ML20027D797
| ML20027D797 | |
| Person / Time | |
|---|---|
| Issue date: | 06/17/1981 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML20027A669 | List:
|
| References | |
| FOIA-82-426, TASK-OS, TASK-RS-042-2, TASK-RS-42-2 REGGD-01.089, REGGD-1.089, NUDOCS 8211100073 | |
| Download: ML20027D797 (42) | |
Text
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3 June 17, 1981 9
i Proposed Regulatory Guide 1.89 Rev. 1 ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS A.
INTRODUCTION Criterion III, " Design Control" and Criteria XI, " Test Control," of Appen-dix B, " Quality Assurance Criteria for Nuclear Power Plants and' Fuel Reprocessing Plants," to 10 CFR Part 50, " Licensing of Production and Utilization Facilities,"
requires that design control measures provide for verifying the adequacy of a specific design feature by design reviews, b) various calculational methods or by suitable qualification testing of a prototype unit under the most adverse conditions and that proof tests be conducted to demonstrate that structures, systems and components will perform satisfactorily in service.
General Design Criteria 1, 2, 4 and 23 of Appendix A to 10 CFR Part 50 and 950.49 " Environmental Qualification of Electric Equi.pment Important to Safety for Nuclear Power Plants," to 10 CFR Part 50, requires that each type of electric equipment be qualified for its application and specified performance requirements, and provides requirements for establishing qualification methods and environmental qualification parameters.
This regulatory guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to design verification of electric equipment for service in light-water-cooled nuclear power plants to assure that the equipment can perform functions that are iLportant to safety.
J 8211100073 821014 PDR FOIA CURRAN 82-426 PDR
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DISCUSSION IEEE Std 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations,"M ated February 28, 1974, was prepared by d
Subcommittee 2, Equipment Qualification, of the Nuclear Power Engineering Committee of the Institute of Electrical and Electronics Engineers, Inc. (IEEE),
j, and subsequently was approved by the IEEE Standards Board on' December 13, 1973.
The standard describes basic procedures for qualifying Class 1E equipment and l
interfaces that are to be used in nuclear power plants and components or equipment of any interface whose failure could adversely affect any Class 1E equipment.
The requirements delineated include principles, procedures, and methods of qualification which, when satisfied, will confirm the adequacy of the equipment design for the performance of safety functions under normal, abnormal, design-basis-event, post-design-basis-event, and containment-test conditions.
Equipment should be qualified to meet its performance requirements under the environmental and operating conditions in which it will be required to func-tion and for the length of time for'which its function is required.
The follow-ing are examples of considerations to be taken into account when determining the environment for which the equipment is to be qualified:
(1) equipment outside containment would generally see a less severe environment than equipment inside containment; (2) equipment whose location is shielded from a radiation source would generally receive a smaller radiation dose than equipment of equal distance from the source but exposed to its direct radiation; (3) equipment required to 1/ Copies may be obtained from the Institute of Electrical and Elect'ronics
~ Engineers, Inc., United Engineering Center, 345 East 47th Street, New York, New York 10017.
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The specific environment for which individual equipment must be qualified will depend on the installed location, the conditions under which it is required to function, and the length of time it is required to operate.
A component to be qualified in a nuclear radiation environment should be exposed to a fluence that simulates the total dose, conservatively calculated, that the component should withstand prior to completion of its intended func-tion.
Dose rates, spectrum, and particle type should be simulated as closely as practicable unless it can be shown that damage is not significantly depen-dent on dose rates, or spectrum, or particle type.
Equipment qualification is predicated on the assumption that qualification testing adequately simulated the environment and service conditions throughout the installed life of the equipment. Where routine maintenance is essential to maintaining equipment in the conditions simulated by the qualification test (e.g., cleanness), it is important that an adequate program of preventive maintenance and quality assurance be established, including minimizing dust accumulation that could degrade the ability of the equipment to function properly.
C.
REGULATORY POSITION The procecures described by IEEE Std 323-1974, "IEEE Standard for Quali-fying Class IE Equipment for Nuclear Power Generating Stations,"M ated Feb-d ruary 28,1974, are acceptable for qualifying electric equipment for service in light-water-cooled nuclear power plants to assure that the equipment can perform functions that are important to safety subject to the following:
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Reference is made in IEEE Std 323-1974, Sections 2, 6.3.2(5), and 6.3.5, to IEEE Std 344-1971, " Guide for Seismic Qualification of Class 1 Elec-tric Equipment for Nuclear Power Generating Stations." The specific applica-
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bility or acceptability of IEEE Std 344 is covered in Regulatory Guide 1.100.
2.
Section 5 of IEEE Std 323-1974 pertains to principles of qualifica-tions including various methods.
In conjunction with Section 5, the selection of a qualification method should be based on the following:
a.
The NRC will not accept analysis alone without supporting test data.
Experience has shown that qualification of equipment without test data may not be adequate to demonstrate functional operability during design basis event conditions.
Analysis may be acceptable provided (a) testing of the equipment is impractical due to size limitations, (b) testing is precluded by the state-of-the-art, and (c) prototype equipment is not available.
3.
Section 6.2 of IEEE Std 323-1974 pertains to establishing perform-ance and environmental requirements.
In conjunction with 6.2(7) of Section 6.2, the following should be used:
a.
Temperature and Pressure Conditions Inside Containment for loss of Coolant Accident (LOCA).
(1) Methods acceptable to the NRC staff, for calculating and establishing the containment pressure and temperature envelopes to which equipment should be qualified are provided below.
Methods for calculating mass and energy re-lease rates are summarized in Appendix A.
The calculations should account for the time dependence and spatial distribution of these variables.
High pressure is not necessarily a limiting condition.
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Dry Containment --Calculate LOCA containment environment using CONTEMPT-LT
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or equivalent industry codes. Additional guidance is provided in Standard 1
Review Plan (SRP) Section 6.2.1.1.A, NUREG-75/087.
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Ice Condenser Containment - Calculate LOCA containment environment using i
LOTIC or equivalent industry codes.
Additional guidance is provided in n
SRP Section 6.2.1.1.8, NUREG-75/087.
2 Boiling Water Reactors (BWRs) 1 Mark I, II and III Containment - Calculate LOCA environment using methods of GESSAR Appendix 38 or equivalent industry codes. Additional guidance i
is provided in SRP Section 6.2.1.1.C, NUREG-75/087.
(2) The test profiles included in Appendix A tn IEEE Std. 323-1974 should not be considered an acceptable alternative in lieu of using plant-specific containment temperature and pressure design profiles unles's plant-specific analysis is provided to verify :.ne applicability of those profiles.
ai b.
Temperature and. Pressure Conditions Inside Containment for Main i
j Ste i Line Break (MSLB).
Methods acceptable to the NRC staff for calculating the environmental l-parameters of a MSLB used for equipment qualification are provided below.
l" (1) Models that are acceptable for calculating containment parameters are listed in Position 3.a(1).
(2) The test profiles included in Appendix A to IEEE Std. 323-1974 should not be. considered an acceptable alternative in lieu of using plant-specific i
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containment temperature and pressure design profiles unless plant-specific j
analysis is provided to verify the applicability of those profiles.
11 c.
Effects of Chemicals i
Guidelines for the chemical spray solution are provided in SRP Sec-tion 6.5.2 (NUREG-75/087), paragraph II, item (e).
For plants which use demineralized water as spray solution, effect of spray should also be considered.
j d.
Radiation Conditions Inside and Outside Containment The radiation environment for qualification of equipment should be based on the normally expected radiation environment over the equipment installed life, plus that associated with the most severe design basis accident (DBA) ~
during or following that which equipment must remain functional.
It should be assumed that the DBA related environmental conditions occur at the most critical point of degradation during the equipment installed life, which may be at the end of its installed life, i
Methods acceptable to the NRC staff for establishing radiation limits for qualification.for BWR and PWR type reactors are provided in the sample calcula-tions in Appendix B and the following:
(1) The source term to be used in determining the radiation environment for equipment qualification associated with a design basis LOCA should consider the most limiting environment associated with the following:
(a) For a LOCA wh'ere the break cannot be isolated, 100% of the core activity inventory of noble gases and 50% of the core activity inventory of -
the halqgens should be assumed to be instantaneously released from the fuel to the containment.
Fifty percent of the cesium activity and 1% of the remaining
" solids" activity inventory in the core should be assumed'to be instantaneously i
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released from the fuel to the primary coolant and carried by the coolant to S-the containment sump.
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(b) For a LOCA where the break can be isolated, 100% of the core i
activity inventory of the noble gases, 50%_of the core activity inventory of
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the halogens, and 50% of the core activity inventory of the cesium and 1% of j
the remaining " solids" activity inventory should be assumed to be instantaneously W
i released (after an initial time delay) and circulated in the primary coolant
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This accident is not expected to produce instantaneous fuel damage.
A 30-minute delay may be assumed for fission product release from the fuel.
Greater delay times should be justified on the basis of-system performance-design that minimizes fission product release.
No noble gases should be assumed circulating in the primary system following system depressurization.
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(2) For all other design basis accidents (e.g., non-LOCA high energy-g line breaks, rod ejection or rod drop accidents) the qualification source terms should be calculated factoring in the percent of fuel. damage assumed in the plant specific analysis (provided in the FSAR). When only fuel clad perfora-tion is postul.ated, the nuclide inventory of the fuel elements breached should 7
be calculated at the end of core life, assuming continuous full power opera-L tion.
The fuel rod gap inventory in the rods should be assumed to be 10% of the total rod activity inventory of iodine and 10% of the total activity inven-F tory of the noble gases (except for Kr-85 for which a release of 30% should.be assumed).
All the gaseous constituents in the gaps of the breached fue1~ rods 3
should be assumed instantaneously released to the primary coolant. When fuel
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melting is postulated the activity inventory of the melted fuel elements should also be calculated at the end of core life assuming full power operation.
For
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this case, 100% of the noble gases, 30% of t!)e halogens, 50% of the cesium h
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should be assumed to be instantaneously released to the primary coolant.
d (3) For a limited number of accident monitoring instrumentation with instrument ranges that extend to the maximum values the selected parameters
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can attain under worst-case conditions, specified in Regulatory Guide 1.97,
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant
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and Environs Conditions During and Following an Accident," the source term should assume an initial release which considers the fission product release groups
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associated with grossly melted fuel.
An acceptable assumption of frac-j tional release for each group are:
noble gases, 100%; I, Br, 100%; Cs, Rb, 100%, Te, 100%; Sr, Ba. 11%; Ru, 8%; and La, 1.3% (individual nuclides are listed in Table VI 3-1 of WASH-1400).
The effect of natural and mechanical containment fission product removal may be considered on a best estimate basis to determine the rate of redistribution of the various groups from the contain-ment atmosphere to other locations.
(4) The calculation of the-radiation environment associated with design 1
basis accidents should take into account the time-dependent transport of re-li leased fission products within various regions of containment and auxiliary structures.
(5) The initial distribution of activity within the containment should be based on a mechanistically mtional assumption.
Hence, for compartmented containments, such as in some BWRs, 100% of the source should be assumed to be initially contained in the drywell.
For ice condenser containments, it should be assumed that 100% of the source is initially contained in the' lower portion a
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The assumption of uniform distribution of activity through-out a compartmented containment at time zero may not be appropriate.
'i (6) Effects of ESF systems, such as containment sprays and containment
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ventilation and filtration, systems, which act to remove airborne activity and j
redistribute activity within containment, should be calculated using the same assumptions used in the calculation of offsite dose.
See SRP Section 15.6.5 1j (NUREG-75/087) and the related sections referenced in the Appendices to that j
section.
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(7) Natural deposition (i.e., plate-out) of airborne activity should be determined using a mechanistic model and best estimates for the model parameters (See Ref. 3, Appendix B).
The assumption of 50 percent instantaneous plate-out of the iodine released from the core should not be made.
Removal of iodine from surfaces by steam condensate flow or washoff by the containment spray may be assumed if such effects can be justified and quantified by analysis or experiment.
(8) The calculated qualification dose should be the sum of the calculated doses of the potential radiation sources at the equipment location (i.e., beta and gamma), and may be established by one of the following:
.(a) The total qualification dose should be equivalent to the total cal-culated dose (beta plus gamma) at the equipment location.
A gamma source (only) may be used for qualification testing provided analysis or tests indicate that the doses and dose rate produce similar damage to that which would occur under accident conditions, i.e., a combination of beta and gamma, or
.i (b) The beta and gamma qualification dose may be determined separately and the testing may be performed using both a beta and gamma test source.
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i Plant specific analysis may be used to justify any reduction in dose or f
dose rate due to equipment location or shielding.
(9) Shielded components need be qualified only to the gamma radiation
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(10) Paints, coatings and coverings on electric equipment should be assumed a
l to be exposed to both beta and gamma dose and dose rates in assessing their resistance to radiation.
Plate-out activity should be assumed to remain on the equipment surface unless the effects of the removal mechanisms, such as spray wash-off or steam condensate flow, can be jusitifed and quantified by analysis or experiment.
(11) Equipment located outside containment exposed to recirculating fluid system should be qualified to withstand the radiation equivalent to that pene-trating the containment, plus the exposure from the recirculating fluid.
(12) Equipment that may be exposed to low level radiation doses [below 4
10 rads] should not be considered to be exempt from radiation qualification, unless analysis supported by test data or operating experience is provided to verify that its dose and dose rates will not degrade the operability of the eq'uipment below acceptable values.
(13) A given component may be considered to be qualified provided it can be shown that the component can be subjected, without failing, to the integrated beta and gamma doses, accounting for beta and gamma dose rates, which are equal to or higher than those levels resulting from an analysis that (1) is similar 3
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fj in nature and scope to that included in Appendix B [ Appendix B uses the source term given in Position C.3.d(1)], ano (2) incorporates appropriate factors perti-j nent to the plant design (e.g., reactor types and power level, containment size).
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Environmental Conditions for Outside Containment
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(1) Equipment important to safety, which are located outside containment 4
f and which could be subjected to high energy pipe breaks, as defined in the
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Standard Review Plan, should be qualified to the conditions resulting from an accident for the duration required. The techniques to calculate the environ-l mental conditions should employ a plant specific model based on good engineering i
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t (2) Equipment important to safety, which are located in general plant S
areas outside containment where equipment is not subjected to a design basis accident environment, should be qualified to the normal and abnormal range of i
environmental conditions postulated to occur at the equipment location.
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(3) Equipment important to safety not served by environmental support systems important to safety, or served by other systems important to safety that may be secured during plant operation or shutdown, should be qualified to the limiting environmental conditions that are postulated for that location, assuming a loss of the environmental support system.
4.
Section 6.3 of IEEE Std. 323-1974 pertains to type test procedures.
The following should be used in conjunction with Section 6.3:
a.
Equipment located in a mild environment defined in Positions 3.e.(2) and (3) are not required to be qualified by test.
The " Design / Purchase" 1
specifications which contain a description of the functional requirements of its specific environmental location during normal and abnormal. environmental 1
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conditions will generally be acceptable.
A well supported maintenance /
surveillance program, in conjunction with a good preventive maintenance program, should be provided to assure that equipment so qualified will function for its design life.
Furthermore, the maintenance / surveillance program data and records should be reviewed periodically (not more than 18 months) to assure that the design qualified life is not suffering thermal and cyclic degradation resulting
.i from the accumulated stresses of service conditions.
b.
Where equipment is located in watertight enclosures, qualification by test should be used to demonstrate the adequacy of such protection. Where equipment could be submerged, it should be identified and demonstrated to be qualified by test to demonstrate seal integrity and functional operability for the duration required.
Shortened test periods and analytical extrapolation should be justified.
c.
Where equipment is located in an area where rapid pressure changes L
are expected, qualification by test should demonstrate that, under the most adverse time dependent relative humidity conditions (superheated steam followed by saturated steam may be a limiting condition) and adverse postulated pressure transient for the equipment location, the equipment seals and vapor barriers will prevent moisture from penetrating into the equipment to the degree necessary to maintain equipment integrity for the length of time the equipment function is required.
d.
The temperature to which equipment is qualified, when exposed to the simulated environment, should be defined by temperature readings as close as e
practical to the component being qualified.
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Performance characteristics of equipment should be verified before, after, and periodically during testing throughout its range of required oper-j ability.
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Chemical spray or demineralized water spray should be incorporated during sittulated event testing at or near the maximum pressure and temperature I{
conditions that would occur when the spray systems actuate.
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g.
Variables indicating functional status of equipment should be moni-tored continuously to assure that spurious failures (if any) have been accounted
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For long-term testing, however, continuous monitoring during o!
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periodic intervals may be used if justified.
1 h.
Expected extranes in power supply voltage range and frequency should be applied appropriately during simulated event environmental testing.
i.
Cobalt-60 or Cessium-137 is an acceptable gamma radiation' source for environmental qualification.
5.
Section 6.3.1.5 of IEEE Std. 323-1974 pertains to margin.
In lieu of other proposed margins that may be found acceptable, the suggested values ij indicated in Section 6.3.1.5, should be used as a guide with the following
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exceptions:
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Quantified margins should be applied to the design parameters' dis-1 1 lj cussed in Position C.3 to assure that the postulated accident conditions have l
been enveloped during testing.
These margins should be applied in addition to 1
j any conservatism applied during the derivation of the specified plant param-F eters unless those conservatisms can be quantified and shown to contain suffi-cient margin.
The margins should (a) account for uncertainties associated with the use of analytical techniques in deriving environmental parameters, 3
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when best estimates methods are used rather than conservative licensing methods,
~i (b) account for uncertainties associated with defining satisfactory performance u
(e.g., when only a few units are tested (c) account for variations in the si commercial. production of the equipment, and (d) account for the inaccuracies in j
the test equipment to assure that the calculated parameters have been adequately
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b.
Some equipment may be required by the design to only perform its safety l
function within a short time period into the event (i.e., less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />),
l and, once its function is complete, subsequent failures are shown not to be
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detrimental to plant safety.
Other equipment may not be required to perform a i
safety function but must not fail within a short time period into the event, j
and subsequent failures are also shown not to be detrimental to plant safety.
Equipment in these categcries should remain functional in the accident environ-ment for a 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 in the acci-dent analysis.
For all other equipment (e.g., post-accident monitoring, recom-i biners, etc.), the 10 percent time margin identified in Section 6.3.1.5 of IEEE Std. 323-1974 should be used.
ij 6.
Section 6.3.3 of IEEE Std. 323-1974 pertains to aging.
In conjunc-
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tion with Section 6.3.3, the following should apply:
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a.
Where synergistic effects have been identified. (e.g., affects resulting from dose rates, and from different sequences of applying qualifica-1,
'j' tion test parameters) they should be accounted for in the qualification i'
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program.
y b.
The calculated operating temperature of the equipment under service ij conditions should be accounted for in thermal aging. The Arrhenius methodology y
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is considered an acceptabe method of addressing accelerated thermal aging.
Other aging methods that can be supported by. tests will be evaluated on a case-by-case basis.
c.
Known material phase changes and reactions should be identified to insure that'no adverse changes occur within the extrapolation limits.
d.
Theagingaccelerationratgand/oractivationenergiesusedduring qualification testing and the basis upon which the rate and/or activation energy was established should be defined, justified and documented.
e.
Periodic surveillance testing under normal service conditions is not considered an acceptable method for on going qualification, unless the testing includes provisions for subjecting the equipment to the limiting service environ-ment conditions (specified in 5 50.49(c) of 10 CFR Part 50).
f.
Humidity effects should be included in accelerated aging unless it can be shown that the effects of relative humidity are negligible.
j g.
The qualified life of the equipment (and/or component as applicable) and the basis for its selection should be defined and documented.
i h.
Qualified life should be established on the basis of the severity of the testing performed, the conservatisms employed in the extrapolation of data, i
the operating history, and in other methods that may be reasonably assumed.
All assumptions should be documented.
(i) An ongoing program to review surveillance and maintenance records to identify potential age-related degradations should be established.
(j) A component maintenance and replacement schedule, which include con-sideration of aging characteristics of the installed components, should be established.
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7.
Sections 6.4 and 6.5 of IEEE Std. 323-1974 discuss qualification by operating experience and by analysis respectively.
The adequacy of these methods should be evaluated on the basis of the quality and detail of the information available in support of the assumptions made.
Operating experi-ence and analysis based on test data may be used where testing is precluded by physical size of the equipment or state of the art of testing. When the analysis
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method is employed because of the physical size of the equipment, tests on vital components of the equipment should be provided.
8.
Components which are part of equipment qualified as an assembly (e.g.,
a motor starter which is part of a motor control center qualified as a whole) may be replaced with components of the same design.
If components of the same design are not used for replacement, the replacement component should r
be designed to meet the performance requirements and be qualified to meet the service conditions specified for the original components.
9.
Section 8 of IEEE Std. 323-1974 pertains to documentation.
In con-junction with Section 8, the documentation should include sufficient informa-l tion to address the required information identified in Appendix C.
A certifi-l cate o.f conformance by itself is not acceptable unless it is accompanied by l
information on the qualification program, including test data or comparable I
l test data from equivalent equipment.
A record of the qualification shall be maintained in a central file to permit verification that each item of electric i
j equipment important to safety is qualified for its application and meets its i
specified performance requirements when subjected to the conditions present
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when it must perform its safety function up to the end of its qualified life.
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IMPLEMENTATION The purpose of this section is to provide information to applicants and
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licensees regarding the NRC staff's plans for using this regulatory guide.
All operating plants and plants which have not received an operating license should meet the provisions of this guide subject to the following:
i (1) For plants which are not committed to either IEEE Std 323-1971 or the November 1974 issue of Regulatory Guide 1.89 (IEEE Std 323-1974) and have been
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tested for only high temperature, pressure and steam, equipment may not need to i
be tested.again to include other service conditions such as radiation and
.. g chemical sprays.
The qualification of equipment for these service conditions may be demonstrated by analysis.
(2) Regarding aging considerations in equipment qualification, for all plants which are not committed to the November 1974 issue of Regulatory Guide 1.89 (IEEE Std 323-1974), a specific qualified life need not be demon-strated.
This position does not, however, exclude equipment using materials that have been identified as being susceptible to significant degradation due to aging.
Component maintenance or replacement schedules should include con-siderations of the specific aging characteristics of the component materials.
Ongoing programs should exist at the plant to review surveillance and main-tenance records to assure that equipment which is exhibiting age-related degradation will be identified and replaces as necessary.
However, plants which are committed to Regulatory Guide 1.73 (IEEE Std 382-1972) and Regulatory l,
Guide 1.40 (IEEE Std 334-1971) should preage the valve operators and the motors.
(3) Beginning with May 23, 1980, replacement components or spare parts used to replace presently installed equipment or components should be qualified
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to the existing standards unless there are sound reasons to the contrary.
Non-
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in stock or was purchased prior to May'23, 1980 are among the factors to be kj considered in weighing whether there are sound reasons to the contrary.
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I' VALUE/ IMPACT STATEMENT FOR REVISION 1 TO REGULATORY GUIDE 1.89, " ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY FOR LIGHT-WATER-COOLED j
NUCLEAR POWER PLANTS"
-i
Background
Regulatory Guide 1.89, " Qualification of Class 1E Equipment for Nuclear i
Power Plants," is being revised to reflect current staff position on equipment
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qualification.
NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-
'1 Related Electric Equipment," was issued for public comment in December 1979.
Subsequent to its issuance for coment, the Commissioners (see " Memorandum and Order", dated May 23,1980) directed the staff to use NUREG-0588 along with
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" DOR Guidelines for Evaluating Qualification of Class lE Electrical Equipment in Operating Reactors," as requirements licensees and applicants must met in order to satisfy the equipment qualification requirements of 10 CFR Part 50, s
Additionally, the Commissioners directed the staff to develop a rule for electric equipment qualification. The rule will be based principally on NUREG-0588 and the 00R guidelines. This revision to Regulatory Guide 1.89 will provide guide-lines for meeting the Commission's equipment qualification rule and is essentially equivalent to the staff position and guidance contained in the proposed revised version of NUREG-0588 which is based on consideration of public comments and lessons learned from TMI-2 in source term definition.
Substantiv'e Changes and Their Value/Imoact 1.
Regulatory Position C.2 (which provided radiological source tenns for equipment qualification tests) was deleted and the following positions were added:
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.] i (a) New Pesition C.2 was added which provides the staff position regarding the various qualification methods (e.g., test, operating experience, analy-sis, on-going. qualification). Testing should be the primary method.. The other methods, when used, should be supported by test data.
'I (b) Position C.3 was added which provides the staff position pertaining to establishing performance and environmental requirements for equipment qual-
.j c4 ification. Methods for establishing temperature and pressure profiles for
- 4 zj loss-of-coolant accident and main steam line break are provided, and a
radiological source tenns are given.
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(c) Position C.4 was added which provides the staff position pertaining n
j to test procedures.
(d) Position C.5 was added which provides the staff position regarding establishing margin in testing requirements.
(e) Position C.6 was added which provides the staff position regarding accelerated aging of equipment as part of the testing procedure.
(f). Position C.7 was added which provides the staff position regarding the use of operating experience and analysis as qualification methods.
.(g) Position C.8 was added which provides the staff position on the use
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i of and qualification of, replacement components.
(h) Position C.9 was added which provides the staff position on the ade-quacy of the documentation of equipment qualification procedures and results.
Value - Tne above provides the staff's position on individual sections of IEEE Std 232-1974. This provides guidance to licensees and applicants using the standard as to what is an acceptable understanding of the standard's re-quirements. These positions should enhance the licensing process.
Impact - The impact could be considerable since the scope of the guide has
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been expanded to include additional equipment.
(Class 1E is only a subset of
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equipment important to safety.) The total impact will depend upon the final amount of additional equipment not previously qualified in accordance with Class 1E
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equipment qualification requirements. The impact of the various positions of the guide should be minimal. The positions merely take established NRC pro-visions and relate them.to appropriate sections of an endorsed voluntary con-sensus standard.
2.
Position C.3.d(3) was added which is not part of NUREG-0588, but which provides a source term for cse in the qualification of certain accident-monitoring instrumentation specified in Regulatory Guide 1.97.
This certain instrumen-4 tation is for the measurement of designated variables whose maximum value extends
+
beyond the values predicted in the design basis accident analysis.
'Value - A source term is provided which will standardize the radiation value for use in qualification of high-level instrumentation specified in Regulatory Guide 1.97, and will eliminate the necessity of source term determination on a case-by-case basis.
This will enhance the licensing process.
Impact - There is no impact. The source term of Position C.3.d(3) is not imposed by this regulatory guide.
It mercly provides an acceptable term im-posed by the provisions of Regulatory Guide 1.97, which is already in effect.
3.
The Implementation Section was modified to provide that the revision to the guide implemented in accordance with the implementation of NUREG-0588 and the D0R Guidelines.
Value - The implementation is consistent with current requirements as imposed by the Commissioners " Memorandum and Order" dated May 23, 1980.
Imoact - There is no impact since no new requirements are imposed.
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9 APPDiDIX A HETHODS FOR CALCUI.ATING MASS AND ENERGY REIIASE E
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APPENDIX A e
METHODS FOR CALCULATING MASS AND ENERGY REIIASE q
' Acceptable methods for calculating the mass and energy release to determine the loss-of-coolant accident (LOCA) environment for PWR and BWR plants are described in the following:
(1) Topical Report WCAP-8312A for Westinghouse plants.
(2)
Section 6.2.1 of CESSAR System 80 PSAR for Combustion Engineering plants.
(3) Appendix 6A of B-SAR-205 for Babcock & Wilcox plants.
(4) a. NEDO-10320'and Supplements 1 & 2 for General Electric plants.
- b. NEDO-20533 dated June 1974 and Supplement 1 dated August 1975 (GE Mark III).
Acceptable methods for calculating the mass and energy release to determine the main steam line break (MSEB) environment are described in the following:
(1) Appendix 6B of CESSAR System 80 PSAR for Combustion Engineering plants.
j (2) Section 15.1.14 of B-SAR-205 for Babcock & Wilcox plants.
(3) Same as item (4) above for General Electric plants.
(4) Topical Report WCAP-8822 for Westinghouse plants.
(Although this Topical Report is currently under review, the use of this method is l
acceptable in the interim if no entrainment is assumed. Reanalysis may be required following the NRC staff r_ view of the entrainment model as presently described.)
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APPENDIX B SAMPLE CALCULATION AND TYPE METHODOLOGY FOR RADIATION QUALIFICATION DOSE e
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e APPENDIX B SAMPLE CALCULATION AND TYPE METHODOLOGY FOR RADIATION QUALIFICATION DOSE This appendix illustrates the staff model for calculating dose rates and integrated doses for equipment qualification purposes.
The doses shown in Figure B-1 below include contributions from several dose point locations in the containment and cover a period of one year following the postulated fis-sion product release. The dose values shown here are provided for illustra-tion only and may not be appropriate for plant specific application for equip-ment qualification levels.
The dose levels intended for qualification purposes should be determined using the maximum time the equipment is intended to func-tion which, for the design basis LOCA event, may well exceed one year.
The beta and gamma integrated doses presented in Tables B-1 and B-2 and Figure B-1 below, have been determined using models and assumptions consistent with those of Regulatory Guide 1.7.
This anafysis is conservative, and fac-tors in the important time-dependent phenomena related.to the action of engi-neered safety features (ESFs) and natural phenomena, such as iodine plate-out, as done in previous staff analyses.
Doses were calculated for two points inside the containment; at the mid-point of the containment (taking sprays and plate out mechanisms into account),
and near the surface of the sump water.
The doses presented in Figure B-1 are l
values for a PWR plant having a containment free-volume of 2.5 million cubic l
feet and a power rating of 4100 W t.
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1.0 Basic Assumptions Used in the Analysis Gamma and beta doses and dose rates were determined for three types of radioactive source distributions:
(1) from activity suspended in the contain-ment atmosphere, (2) from activity plated out on containment surfaces, and (3) from activity mixed in the containment sump water.
Thus, a given piece of equipment may receive a dose contribution from any or all of these sources.
The amount of dose contributed by each of these sources is determined by the location of the equipment, the time-dependent and~ location-dependent distribu-tion of the source, and the effects of shielding.
g, Following the Three Mile Island Unit 2 (TMI-2) accident, the staff con-cluded that a thorough examination of the source term assumptions for equip-ment qualification was warranted.
It is recognized, however, that the TMI-2 accident represents only one of a number of possible accident sequences lead-i.ng to a release.of fission products, and that the mix of fission products released under various core conditions could vary substantially.. Current rulemaking proceedings are reevaluating plant siting policy, degraded cores, minimum requirements for engineered safety features and emergency preparedness.
These rulemaking activities also included an examination of fission product I
releases under degraded core conditions. While the final resolution of the source term assumptions is conditioned on the completion of these rulemaking L
efforts, the staff believes it is prudent to inc'orporate the knowledge gained l
l' of fission product behavior from the TMI-2 accident in defining source term assumptions for equipment qualification.
i Based upon release estimates in the Rogovin Report (Ref. 1), the staff assumptions for noble gas and iodine releases are still conservative.
- However, the report estimates that thb TMI-2 release contained between 40 and 60 percent B-2
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of the Cs-134 and Cs-137 core activity in the primary system water, in the con-tainment sump water, and in'the auxiliary building tank.
Comparison of the j
integrated dose from the TMI-2 cesium release to the previous staff assumption of "1% solids" shows that the "1% solids" assumption may not be conservative for ch equipment required to function for time perio'ds exceeding thirty days.
The' staff feels that as a first step toward modification of the TID-14844 source term in
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the direction indicated by the TMI-2 experience, it may be prudent to factor i
-j in a cesium release'in addition to the previously assumed "1% solids." As a result, the revised regulatory positions propose a cesium release of 50 percent of the core activity inventory (see Positions C.3.d(1) and (2).
The assumed cesium release implies no substantial departure from, and is consistent with, the degraded core conditions previously implied by the assumed release of a 50 percent core iodine activity.
This change in assumptions would have particular significance for the qualification of equipment in the vicinity of recirculating-fluids and for equipment required to function for time periods exceeding 30 days.
The assumption of" concurrent release of cesium and iodine also is consistent 4
with the findings of recent source term studies reported in NUREG-0772 (Ref. 2).
This report also concluded that the expected predominant form of iodine released during accidents is cesium iodide (CsI).
Although the CsI form'is not specifically addressed in this report, it is evident that either CsI, or It and.Cs would, in the long term, be located primarily in the reactor water and the containment sump water (PWR) or suppression pool (BWR).
The staff recognizes that the revised source terms contained in this report are interim values and that the conclusions from the report cited above, as well as further results from current research efforts in the source term area, should ultimately form the basis for any revision of source term assumptions.
Any revision of the source term assumptions, such as l
the incorporation of additional radionuclides, would be factored into the guide l~
j before it is issued as an effective' guide.
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2.0 Assumptions Used in Calculatina Fission Produce Concentrations
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This section discusses the assumptions used to simulate the PWR and BWR
- l containments for determining the time-dependent and location-dependent dis-d tributionofnoblegasesandiodiocEairbornewithinthecontainmentatmos-
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-j phere, plated out on containment surfaces, and in the sump water.
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The staff has developed a computer program, TACT, (to be published) that
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models the time-dependent behavior of iodine and noble gases within a nuclear power plant. 'The DCT code is used routinely by the staff for the calculation A
pj of the offsite radiological consequences of a LOCA, and is an acceptable method 1
i for modeling the transfer of activity from one containment region to another 1,,
-l and in modeling the reduction of activity due to the action of ESFs. Another l
staff code, SPIRT (Ref. 3), is used to calculate the removal rates of elemental F
iodine by plate-out and sprays.
These' codes were used to develop the source ters estimates.
The folicwing assumptions were also used to calculat,e the distribution of radioactivity within the containment following a design basis LOCA.
2.1 PWR Dry Containments s
j The source terms used in the analysis assumes that 50 percent of the a.
jo core iodines and 100 percent of the core noble gases were released instantaneously to the containment atmosphere, 50 percent of the core \\
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!.j inventory of cesium and 1% of the. remaining " solid" activity inventory e
is released from the core and carried with the primary coolant directly i
s to the containment sump.
(Note: 'The integrated dose from a'"1% solids",
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release of TID 14844 is approximately equal to the M egrated dose from 50 percent cesium release for the initial 30-day period.)
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The containment free volume was taken as 2.52 x 106 3
ft. Of this 6
3 volume, 74 percent or 1.86 x 10 ft is assumed to be directly
,y covered by the containment sprays.
(Plants with different contain-ment free volumes should use plant specific valves.)
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c.
6.6 x 10 ft of the containment free volume is assumed unsprayed, which includes regions within the main containment space under the
[j containment dome and compartments below the operating floor level.
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d.
The ESF fans are assumed to heve a design flow rate of 220,000 cfm
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in the post-LOCA environment. Mixing between all major unsprayed
's regions and compartments and the main sprayed region is assured.
Air exchange between the sprayed and unsprayed region was assumed to e.
be one-half of the design flow rate of ESF fans. ' Good mixing of the k.
containment activity between the sprayed and unsprayed regions is
. assured by natural convection currents and ESF fans.
f.
The' containment spray system was assumed to have two equal capacity trains, each designed to inject 3000 gpm of boric acid solution into the containment.
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Trace levels of hydrazine was assumed added to enhance the removal of iodine.
h.
The spray removal rate constant (lambda) was assumed and calculated usin( ^Je staff's SPIRT program, conservatively assuming only one B-5
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spray train operation and an elemental iodine instantaneous parti-tion coefficient (H) of 5000.
The calculated value of the elemental
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iodine spray removal constant was 27.2 hr
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Plate-out of iodine on containment internal surfaces was modeled as i
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-a first-order rate removal process and best estimates for model parameters were assumed.
Based on an assumed total surface area 5
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'i within containment of approximately 5.0 x 10 fg.
The calculated iIl value for the overall elemental iodine plate-out constant was
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1.23 hr The assumption that 50 percent of the activity is ll instantaneously plated-out should not be used.
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The spray removal and plate-out process were modeled as competing iodine removal mechanisms.
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A spray removal rate constant (A) for particulate iodine concentra-tion was calculated using the staff's SPIRT program (Ref. 2).
The
-1 staff calculated a value of A = 0.43 hr and allowed the removal of particulate iodine to continue until the airborne concent. ration was i;
i 4
reduced by a factor of 10.
The organic iodine concentration in the containment atmosphere is assumed not to be affected by either the containment spray or pl. ate-out removai mechanisms.
1.
The sprays were assumed to remove elemental iodine until the instan-f taneous concentration in the sprayed region was reduced by a factor of 200.
This is necessary to achieve an equilibrium airborne iodine concentration consistent with previous LOCA analyses.
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A relatively open (not compartmented) containment was assumed, and the large release was uniformly distributed in the containment.
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This is an adequate simplification for dose assessment in a PWR con-tainment, and realistic in terms of specifying the time-dependent radiation environment in most areas of the containment.
n.
The analysis assumed that more than one species of radioactive J
iodine is present in a design basis LOCA.
The calculation of the 1
post-LOCA environment assumed that 2.5 percent of the core inventory t
of the iodine released is associated with airborne particulate mate-rials and 2 percent of the core inventory of the iodine released formed organic compo'unds.
The remaining 95.5 percent remained as elemental iodine.
For conservatism this composition was assumed present at time t = o.
(These assumptions concerning the iodine form are consistent with those of Regulatory Guides 1.3 and 1.4 when a plate-out factor of 2 is assumed for the elemental form.)
o.
For all containments, no leakage from the containment building to the environment was assumed.
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Removal of airborne activity by engineered safety features may be assumed when calculating the radiation environment following other non-LOCA design basis accidents provided the safety features systems are automatically activated as a result of the accident.
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g 2.2 PWR Ice Condenser Containments The assumptions and methods presented for the calculation of the radia-tion environment in PWR dry containments are appropriate for use in calculat-ing the radiation environment following a design basis LOCA for ice condenser
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containments with the following modifications:
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The source should be assumed to be initially released to the lower 1i containment compartment. The distribution of the activity should be i
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based on the forced recirculation fan flow rates and the transfer 1
rates through the ice beds as a function of time.
t b.
Credit may be taken for iodine removal via the operation of the ice beds and the spray system.
A time-dependent removal efficiency con-sistent with the steam / air mixture for elemental iodine may be assumed.
c.
Removal of airborne iodine in the upper compartment of the contain-ment by the action of both plate-out and spray processes may be assumed provided that these removal processes are evaluated using the ' assumptions consistent with items h through 1 in Section 2.1 above and plant-specific parameters.
2.3 BWR Containments The assumptions and methods presented for the calculation of the radia-tion environment in PWR dry containments are appropriate for use in calculat-ing the radiation environment following a design basis LOCA for BWR's with the following modifications:
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A decontamination factor (DF) of 10 should be assumed for both the-
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elemental and particulate iodine as the iodine activity passes
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through the suppression pool.
No credit should be taken for the
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removal of organic iodine or noble gases in the suppression pool.-
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b.
For Mark III designs, all of the activity passing through the suppres-
.i sion pool should be assumed instantaneously and uniformly distributed within the containment.
For the Mark I and Mark II designs, all of the activity should be assumed initially released to the dry well j
area and the. transfer of activity from these regions via containment 1
leakage to the surrounding reactor building volume should be used to predict the qualification levels within the reactor building (secondary containment).
c.
Removal of airborne iodine in the dry well or reactor building by both the plate-out and the spray process may be assumed provided the effectiveness of these competing iodine removal processes are evaluated using the assumptions consistent with items h through 1 in Section 2.1 above and plant-specific parameters.
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d.
The removal of airborne activity from the reactor building by opera-tion of the Standby Gas Treatment System (SGT$) may be assumed.
3.0 Moos. for Calculating the Dose Rate of Airborne and Plate-out Fission Products The beta and gamma dose rates and integrated doses from the airborne activity within the containment atmosphere were calculated for a midpoint in B-9 n
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the containment.
The containment was modeled as a cylinder of equal height and diameter.
Containment shielding and internal structures were neglected.
4 Because of the short range of the betas in air, the airborne beta doses were calcualted using an infinite medium approximation.
This is shown in i
Reference 4 to result in only a small error.
The airborne beta doses are not 9
expected to be significantly reduced by the presence of containment internal.
structures. For beta dose calculations for equipment located on the coatain-ment walls or on large internal structures, the semi-infinite beta dose model may be used.
l' The gamma dose rate contribution from the plated-out iodine on containment surfaces to the point on the centerline was also included. The.model calcu-I lated the plate-out activity in the containment assuming only one spray train and one ventilation system were operating.
It should be noted that wash-off f
by the sprays of the plated-out iodine activity was not addressed in this evaluation.
Finally, all gamma doses were multiplied by a correction factor of 1.3 as suggested in Reference 4 to account for the omission of the contribution from j
the decay chains of the isotopes.
4.0 Model for Calculating the Dose Rate of Sump Fission Products The staff model assumed the washout of airborne iodine from the contain-ment atmosphere to the containment sump.
For a PWR containment with sprays and good mixing between the sprayed and unsprayed regions, the elemental
' t iodine (assumed constituting 91 percent of the released iodine) is very rapidly washed out of the atmosphere to the containment sump (typically, 90 percent of the airborne iodine in less than 15 minutes).
4 B-10
The dose calculations assumed a time-dependent iodine source.
(The differ-ence between the integrated dose assuming 50 percent of the core iodine imme-diately available in the sump versus a time-dependent sump iodine buildup is notsignificant.)
The " solid" fission products should be assumed instantaneously carried by the coolant to the sump and uniformly distributed in the sump water. The gamma 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 rates should be calculated including an estimate of the effects of buildup.
5.0 Conclusion The values given in Tables B-1 and B-2 and Figure B-1 for the various locations in the containment provide an estimate of expected radiation qualifi-cation values for a 4100 MWt PWR design.
The NRC Office of Research is continuing its research efforts in the area of source terms for equipment qualification following design basis accidents.
As more information in this area becomes available, the source terms and staff models may change to reflect the new information.
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TABLE B-1 SUMfARY TABLE OF ESTIMATES FOR TOTAL AIRBORNE GAMfA DOSE CONTRIBUTCRS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER TIME AIRBORNE IODINE AIRBORNE NOBLE GAS PLATE-0UT IODINE TOTAL DOSE r
(HRS)
DOSE (R)
DOSE (R)
DOSE (R)
(R) 0.00 0.03 4.82+4 7.42E+4 1.69+3 1.24E+5 0.06 8.57+4 1.39E+5 3.98+3 2.29E+5 0.09 1.09+5 1.98E+5 7.22+3 3.14E+5 0.12 1.25+5 2.51E+5 1.10+4 3.87E+5 j
0.15 1.38+5 3.01E+5 1.52+4 4.54E+5 0.18 1.47+5 3.48E+5 1.96+4 5.15E+5 0.21 1.55+5 3.92E+5 2.41+4 5.71E+5 0.25 1.64+5 4.49E+5 3.03+4 6.43E+5 0.38 1.87+5 6.19E+5 5.05+4 8.57E+5 0.50 2.03+5 7.61E+5 6.90+4 1.03E+6 0.75 2.36+5 1.03E+6 1.06+5 1.37E+6 1.00 2.66+5 1.26E+6 1.40+5 1.67E+6 2.00 3.62+5 2.04E+6 2.61+5 2.66E+6 5.00 5.50+5 3.56E+6 5.40+5 4.65E+6 8.00 6.63+5 4.38E+6 7.47+5 5.79E+6 24.0 1.01+&
6.26E+6 1.45+6 8.72E+6 60.0 1.31+6 7.16E+6 2.10+6 1.06E+7 96.0
'1.45+6 7.56E+6 2.39+6 1.14E+7 192.
1.68+6 8.29E+6 2.86+6 1.28E+7 298.
1.85+6 8.76E+6 3.19+6 1.38E+7 394.
1.95+6 8.85E+6 3.41+6 1.42E+7 560.
2.07+6 9.06E'6 3.64+6 1.48E+7 720.
2.13+6 9.15E+6 3.76+6 1.50E+7 i
888.
2.16+6 9.19E+6 3.83+6 1.52E+7 1060 2.18+6 9.21E+6 3.87+6 1.53E+6 1220 2.19+6 9.21E+6 3.89+6 1.53E+7 1390 2.20+6 9.21E+6 3.90+6 1.53E+7 1560 2.20E+6 9.22E+6 3.91+6 1.53E+7 1730 2.20E+6 9.22E+6 3.91+6 1.53E+7 1900 2.20E+6 9.22E+6 3.92+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 9.26E+6 3.92E+6 1.54E+7 8000 2.20E+6 9.27E+6 3.92E+6 1.54E+7 8710 2.20+6 9.28E+6 3.92+6 1.54E+7 TOTAL 1.54E+7 B-12
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SUMMARY
TABLE OF ESTIMATES FOR TOTAL AIRBORNE BETA DOSE CONTRIBUTORS
-i IN CONTAIhTENT TO A POINT IN THE CONTAINMENT CENTER
,.j TIME AIRBORNE IODINE AIRBORNE NOBLE GAS TOTAL DOSE j
(HRS)
DOSE (RADS)+
DOSE (RADS)+
(llADS)+
1'j 0.00 1
0.03 1.47+5 5.48+5 6.95+5 1
0.06 2.62+5 9.86+5 1.25+6 0.09 3.33+5 1.35+5 1.68+6 0.12 3.83+5 1.65+6 2.03+6
- 0. 15 4.20+5 1.91+6 2.33+6 j
0.18 4.49+5 2.14+6-2.59+6 i
0.21 4.73+5 2.35+6 2.82+6 0.25 5.00+5 2.60+6 3.10+6
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0.38 5.67+5 3.30+6 3.87+6 0.50 6.15+5 3.86+6 4.48+6 0.75 7.13+5 4.89+6 5.60+6
-l 1.00 8.00+5 5.81+6 6.61+6 2.00 1.07+6 9.02+6 1.01+7 5.00
- 1. 58 +6 1.65+7 1.81+7 8.00 1.88+6 2.20+7 2.39+7 24.0 2.87+6 4.08+7 4.37+8 60.0 3.89+6 6.15+7 6.54+7 96.0 4.37+6 7.48+7 7.92+7 192.
5.14+6 1.00+8 1.05+8 298.
5.64+6 1.17+8 1.23+8 394.
5.99+6 1.25+8 1.31+8 560.
6.34+6 1.34+8 1.40+8 720.
6.53+6 1.39+8 1.46+8 888.
6.63+6 1.42+8 1.49+8
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1060 6.69+6 1.44+8 1.51+8 1220-6.73+6 1.45+8 1.52+8 l
1390 6.75+6 1.47+8 1.54+8 1560 6.76t6 1.49+8 1.56+8 1730 6.76+6 1.51+8 1.58+8 1900 6.76+6 1.52+8 1.59+&
L, 2060 6.76+6 1.54+8 1.61+8 2230 6.77+6 1.55+8 1.62+8 2950 6.77+6 1.62+8 1.69+8 3670 6.77+6 1.69+8 1.76+8 4390 6.77+6 1.76+8 1.83+8 5110 6.77+6 1.83+8 1.90+8 li 5830 6.77+6 1.89+8 1.96+8 ll 6550 6.77+6 1.96+8 2.03+8 j,
7270 6.77+6 2.03+8 2.10+8 8000 6.77+6 2.09+8 2.16+8 h
8710 6.77+6 2.16+8 2.23+8 TOTAL 2.23+8 i!*
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- j REFERENCES Q
1.
Mitchell Rogovin, George T. Frampton, Jr.,- et al, "Three Mile Island--
4 a report to the Commissioners and to the Public" NUREG/CR-1250, Volume II, h
Part 2.
2.
NUREG-0772, " Technical Basis for Estimating Fission Product Behavior During LWR Accidents."
3.
A. K. Postma and P. Tam, " Technological Bases for Models of Spray Wash-out and Airborne Contaminants in Containment Vessels," USNRC Report NUREG/CR-0009, November 1978.
Available for purchase from National Tech-nical Information Service, Springfield, Virginia 22161.
4.
M. J. Kolar and N. C. Olson, " Calculation of Accident Doses to Equipment Inside Containment of Power Reactors," Vol. 22, pp. 808-809 in Transac-i tions of the American Nuclear Society, 1975. Available from technical libraries.
BIBLIOGRAPHY A.K. Postma and R. Zavadoski, " Review of Organic Iodide Formation Under Accident Conditions in Water Cooled Reactors," WASH-1233, October 1972, pp. 62-64.
Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
E. A. Warman and E. T. Boulette, " Engineering Evaluation of Radiation Environment in LWR Containments," Vol. 23, pp. 604-605 in Transactions of the American Nuclear i
Society, 1976.
Available from technical libraries.
D. C. Kocher, ed., " Nuclear Decay Data for Radionuclides Occurring in Routine Releases froc Nuclear Fuel Cycle Facilities," ORNL/NUREG/TM-102, August 1977.
Available for. purchase from the National Technical Information Service, Springfield, Virginia 22161.
E. Normand and W. R. Determan, "A Simple Algorithm to Calculate the Immersion Dose," Vol. 18, pp. 358-359 in Transactions of the American Nuclear Society, 1974. Available from technical libraries.
R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, " Fission Product Source Terms for the LWR Loss-of-Coolant Accident:
Summary Report," USNRC Report NUREG/CR-0091, May 1978.
Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
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APPEIDII C E
QUALIFICATION DOCUMENTATION FOR ELECTRIC EQUIPMENT DTORTANT TO SAFETT e
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APPENDIX C QUALIFICATION DOCmfENTATION FOR ELECIRIC EQUIPMENT EfPORTANT TO SAFEIT In order to ensure that an environmental qualification program conforms with General Design Criteria 1, 2, 4 and 23 of Appedir A and Sections III and II of Appendiz 3 to 10 CFR Part 50, and to the national standards mentioned in Part II " Acceptance criteria" (which includes IZZZ Std. 323) con *=ined in Senad=rd Review Plan Section 3.11, the following information on the qns14 f4 cation program is required for all electric equipment important to safecy.
' e 1.
Identify all electric equipment.important to safety and provide the following:
a.
Type (functional designation)
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b.
Manufacturer c.
Mannfacturer's type number and model number i
d.
The equipment should include the following, as applicable:
j (1) Switchgear (2) Motor control centers e
ll (3) Valve operators (4) Motors (5) Iogic equipment (6) Cable (7) Diesel generator control equipment (81 Sensors (pressure, pressure differential, temperature and neutron)
(9) Limit switches.
(,
(10) Heaters (11) Fans (12) Control boards (13) Instrument racks and panels (14) Connectors (15) Electrical penetrations (16) Splices (17)'Termisal blocks 2.
Categorize the equipment identified in item 1 above into one of the following categories:
a.
Equipment that will experience the envirec= ental conditions of design basis accidents for,rhich it must function to sitigate said accidents, and that will be qualified to demonstrate operability in the accident environment for the time requi:=d for accident nitigation with safety sargin to failure.
4 l-b.
Iquipment that will experie=ce environmental conditions of design j.
basis accidents through which it need not function for mitigation of said accidents, but through which it =ust not fail in a = ann =:
detri= ental to plant safety or accident mitigation, and that will be qualified to denenstrate the capability to withstand any accidant l
environment for the ti=e during which it must not fail with safety l
margin to failure.
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l Equipment that win experience environ = ental conditions of design c.
basis accidents through which it need not function for mitigation of j
said accidents, and whose failure (in any mode) is deemed not detrimental to plant safety or accident mitigation, and need not be qualified for any accident environment, but will be qualified for its non-accident service environment.
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d.
Equipment that win not experience environmental conditions of design basis accidents and that will be qualified to demonstrate operability under the expected extremes of its non-accident service environment. This equipment would normally be located outside the reactor containment.
3.
For each type of equipment in the categories of equipment listed in item 2 above, provide separately the equipment design specification requirements, including:.
a.
The system safety function requirements.
b.
An environmental envelope as a function of time that includes all extreme parameters, both maximum and m%um values, expected to
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occur during plant shutdown, normal operation, abnormal operation, and any design basis event (including 10CA and MS13), including post-event conditions.
Time required to fulfill its safety function when subjected to any c.
of the extremes of the envirocment envelope specified above.
d.
Technical bases should be provided to justify the placement of each type equipment in the categories 2.b and 2.c listed above.
4 Provide the qualification test plan, test setup, test procedures, and acceptance cr!.teria. for at least one cf each group of equipment of item 1.d as appropriate to the category identified in item 2 ybove.
If any method other than type testing was used for qualification (operating experience, analysis, combined qualification, or ongoing qualification),
t describe the method in sufficient detail to permit evaluation of its adequacy.
5.
For each category of equipment identified in item 2 above, state the actual qualification envelope simulated during testing (defining the duration of the hostile environ =ent and the margin in excess of the design requirements). If any method other than type testing was used for qualification, identify the method and define the equivalent
" qualification envelope" so derived.
.i 6.
A su= mary of test results that demonst ates the adequacy of the qualification program. If analysis is used for qualifiestion, justification of all analysis assu=ptions must be provided.
7.
Identification of the qualification docu=ents which contain detailed supporting information, including test data, for items 4, 5, and 6.
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