:
IL Sensors (pressure, pressure differential, tempera- ture, flow and level, neutron, and other radiation) a. The system safety function requirements for i Limit switches equipment in categories 3.a, 3.b, and 3.d.
j. Heaters k. Fans b. An environmental envelope as a function of L Control boards time that includes all extreme parameters, both maxi- m.Instrument racks and panels mum and minimum values, expected to occur during n. Electric penetrations plant shutdown and design basis accident (including t
o. Splices LOCA and MSLB), including postaccident conditions, p. Terminal blocks for equipment in categories 3.a and 3.b.
2. For each item of equipment identified in 1, c. Length of time equipment in categories
3. a and
1; provide the following:
a.
b.
c.
Type (functional designation)
Manufacturer Manufacturer's type number and model number
3.b must perform its safety function when subjected to any of the limiting environment specified above.
d. The technical bases that justify the placement of each item of equipment in categories 3.b, 3.c, and d. Plant ID/tag number and location 3.d.
t-i -.
3. Categorize the equipment identified in item I into 5. For each item of equipment identified in categor- one of the following categories: ies 3.a and 3.b, state the actual qualification envelope simulated during testing (defining the duration of the a. Equipment that will experience the environmental environment and the margin in excess of the design conditions of design basis accidents through which it requirements). If any method other than type testing must function to mitigate such accidents; it must be was used for qualification, identify the method and qualified to demonstrate operability in the accident define the equivalent "qualification envelope" so derived.
environment for the time required for accident mitigation r with safety margin to failure. 6. Provide a summary of test results that demon- strates the adequacy of the qualification program. If any b. Equipment that will experience environmental analysis is used for qualification, justification of all conditions of design basis accidents through which it analysis assumptions must be provided.
need not function for mitigation of such accidents but through which it must not fail in a manner detrimental 7. Identify the qualification documents that contain to plant safety or accident mitigation; it must be quali- detailed supporting information, including test data, for p fied to demonstrate the capability to withstand any items 5 and 6.
1.89-17
I
VALUE/IMPACT STATEMENT
Background requirements for equipment qualification. Methods for establishing temperature and pressure profiles for a The Commission (in Memorandum and Order CLI-80- loss-of-coolant accident and main steam line break are
21 dated May 23, 1980) directed the staff to use provided, and radiological source terms are given.
NUREG-0588, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," 3. Regulatory Position C.3, which provides the staff along with a document entitled "Guidelines for Evaluat- position pertaining to test procedures.
ing Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines, 4. Regulatory Position C4, which provides the staff January 14, 1980) as requirements that licensees and position regarding establishing margin in testing require- applicants must meet in order to satisfy the equipment ments.
qualification requirements of 10 CFR Part 50. Subse- quently, the Commission approved a final rule for 5. Regidatory Position C.5, which provides the staff electric equipment qualification (§ 50.49 of 10 CFR position regarding aging of equipment.
Part 50). Revision 1 to Regulatory Guide 1.89 will provide an acceptable mAethod for meeting the require- 6. Regulatory Position C6, which provides the ments of § 50.49. staff position regarding qualification of replacement equipment.
Substantive Changes and Their Value/Impact
7. Regulatory Position C.7, which provides the staff The following positions were added in Revision I to position on the documentation of equipment quali- Regulatory Guide 1.89: fication procedures and results.
1. Regulatory Position C., which adds to the scope Value - This guide provides the staff's views on of the guide non-safety-related electric equipment whose individual sections of IEEE Std 323-1974 and describes failure under postulated environmental conditions could acceptable methods for meeting the requirements of prevent satisfactory accomplishment of safety functions (for example, the associated circuits defined in Regula- tory Guide 1.75, "Physical Independence of Electric Systems") and certain postaccident monitoring equipment.
§ 50.49 of 10 CFR Part 50. This guide should enhance the licensing process.
Impact - This regulatory guide does not impose any i
new costs or obligations on licensees or applicants.
2. Regulatory Position C.2, which provides the staff Thus, no impact will result from issuance of this guide position on establishing performance and environmental with respect to requirements in effect at this time.
4
1.89-18