ML20027D800
| ML20027D800 | |
| Person / Time | |
|---|---|
| Issue date: | 08/07/1981 |
| 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 8211100081 | |
| Download: ML20027D800 (53) | |
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ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY FOR 6 NUCLEAR POWER PLANTS 1
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, by 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 $50.49 " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," to 10 CFR Part 50, requires that each type l
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 important to safety.
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A11 8211100081 821014 PDR FOIA CURRAN 82-426 PDR
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1 B.
DISCUSSION b
i IEEE Std 323-1974, "IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations,"E ated February 28,1974,.was prepared by d
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Subcommittee 7, Equipment Qualification, of the Nuclear Power Engineering Committee of the Institute of Electrical and Electronics Engineers, Inc. (IEEE),
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and subsequently was approved by the IEEE Standards Board on December 13, 1973.
The standard describes basic procedures for qualifying Class 1E equipment and 1
interfaces that are to be used in nuclear power plants and components or equipment j
of any interface who:e 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-i 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-l ing are examples of ca,nsiderations 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
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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 radiatica; (3) equipment required to 1/ Copies may oe obtained from the Institute of Electrical and Electronics Engineers, Inc., United Engineering Center, 345 East 47th Street, New York, i
New York 10017.
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initiate protective action would generally be required for a shorter period of time than instrumentation required to follow the course of an accident. The specific environment for which individual equipment must be qualified will depend 5
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 te completion of its intended func-l*
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.
SEM 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 establisi:ed, including minimizing dust accumulation that could degrade the ability of the equipment to function properly.
C.
REGULATORY POSITION The procedures described by IEEE Std 323-1974, "IEEE Standard for Quali-fying Class IE Equipment for Nuclear Power Generating Stations,"E 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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Regulatory Position 3.d(12) of this guide addresses qualification of equip-4 ment exposed to low-level radiation doses. Numerous studies have compiled rad-iation effects data on all classes of organic compounds and show that the least radiation resistant compounds have damage thresholds greater than 10'+ rads and would remain functional with exposures somewhat above the threshold value. Thus, for organic materials, radiation qualification may be readily justified by existing test data and/or operating experience for radiation exoosures below 10
lrads. However, for electronic components, studies have shown failures in metal-
{ oxide-semi-conductordevicesat3.5x10 3 rads. Therefore, radiation qualification i
for electronic components may likely have a lower e.xposure threshold,
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1.
Section 50.49, " Environmental and Seismic Qualification of Electric Equipment for Nuclear Pcwer Plants," to 10 CFR Part 50 requires that essential electric systems and equipment be qualified to perfonn their intended functions.
Typical essential systems and equipment which mitigate accidents are listed in Appendix E of this guide. Additional equipment which, if it malfunctioned or failed when subjected to accident or seismic conditions, could negate the t-safety function of essential systems and equipment should also be qualified for the accident and seismic conditions. An example of such additional equip-ment is the rod control system which could, if it were to fail due to an acci-dent or seismic event, cause the reacto. shutdown system to fail to prevent fuel damage from exceeding precribed limits. Other examoles of Additional equipment are the steam generator level control which cou?d fail such that it would cause the loss of sufficient cooling for the reactor core, and the pres-
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surizer pressure control which could fail such that core cooling would be in-adequate. Associated circuits, as defined in Regulatorf Guide 1.75, should also be qualified since the failure of associated circuits could cause a fail-ure of essential systems and equipment.
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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-bility or acceptability ~of IEEE Std 344 is covered in Regulatory Guide 1.100.
h.
Section 5 of IEEE Std 323-1974 pertains to principles of qualifica-tions including various methods.
In conjunction with Secticn 5, the selection of a qualification method should be based on the following generaHy Jry feu. of esNn_9 a.
The NRC will not accept analysis den: ithe t _ cuppcrting tat dh 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 (s) testing of the equipment ar' is impractical due to size limitations
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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.
Temoerature 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-The calculations should account for lease rates are summarized in Appendix A.ke-rsy,1Ae fech ( c 1, erg,p's%Ller,
& e.>campfe. when cons thetimedependenceandspat'ialdistributionofthesevariables.%ighpressure le A
huyle not necessarily a. limiting condition.
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Pressurized Watar Reactors (PWRs)
Dry Containment --Calculate LOCA containment environment using CONTEMPT-LT or equivalent industry codes. Additional guidance is provided in Standard Review Plan (SRP) Section 6.2.1.1.A, NUREG-75/087.
Ice Condenser Containment - Calculate LOCA containment environment using R
LOTIC or equivalent industry codes. Additional guidance is provided in i
SRP Section 6.2.1.1.B, NUREG-75/087.
Boiling Water Reactors (BWRs)
Mark I. II and III Containment - Calculate LOCA environment using methods of GESSAR Appendix 38 or equivalent industry codes. Additional guidance is prov ded in SRP Section 6.2.1.1.C, NUREG-75/087.
gegg /4ey (2)4 khe test profiles included in Appendix A to IEEE Std. 323-1974, should not be considered an acceptable alternative in lieu of using plant-specific containment temperature and pressure design profiles unless plant-specific analysis is provided to verify the applicability of those profiles.
b.
Temoerature and Pressure Conditions Inside Containment for Main Steam 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.
(1) Models' that are acceptable for calculating containment parameters are listed in Position 3.a(1).
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Spes (2)j De test profiles included in Appendix A to IEEE Std. 323-1974 should not be considered an acceptable alternative in lieu of using plant-specific S
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1 containment temperature and pressure design profiles unless plant-specific analysis is provided to verify the applicability of those profiles.
c.
Effects of Chemicals Guidelines for the chemical spray.sakshssa 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.
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
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during or following that which equipment must remain functional.
It should be accdfent assumed that the.GiWrelated 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.
Methods acceptable to the NRC staff for establishing radiation limits for qualification for BWR and PWR type reactors are provided in the sample calcula-l 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 essegueuuks LOCA should consider the most limiting environment associated with the following:
(a) For a LOCA where the break cannot be isolated,100% of the core activity inventory of noble gases and 50% of the core activity inventory of the halogens should be assumed to be instantaneously released from the fuel to Qsnn PdN the containment. Fifty percent of the cesium activity and 1% of the remaining,
/solidsfactivity inventory in the core should be assumed to be instantaneously 6
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released from the fuel to the primary coolant and carried by the coolant to the containment sump.
i (b) For a LOCA where the b,reak can be isolated, 100% of the core activity inventory of the noble gases, 50% of the core activity inventory of d"g50% of the core activity inventory of the cesium and 3% of the haloge an fdsion the remain ng # solids # activity inventory should be assumed to be instantaneously
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released (after an initial time delay) and circulated in the primary coolant system. This accident is not expected to produce instantaneous fuel damage.
A 30-minuta delay may be assumed for fission product release from the fuel.
Greater delay times should be justified on the basis of system perfermance
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design that minimizes fission product release. No noble gases should be assumed circulating in the pr.imary system following system depressurization.
(2) For all other design basis accidents (e.g., non-LOCA high energy 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 postulated, the nuclide inventory of the fuel elements breached should be calculated at the end of core life, assuming continuous full power opera-tion. The fuel rod gap inventory in the rods should be assumed to be 10% of i
l the total rod activity inventory of iodine and 10% of the total activity inven-l tory of the noble gases (except for Kr-85 for which a release of 30% should be l
assumed). All the gaseous constituents in the gaps of the breached fuel rods i
should be assumed instantaneously released to the primary coolant. When fuel 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 this case,100% of the noble gases, 50% of the halogens, 50% of the cesium 7
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inventoryand1%oftheremaining[ solids #activityinventoryintheseelements should be assumed to be instantaneously released to the primary coolant.
c/,ameh (3) For a limited number of accident monitoring instrumentation with
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4S can attain under nr.s.tase)se conditions specified in Regulatory Guide 1.97, g
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," the source term should 415e ewArboment and assume an initir; release which considers the fission product release groups 4
associated with grossly melted fuel. An acceptable assumption of frac-tional release for each group are: noble gases, 100%; I, Br, 100%; Cs, Rb, 100"c, 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 s
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 basis accidents should take into account the time-dependent transport of re-i leased fission products within various regions of containment and auxiliary structures.
(S) The initial distribution of activity within the containment should 8
r be based on 4 mechanistic i Y
' -..d assumptions. 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 8
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of the containment. The assumption of uniform distribution of activity through-out a compartmented containment at time zero may not be appropriate.
4 (6) Effects of ESF systems, such as containment sprays and containment ventilation and filtration systems, which act to remove airborne activity and redistribute activity within containment, should be calculated using the same assumptions used in the calculation of offsite dose.
See SRP Section 15.6.5 (NUREG-75/087) and the related sections referenced in the Appendices to that section.
(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 (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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(9 Shielded components need be qualified only to the gamma radiation dose.
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'i tweb required, provided an analysis or test shows that the sensitive portions l
,i of the component or equipment are not exposed to significant beta radiation 1;l dose rates or that the effects of beta radiation heating and secondary radia-3j tion have no deleterious effects on component performance.
j (10) 9mmucc. batings and coverings on electric equipment should be assumed j
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-l l-trating the containment, plus the exposure from the recirculating fluid.
L (12) Equipment that ma.y be exposed to low level radiation doses N '
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-4 should not be considered h exempt from radiation qualificationbesNg, g
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- :isme analysis supported by test data or operating experience h r-' * ' w44M Ces ga veriff that ites dose and dose rates will not degrade the operability of the 4
equipment below acceptable values.
(13) A given component may be considered to be qualified provided it can 1
I be shown that the component can be subjected, without failing, to the integrated l
beta and gamma doses, accounting for beta and gamma dose rates, which are equal 1
to or higher than those levels resulting from an analysis that (1) is similar 10 e
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Environmental Conditions for Outside containment (1) Equipment important to safety W(abesummes located outside containment 1%nC
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and M.4en could be subjected to high energy pipe breaksj as defined in the Standard Res.ew Planj should be qualified to the conditions resulting from an accident for the duration required. The techniques to calculate the environ-mental conditions should employ a plant specific model based on good engineering judgment.
/S (2) Equipment important to safety, which as located in general plant areas outside containment where equipment is not subjected to a design basis accident environment, should be qualified to the normal and abnormal range of environmental conditions postulated to occur at the equipment location.
(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.
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" specifications which contain a description of the functional requirements of its specific environmental location during normal and abnormal environmental 11
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surveillance program, in conjunction with a good preventive maintenance program, d
should be provided to assure that equipment so qualified will function for its 2
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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 shouldbejustified.
c.
Where equipment is. located in an area where rapid pressure changes are expected, qualification by test should demonstrate that, under the most adverse time dependent relative humidity condition's (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 temperatura to which equipment is qualified, whan exposed to the simulated environment, should be defined by temperature readings gs close sg was===M 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-M W
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Chemical spray or demineralized water spray should be incorporated during simulated event testing at or near the maximum pressure and tamperature conditions that would occur when the spray systems actuate.
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tored continuously to assure that spurious failures (if any) have been accounted l
for during testing.
For long-term testing, however, continuous monitoring during L periodic intervals may be used if justified.
Expected extremes in power supply voltage range and frequency should be applied appropriately during simulated event environmental testing.
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Cobalt-60 or Ces[ium-137 is-e-acceptable gamma radiation sourcejfor environmental qualification.
h.
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 indicated in Section 6.3.1.5, should be used as a guide with the following exceptions:
a.
Quantified margins should be applied to the design parameters dis-cussed in Position C.3 to assure that the postulated accident conditions have been enveloped during testing.
These margins should be applied in addition to any conservatism applied during the derivation of the specified plant param-eters unless those conservatisms can be quantified and shown to contain suffi-cient margin. TM margins should (a) account for uncertainties associated with the use of analytical techniques in deriving environmental parameters, 13 e-
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.l (e.g., when only a few units are tested (c) account for variations in the commercial production of the equipment, and (d) account for the inaccuracies in 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 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 />),
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and, once its function is complete, subsequent failures are shown not to be detrimental to plant safetyi Other equipment may not be required to perform a safety function but must not fail within a short time period into the event, and subsequent failures are also shown not to be detrimental to plant safety.
Equipment in these categories should remain functional in the accident environ '
ment for a period of at least I hour in excess of the time assumed in the acci-dent analysis.
For all other equipment (e.g., post-acci.'.ent monitoring, recem-biners, etc.), the 10 percent time margin identified in Section 6.3.1.5 of IEEE Std. 323-1974 should be used.
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Section 6.3.3 of IEEE Std. 323-1974 pertains to aging.
Inconjunc-tion with Section 6.3.3, the following should apply:
a.
Where synergistic effects have been identified, (e.g., effects l
Le combiwa/4n wh% hmhenx/we resulting from dose rates, and from different sequences of applying qualifica-A tion test parameters) they should be accounted for in the qualification program.
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b.
The ubum'asess operating temperature of the equipment under service conditions should be accounted for in thermal aging. The Arrhenius methodology 14 m
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is considered an acceptabe method of addressing accelerated thermal aging.
q Other aging methods that can be supported by tests will be evaluated on a case-1 J
by-case basis.
c.
Known material phase changes and reactions should be identified to 1
insure that no adverse changes occur within the extrapolation limits.
d.
The aging acceleration rate and/or activation energies used during 4
i qualification testing and the basis upon which the rate and/or activation energy e
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 S 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.
g.
The qualified life of the equipment (and/or component as applicable) and the basis for its selection should be defined and documented.
h.
Qualified life should be established on the basis of the severity of the testing perforired, the conservatisms employed in the extrapolation of data, 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 r^'
' 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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Sections 6.4 and 6.5 of IEEE Std. 323-1974 discuss qualification by operating experience and by analysis respectively. The adequacy of these i
j 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 1
physical size of the equipment or state of the art of testing. When the analysis method is employed because of the physical size of the equipment, tests on vital components of the equipment should be provided.
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 Nplacement, the replacement Component should be designed to meet the performance requirements and be qualified to meet the service conditions specified for the original components.
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Section 8 of IEEE Std. 323-1974 pertains to documentation. In con-junction with Section 8, the documentation should include sufficient informa-tion to address the required information identified in Appendix C.
A certifi-cate of conformance by itself is not acceptable unless it is accompanied by information on the qualification program, including test data or comparable 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 equipment important to safety is qualified for its application and meets its specified performance requirements when subjected to the conditions pre.ent when it must perform its safety function up to the end of its qualified life.
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D.
IMPLEMENTATION I
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i The purpose of this section is to provide inferination to applicants and s
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i licensees regarding the NRC staff's plans for using this regulatory guide.
G.11 operating plants and plants which have not received an operating license should meet the provisions of this guide subject to the following:
(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{andhavebeen f
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tested for only high temperature, pressure and steam, equipment may not need to be tested again to include other service conditions such as radiation and 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 197A issue of Regulatory Guide 1.89 IEEE Std 323-19747, 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 See Apperidix.0)intenance or replacement schedules should include con-to aging (g Component ma siderations of the specific aging characteristics of the component mate' rials.
Ongoing programs should exist at the plant to review surveillance and main-tenance records to assure that equipment which is exhibiting age-related#'e *
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- "'N degradation will be identified and replaces as necessary. 'Howeverl, plants hondg.
9 h h a.T committed to Regulatory Guide 1.73 IEEE Std 382-1972$ and Regulatory 4
,1 u.h4 Guide 1.40{IEEEStd 334-19711 C 'd-d % e m e 4 th: 1 (3) Beginning with May 23, 1980, replacement components or spare parts used to replace presently installed equipment or components should be qualified 17
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i a ctor'NT t 'M : :;tir.; :trtrt unless there are sound reasons to the contrary. -Non-availability and/or the fact that the component to be used as a replacement is
-4j in stock or was purchased prior to May 23, 1980 are among the factors to be a~
considered in weighing whether there are sound reasons to the contrary.
h) 1 4
1
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e APPEND O A ETHODS FOR CAI.CUI.ATING FASS AND ENERGY FrrrAsr O
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' l
' l
'4 APPINDII A
. -f METHODS 70R CALCULATING MASS AND ENERGY F " tSE a
4
' Acceptable methods for calculating the mass and energy release to determine
'i the loss-of-coolant accident (LOCA) environment for P%2 and B'ai plants are ij described in the following:
(1) Topical Report WCAP-8312A for Westinghouse plants.
4 '
- 1
.]
(2) Section 6.2.1 of CISSAR System 30 PSAR for Cc=bustion Engineering
,i plants.
s
.3
- )
(3) Appendix 6A of 3-SAR-205 for Babcock & Wilcox plants.
i 2
(4) a. NEDO-10320 and Supplements 1 & 2 for General Electric plants.
- b. NEDO-20533 dated June 1974 and Supple =ent 1 dated August 1975 (GI Mark III).
Acceptable methods for calculating the mass and energy release to determine the main steam line break (MSL3) environment are described in the following:
(1) Appendix 6B of CISSAR System 80 PSAR for Combustion Engineering plants.
(2) Section 15.1.14 of B-SAR-205 for Babcock & Wilcox plants.
(3) Same as item (4) above for General Electric plants.
l (4) Topical Report WCAP-8822 for Westinghouse plants.
(Although this Topical Report is currently under review, the use of this method is acceptable in the interim if no entrainment is assumed. Reanalysis may be required following the NRC staff review of the entrainment l:
model as presently described.)
l.
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s.
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-2:
t o
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,-4
- i
' i:
i APPENDIX B II SAMPLE CALCULATION AND TYPE METHODOLOGY FOR RADIATION QUALIFICATION DOSE l
L l
L i
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+
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.._. m.
a.
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J 5
h APPENDIX B
- 1 SAMPLE CALC'JLATION AND TYPE METHODOLOGY FOR RADIATION QUALIFICATION 00SE 4
This appendix illustrates the staff model for calculating dose rates and t
integrated doses for equipment qualification purposes. The doses shown in Figure B-1 W include contributions from ~ Q &,e A f_
U
- ;
- 0 i
- r...; in o
the containment and cover a period of one year following the postulated fis-
'r i
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
&1 Figure 8-1 k ima, have been determined using models and assumptions etag!:=e=nn 4 4 d;, _-A.
n_'__
- - 7.
This analysis is conservative, and fac-m w--
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.
I o
Doses were calculated forg ur point,( inside the containment 4 ht the mid-e pointofthecontainment]
king sprays and plate-out mechanisms into account 2-,e a.
e,,
,w
...,_.,'w.
The doses presented in Figure B-1 are 1
values for a PWR plant having a containment free-volume of 2.5 million cubic l~
feet and a power rating of 4100 MWt.
l i
B-1 l
/
+-
.e..
s..,_...
. u.._...a.
.i
-J l
- 1. 0 Basic Assumptions Used in the Analysis a.4 d h Gamma and beta doses and dose rates )epe' detennined for three types of
(
3 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. J5gs:f[given piece of equipment may receive a dose contribution from any or all of these sources.
The count 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 sour <:e, and the effects of shielding.
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-ing to a release of fission products, and that the mix of fission produ.is 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 releases under degraded core conditions. While the final resolution of the f
source term assumptions is conditioned on the completion of these rulemaking efforts, the staff believes it is prudent to incorporate the knowledge gained of fission product behavior from the TMI-2 accident in defining source term l
assumptions for equipment qualification.
1 Based upon release estimates in the Rogovin Report (Ref. 1), the staff 4sm f., &
assumptions for noble gas and iodine releases Akt still conservative. However, a
the report estimates that the THI-2 release contained between 40 and 60 percent B-2 l
..._m.._a _ _ w..c..m..
u.. n. ;
.c. _, 4.
u....:... s. -
+
e of the Cs-134 and Cs-137 core activity in the primary system water, in the con-I tainment sump water, and in'#erauxiliary building tanks, Comparison of the integrated dose from the TMI-2 cesium release to the previous staff assumption a-nn d.L r d '4 /* A%-
of "l% solitis" shows tnat "1% solids" gases (5%9rmay notyconserv'ative for 4
g
- b equipment required to function for time periods exceeding thirty days. The staff feels that as a first step toward modification of the TID-14844 source term in the direction indicated by the TMI-2 experience, it may be prudent to factor
~
1 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 l
with the findings of recent source term studies reported in NUREG-0772 (Ref. 2).
I This report also concluded that the expected predominant form of iodine released during accidents is cesium iodide (CsI).,Although the Csl form is not specifically addressed in this report, it is evident that either CsI, or I2 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 the incorporation of additional radionuclides, would be factored into the guide
'before it is issued as an effective' guide.
B-3
)
w.m..
.a.
.i i.
i
- 2. 0 Assumotions Used in Calculatino Fission Produce Concentrations
- l This section discusses the assumptions used to simulate the PWR and BWR containments for determining the time-dependent and location-dependent dis-i Ej tribution of noble gases and todines airborne within the containment atmos-y
,i phere, plated out on containment surfaces, and in the sump water.
~
The staff has developed a computer program, TACT, (' ': _ E L.. O that j
.e:dels the time-dependent behavior of iodine and noble gases within a nuclear l
power plant. The TACT code is used routinely by the staff for the calculation i
of the offsite radiological consequences of a LOCA, and is an acceptable method i
for modeling the transfer of activity from one containment region to another and in modeling the reduction of activity due to the action of ESFs. Another staff code, SPIRT (Ref. 3), is used to calculate the removal rates of elemental iodine by plate-out and sprays. These codes were used to develop the source term estimates. The following assumptions were also used to calculate the distribution of radioactivity within the containment following a design basis LOCA.
2.1 PWR Orv Containments a.
The source terms used in the analysis assumes that 50 percent of the core iodines and 100 percent of the core noble gases were released instantaneously to the containment atmosphere, 50 percent of the core inventory of cesium and 1% of the remaining " solid" activity inventory is released from the core and carried with the primary' coolant directly I
to the containment sump.
("-tr.
"r # ' :
-'*d
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rala m w # T!: '.46*
.3 app rux. iai.e.j ;#al
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...:..... - u.a..:. :. p.w. s
.a..
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6 3
b.
The containment free volume was taken as 2.52 x 10 ft. Of this volume, 74 percent or 1.86 x 10 ft is assumed to be directly 6
3 covered by the containmen't sprays.
(Plants with different contain-U
[]
ment free volumes should use plant specific valves.)
5 3
c.
6.6 x 10 ft of the contaidment free volume is assumed unsprayed, which includes regions within the main containment space under the containment dome and compartments below the coerating floor level.
P d.
The ESF fans are assumed to have a design flow rate of 220,000 cfm in the post-LOCA environment. Mixing between all major unsprayed regions and compartments and the main sprayed region is assured.
e.
Air exchange between the sprayed and unsprayed region was assumed to be one-half of the design flow rate of ESF fans. Good mixing of the 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 gpe of bnric acid solution into the containment.
g.
Trace levels of hydrazine was assumed added to enhance the removal of iodine.
h.
The spray removal rate constant (lambda) was assumed and calculated using the staff's SPIRT program, conservatively assuming only one B-5 l
l
~
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~~. --
.~
l i
- .i
- 5 spray train operation and an elemental iodine instantaneous parti-
-)
tion coefficient (H) of 5000. The calculated value of the elemental
~1 iodine spray removal constant was 27.2 hr 4
1 i.
Plate-out of iodine on containment internal surfaces was modeled as
- }
a first-order rate removal process and best estimates for model parameters were assumed. Based on an assumed total surface area 5
2 within containment of approximately 5.0 x 10 ft. The calculated i
value for the overall elemental iodine plate-out constant was
~1 1.23 hr The assumption that 50 percent of the activity is
.J instantaneously plated-out should not be used.
j.
The spray removal and plate-out process were modeled as competing iodine removal mec*anisms.
I I
k.
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 concentration was 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 plate-out removal mechanisms.
The sprays were assumed to remove elemental iodine until the instan-1.
t I
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.
B-6
.........x
..a..
. c..~ s a :. :..... u..; 0..x..
... ;.T 1
m.
A relatively open (not compartmented) containment was assumed, and the large release was uniformly distributed in the contairment.
This is an adequate simplification for dose assessment in a PWR con-5 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 iodine is present in a design basis LOCA. The calculation of the post-LOCA environment assumed that 2.5 percent of the cere inventory of the iodine released is associated with airborne particulate mate-rials and 2 percent of the core inventory of the iodit e rdeased formed organic compounds. The remaining 95.5 percent remained as elemental iodine.
For conservatism this composition was assumed present at time t =
ese 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.
'p.
Removal of airborne activity by engineered safety features may be i
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, d
L B-7
l.... w. -
.a z,.. u
.w...:. 2
~
<r 2.2 PWR Ice Condenser Containments The assumptions and methods presented for the calculation of the radia-4 I
tion environment in PWR ory containments are appropriate for use in calculat-ing the radiation environment following a design basis LOCA for ice condenser 4
- i containments with the following modifications
a.
The source should be assumed to be initially released to the lower containmant ccmpartment. The distribution of the activity should be
]
based on the forced recirculation fan flow rates and the transfer i
rates through the ice beds as a function of time.
b.
Credit may be taken for iodine removal via the operation of the ica 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:
e B-8 e
yee 6-
- m.. ~,. -,,
.... -.i.~. -..
.,.... ~. a
. 4 1
1:
e A decontamination factor (DF) of 10 Oi be assumed for both the a.
elemental and particulate fodine as the iodine activity passes j
through the suppression pool. No credit should be taken for the removal of organic iodine or noble gases in the suppression pool.
b.
For Mark III designs, all of the activity passing through the suppres-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 4
area and the transfer of activity from these regions via containment 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 I
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-soecific parameters.
d.
The removal of airborne activity from the reactor building by opera-tion of the Standby Gas Treatment System (SGTS) may be assumed.
3.0 Model 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 w'ere calculated for a midpoint in
~
B-9
^
l d
4
~
the containment. The containment was modeled as a cylinder of equal height and diameter.
Containment shielding and internal structures were neglected.
]
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 f
Reference 4 to result in only a small error. H ?'t;c;i.tei.....at m
,.w + - g : 17<<r.,,j -g g 33;,,,, _ g _,n:_,,,,,n,u, c'n :t_....
For beta dose calculations far equipment located on the contain-
\\
ment walls or on large internal structures, the semi-infinite beta dose model may be used.
i
~~
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-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 by the sprays of the plated-out iodine activity was not addressed in this j
evaluation.
i Finally, all gamma doses we,re multiplied by a correction factor of 1.3 as i
suggested in Reference 4 to account for the omission of the contribution from the decay chains of the isotopes.
1 l
4.0 Model for Calculating the Dose Rate of Sume Fission Products l
The staff model assumed t e washout of airborne iodine from the contain-l ment atmosphere to the containment sump.
For a PWR containment with sprays
{
and good mixing between the sprayed and unsprayed regions, the elemental 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 I
the airborne iodine in less than 15 minutes).
s n s.... J. a. a;. w + = & 4 Jg&
RJ.: 4, m,d -
L.
a.
4 s,s c -
xg U.4; W a u A.4.
+...d,d A.s c, ?, s.;
A A AY% M f A -*s-B-10 A, A-J' ~' j ' ' ".
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1 A A' V
d %.
. l.. :.
..._..........._s A
T Thedosecalculationsassume,4(atime-dependentiodinesource. (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 not significant.)
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.
1 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 Offica 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.
B-11
TABLE B-1 SUMMART TABIZ OF ESTIMATES FOR TOTE AIRBORNE GA".A DOSE CONTRI3UTORS IN CONTAINMENT TO A POINT IN THE CONTAINMENT CENTER TIME AIRBORNE IODINE AIRBORNE NOBIE GAS PLATE-0UT IODINE TOTAI, DOSE (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
- 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 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 1.50Et7 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 1.53Et7 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 2.20Et6 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 3.92L+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 TOTE 1.54E+7 B-12
,e e e6
-%-w+-.
s TAatZ B-2 S122fARY TABLE OF ESTDfATES FOR TOTAL AIRBORNE BETA DOSE CONTRIBUTORS l
IN CONTADCfENT TO A POINT IN THE CONTADCfENT CINTIR TIME AIRBORNE IODINE AIRBORNE NOBLE GAS TOTAL DOSE (HRS)
DOSE (RADS)+
DOSE (RADS)+
(RADS)+
0.00 0.03 1.47+5' 5.48+5 6.95+5 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 0.18 4.49+5 2.14+6-2.59+6 0.21 4.73+5 2.35+6 2.82+6 0.25 5.00+5 2.60+6 3.10+6 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 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 4.08t7 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 1.0$+8 298.
5.64+6 1.17+8 1.23+8 394.
5.99+6 1.25+8 1.31+8 L
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 1060 6.69+4 1.44+8 1.51+8 1220 6.73+6 1.45+8 1.52+8 1390 6.75+6 1.47+8 1.54+8 1560 6.76+6 1.49+8 1.56+8 1730 6.76+6 1.51+8 1.58+8 1900 6.76+6
- 1. 5 2+8-1.59+t 2060 6.76+6 1.54+8 1.61+8 2230 6.77+6 1.55+8 1.62+8 2MG 6.77+6 1.62+8 1.69+8 3670 6.77+6 1.69t8 1.76+8 4390 6.77+6 1.76+8 1.83+8 5110 6.77+6 1.83+8 1.90+8 5830 6.77+6 1.89+8 1.96+8 6550 6.77+6 1.96+8 2.03+8 7270 6.77+6 2.03+8 2.10+8 8000 6.77+6 2.09+8 2.16+8 8710 6.77+6 2.16+8 2.23+8 TOTAL 2.23+8 B-13 9
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d 1-j REFERENCES i:
R 1.
Mitchell Rogovin, George T. Frampton, Jr., et al, "Three Mile Island--
l!
a report to the Commissioners and to the Public" NUREG/CR-1250, Volume II, i
Part 2.
2.
NUREG-0772, " Technical Basis for Estimating Fission Product Behavior During J
LWR Accidents."
l;.
M.
3.
A. K. Postm and P.
am, " Technological Bases for Models of Spray Wash-out and Air orne Contaminants in Containment-Vessels," USNRC Report NUREG/CR-0009, November 1978. ' Available for purchase from National Tech-j nical Information Service, Springfield, Virginia 22161.
i 4.
M. J. Kolar and N. C. Olson, " Calculation of Accident Ooses to Equipment Inside Containment of Power Reactors," Vol. 22, pp. 808-809 in Transac-tions of the American Nuclear Society,1975. Available from tecnnical libraries.
~
BIBLIOGRAPHY
(
y 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 i
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 l
Society, 1976. Available from technical libraries.
s T C. Kocher, ed., " Nuclear Oecay Data for Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," ORNL/NUREG/TM-102, August 1977.
Available for purchase from the National Technical Information Service,
' Springfield, Virginia 22161.
i l*
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,
.,l' 1974. Available from technical 11oraries.
R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, " Fission Product Source i -
l Terms for the LWR Loss-of-Coolant Accident:
Summary Report," USNRC Report i
NUREG/CR-0091, May 1978. Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
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AP?ENDIX C QUALIFICATION DOCUMENTATION Folt "mTC EQUIPMENT IMPORTANT TO SAFETT 4
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QUALIFICATION Dommog yog, nvr r'tTC EQUIPMENI IMPOBnNT TO SAFra
!l In order to ensare that an environmental qualification program conforms with General Design Criteria 1, 2, 4 and 23 of Appendix A and Sections III ard II j
of Appendiz 3 to 10 C32 Part 50, and to the national s.asdards sentioned in
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Part II "Acc=ptance Czite:ia" (which iselndes IEEE Std. 323) contained in y
Standard Review Plan Section 3.11, the following info =mation on the
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gn=14 Meation program is required for all electric equipment important to safety.
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Identify all elect 1c equipment.important to safety and provide the following:
a.
Type (functional designation) b.
Manufacturer c.
Manufacturer's type number and model number d.
The equipment should include de followi=g, as applicable:
(1) Switchgear (2) Motor control cacters (3) Valve operators (4) Motors (5) Lagic. equip:sent (6) Cable (7) Diesel generator control equipment (81 Sensors (pressure, pressure differescial, ta=perature and neutron) i!
(9) Limit switches.
(10) Heaters (11) Tans (12) Control boards t
(13) I=st:nment racks and panels (14) Connectors (15) Electrical penetrations (16) Splicas (17) Termisal blocks 2.
Categori:e the equipment identified in ite: I above isto one of the following categories:
Equipment that will experience the environ =estal conditio:s of a.
design basis accidents for,enich it crust function es sitigate said accidents, and that will be qualified to demonstrate operabilig is i
l the accident envirec=ent for the ti::e required for accident zitigation with safety sargin to failure.
b.
Equipment that will experience envirec=e=.at conditions of design l
basis accides_s through which it need not f==ction for mitigation of said accide: s, but drou;h which it =s
- o fail is a =a::e:
det:d= ental to plant safety or accident mitigation, a=d that will be qualified to de=enst:sta the espabilig to wi.hsxd any actid:::
enviro =ent for the time duri:g which it must not fail vi h safety sargis to failu:3.
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.m Iquipment that will experience enviro = mental conditions of design c.
basis accidents through which it need not function for nitigacion of j
said accidents, and whose failure (in any mode) is de==ed not detrimental to plant safety or accident mitigation, and need not be 1
qualified for any accident environment, but will be gnmH fied for its non-accident servica environment.
Equir-ae thstr will not experienca environnental conditions of 3
d.
design basis accidents and that will be q==Wied to demonst:sta 1
operability undar the expectad ext==mes of its non-accident servics This equipment would no== ally be located outside the 1
. environment.
reactor connN-at.
For each type of equipment in the categories of equipment listed in
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j 3.
item 2 above, provide separately the equipment design specification j-requirements, including:.
The system safety function :=quirements.
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l An environmental envelope as a function of time that includes all b.
ext =me parameters, both maximum and Wmm values, expec.ed to
- ~~
occur during plant shutdown, normal operation, abnormal operation, and any design basis event. (including I,0CA and MSI3), including post-event conditions.
Time required to fulfill its safety function when subjected to any c.
of the extremes of the environmenn envelope specified above.
Technical bases thould be provided to justify the place =ent of each d.
type equipment in the categories 2.b and 2.c listed above.
Provide the qualificatf.on test plan, test setup, test procedures, and 4.
acc.ptance criteria. fod at least one of each group of equipment of If item 1.d as appropriate to the category identified in item 2 above.
any method other than type testing was u. sed for qualification (operating experience, analysis, cembined qualification, or ongoing qualification),
describe the method in.gufficient detail to permit evaluation of its adequacy.
I For each category of est*j.; ment identified in item 2 above, state the 5.
actual qualification en@ lope simulated during testing (defi=ist the duration of the hostile,:svironment and the sargin in excess of -Jze design requi.==ents). M any method other than type tes ing was used for qualification, identify 'che method and define the equivalent
" qualification envelope" so derived.
A summary of test resulta that denonstrates the adequacy of he 6.
If analysis is used for qualification, qualification prog:sm.
justification of all analysis assu=ptio=s =ust be provided.
7.
Identification of the qualification docunents which contain detailed supporting information, including test data, for itens 4, 5, and 6.
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-l APPENDIX D THERFAL AND RADIATION AGING DEGRADATION 4
0F SELECTED MATERIALS 4
4 Table D-1 is a partial list of materials *hich may be found in a nuclear power plant along with an indication of the material susceptibility to radiation and thermal aging.
0 i
Susceptibility to significant thennal aging in a 45 C environment and nonnal atmosphere f5r 10 or 40 years is indicated by an (*) in the appro-priate column. Significant aging degradation is defined as that amount of degradation that would place in substantial doubt the ability of typical equipment using these materials to function in a hostile environment.
. Susceptibility to radiation damage is indicated by the dose level and the observed effect identified in the column headed BASIS. The meaning of the tenns used to characterize the dose effect is as follows:
i Threshold - Refers to damage threshold, which is the radiation I
e exposure required to change at least one phys'ical property of the material.
Percent Change of Property - Refers to the radiation exposure e
required to change the physical property noted by the parcent.
Allowable - Refers to the radiation which can be absorbed before e
serious. degradation occurs.
The infonnation in this appendix is based on a literature search of sources including the National Technical Infor=ation Service (NTIS), the National Aeronautics and Space Administration's Scientific and Technical Aerospace Report (STAR), NTIS Government Report Announcements and Index (GRA), and D-1
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[arious manufacturers' data reports. The materials list is not to be con-j sidered all inclusive, neither is it to be used as a basis for specifying e
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The list is soley intended for use by the licensee in making judgments as d
to the possibility of a particular material in a particular application be-ing susceptible to significant degradation due to radiation or thermal aging.
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The data base for thermal and radiation aging in engineering materials is rapidly expanding at this time. As additional information becomes available, G-l the licensee is expected to utilize the data as applicable.
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s APPENDIX 7-PICAL EQUIPMENT / FUNCTIONS FOR ACCIDENT MITIGATION l
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- APPENDIX [
TYPICAL EQUIPMENT / FUNCTIONS FOR ACCIDENT MITIGATION Engineered Safeguards Actuation Reactor Fratection d
Containment IsoTation Steamline Isolation Main Feedwater Shutdown and Isolation Emerge::cy Power 1
l Emergency Core Cooling Containment Heat Removal Containment Fission Product Removal
'i I'
Containmnt Combustible Gas Control Auxiliary Feedwater Containment Ventilation l
Containment Radiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)
Ventilation for Areas Containing Safety Equipment Component Cooling i
Emergency Shutdown 3
Post Accident Sampling and Monitoring 3
Radiation Monitoring 3
Safety Related Display Instre.entation i
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s 1These systems will differ for PWRs and SWRs, and for old r and newer In each case the system features which allow f transfer to pl ants.
recirculation cooling mode and establishment of long term cooling witn boron precipitation control are to be considered as part of the system to be evaluated.
R 2 Emergency shutdown systems include those systems used to bring the plant to a cold shutdown condition following accidents which de not result in a breach of the reactor coolant pressure boundary together Examples with a rapid depressurization of the reactor coolant system.
of such systems and equipment are the RHR system, PORVs, RCIC, pressurizer a
sprays, chemical and volumi control system, and steam dump systems.
More specific identification of these types of equipment can be found 3
in the. plant emergency procedures and in Tables 1 and 2 of Regulatory Guide 1.97, Categories 1 and 2.
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VALUE/ IMPACT STATEMENT FOR REVISION 1 TO REGULATORY GUIDE 1.89, " ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO ';AFETY FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS" Background _
Regulatory Guide 1.89, " Qualification of Class lE Equipment for Nuclear 1
Power Plants," is being revised to reflect current staff position on equipment qualification, NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-j Related Electric Equipment," was issued for public coament in December 1979.
Subsequent to its issuance for ccmment, the Commissioners (see " Memorandum and i
Order", dated May 23,1980) directed the staff to use NUREG-0588 along with l
"00R 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.
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 DOR guidelines. This revision to Regulatory Guide 1.89 will provide guide-lines for meeting the Commission's equipment qualification rule and is essentially i
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 frem TMI-2 in source term definition.
Substantive Changes and Their Value/Imoact Regulatory Position C.2 (which provided radiological source terms for 1.
equipment qualification tests) was deleted and the following positions were added:
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' 2-(a) New Position 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 shoul'd'be the primary method. ~ The l^ '
i other methods, when used, should be supported by test data.
(b) Position C.3 was added which provides the staff position pertaining to establishing performance and environmental requirements for equipment qual-ification. Methods for establishing temperature and pressure profiles for loss-of-coolant accident and main steam line break are provided, and radiological source tenns are given.
(c) Position C.4 was added which provides the staff position pertaining 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 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 - The above provides the staff's position on individual sections of t
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.
Imaact - The impact could be considerable since the scope of the guide has been expanded to include additional equipment. (Class lE is only a subset of l
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equipment important to safety.) The total impact will depend upon the final y
a amount of additional equipment not previously qualified in accordance with Class lE
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The impact of the various positions of 1
equipment qualification requirements.
fj the guide should be minimal. The positions merely take established NRC pro-j visions and relate them to appropriate sections of an endorsed voluntary con-
- 4 sensus standard.
q
'j 2.
Position C.3.d(3) was added which is not part of NUREG-0588, but which I
provides a source tenn for use in the qualificatio. of certain accident-monitoring 1
instrumentation specified in Regulatory Guide 1.97.
This certain instrumen-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 detemination on a case-by-case basis. This will enhance the licensing process.
Imoact - There is no impact. The source tenn of Position C.3.d(3) is not imposed by this regulatory guide.
It merely provides an acceptable tenn 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 00R 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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1.
l YANNEE ATOMIC ELECTRIC COMPANY
/
2.C.2.2 1
1671 Worcester Road, framingham, Maucchusetts 01701 m 81-20 YkNKEE w
e September 14, 1981 i
o Honorable Nunzio Palladino, Chairman i
U.S. Nuclear Regulatory Commission l
Washington, D. C.
20555
Dear Chairman Palladino:
Having lived through the initial stages of the NRC Staff's planned program
~
on equipment qualification and having considered where the Staff is headed, we are deeply concerned that the general approach used is inconsistent with sound j
regulatory policy. We believe your immediate attention to this matter is imperative.
NRC and indus try resources are presently being heavily taxed in a herculean effort merely to produce documentation to satisfy the Staff's interim methods which are evolving for compliance with General Design Criteria 2 and 4 for environmental qualification of electrical equipment. We anticipate a minimum optimistic resource commitment of two to three million dollars per plant to etztply with the current Staff requirements. This is the direct result of NRC establishing.immedia tely effective interim docu=entation requirements in the form of orders on individual plants.
I Sound regulatory policy mandates that im=ediately effective interim l
requirements should be used sparingly and only when justified by immediate l
public health and safety concerns. No such jus tification has been shown in l
this ins tance. Rather, the NRC Staff, Commission and industry have consistently stated that there are no in:=ediate public health and safety concerns on the equipment qualification issue.
I The NRC is proceeding on this course of action without having analyzed the reduction, if any, in risk that such interim requirements vill achieve and without having performed a value-impact analysis as required by Co= mission policy. For such actions, it is Commission policy to prepare value-i= pact analyses to the standards set forth in the Commission's " Guidelines For Conducting Value-Impact Analysis" dated Jar.uary 1978.
(43FR34358 - 8/3/78).
In a recent ACRS meeting, Dr. Kerr expressed these same concerns (Attachment 1).
I Compounding the current problem, the Staff plans to propose, through rulemaking, their final precise documentation requirements. The potential harm created by such a two phased regulatory approach includes the possibility that the interim requirements imposed will be modified af ter full consideration in rule =aking, and that such modifications will thus result in additional and unwarranted plant outages to i=plement changes, backfits of items previously corrected pursuant to the interi: require =ents, and duplicative con:mitment of industry and agency resources. Here, the likelihood for changes is significant because again no value-i= pact analysis is being performed to determine the need and whether there are acceptable alternatives that would not require as significant a resource ce==it=ent.
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- Honsrsblo Nunzio Palledins, Chairman September 14, 1981 Page 2 l.
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Unfortunately, this dismal scenario is apparently just the tip of the
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-iceberg because a similar process is planned in the near future regarding mechanical equipment qualification, seismic qualification, and qualification of electrical equipment exposed to mild environments. Rescurce commitments will, in all likelihood, result in additional expenditures well in excess of
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those previpusly stated.
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The past experience with the interim requirements approach has resulted in costly and unnecessary backfits (e.g., fire protection regulations) sad
'l clearly supports' our position that this approach is unwarranted and should not be condoned by the Commissioners.
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3 In addition to our concern that the " interim" regulatory approach is inappropriata, we are equally disturbed that the failure of the NRC to integrate and prioritize these and other Commission requirements necessitating if' major resource commitments may result in a degradation of overall plant safety. Absent a reasoned decision on the incremental benefits to the public health and saf ety to be gained by the documentation effort, and absent a l
broader prioritization of the numerous requirements recently issued by the NRC (e.g., fire protection,, emergency planning, TMI-related requirements) and i
other tasks facing licensees, such a commitment of industry resources should not be required.
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We strongly urge the Commission to direct the Staff to terminate the
" interim" approach to regulation (including the forthcoming " interim" requirement for seismic qualification)'in favor of a more reasonable deliberate policy which requires a thorough evaluation of the need for new i
requirements. The evaluation should be supported by realistic value-impact
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analyses.
If the evaluation reveals a need for new requirements, sufficient flexibility shocid be provided in the new requirements to allow for plant specific value-impact analyses and/or plant specific alternatives. Finally.
all NRC imposed requirements should be integrated and prioritized with reasonable completion dates that take into consideration the effect on plant '
i safety of over-extending indus try and NRC resources.
We request that our concerns and suggestions be -considered by the i
Commission in the near future. We believe Commission attention to these matters is timely and imperative, and will benefit the Public, NRC and the
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regulated industry.
Very truly yotirs, YANKEE ATOMIC ElICTRIC COMPANY D. E. Vandenburg Senior Vice Pres a.t cc: Com=issioner Ahearne Commissioner Roberts
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Commissioner Gilinsky Commissioner Bradford t
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37
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- ACRS SUBCO.W ITTEI ON INVIR0!MINTAL QUALIFICATION
'4 MEETING ON JULY 22,1981 (Pages 37-39)
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1 MR. KIRR:
In terms of temperature it vould be the
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2 large 10CA and the large steamline break.
Usually the LOC 1 r
3 is a little bit high,er.
In terms of radiation one of our I
4 msin concerns was whether equipment which is close to the 1
5 circulating fluid lines is properly qualified for the 6 environ = ental conditions.
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7 ER. KERE:
Okay.
Yov, a second question is has
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8 there been any effort on the part of Standards or whoever to 1
9 look at this program and estimate the decrease in risks tha t 10 it is likely to produce?
11 MR. SUllIVAN:
Not on th.e. part of the Office" of 12 Res ea rch.
- There is no more Office of Standards.
I just 13 va n t to m ak e th at poin t.
In th e Office of-Research, no, as
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14 f ar as I know there has been effort to quantify any decrease
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15 in risks.
EE. KIRE:
I think everybody involved agrees tha.t 16 17 this is going to cost a lot of money.
It would seen.to me 18 that one would want to have some general idea that one would!I is predict a concomitant decrease in risk.
Indeed, I'vas statement that no 20 puzzled when I read the value impact statement that 21 mention was made of this.other than a passing 22 the public health and safety would be enhanced by this rule.-
I I would have expected that one might ask somebody 23 to try to balance the risk decrease that was being bought b; 24 25 what I would guess might be quite a lot of money.
Is this*
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'I 1 coing to be pretty e xp e nsiv e ?
- i ER..SD11IVAN
Yes. 'To implement the entire e
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yes.
All I can sa y g 3 program that Dr. Roszteczy' has undervay, We 4 is /that in Researrh we have not done any such study.
study rather crude qualitative a
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5 have done simply I guess a is the more assurance tha t you have that equip =ent 6 that but that 7 qualified, the more assurance you have of safety,
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S is it.
seems to me, for exarple, a
-f ER. XERE:
It just 9
helpful in order to establish 10 study of this kind migh t be I
of things that one vould do first.
emphasis in the search 11 you are faced with this to remind you because 12 d on ' t have people who operate plants to do 13 everr day that we are asking of things and they can't.do them all s,imultaneously.
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14 a lo t an extr'emely high priority need or it may be a 15 This may be 16 low priority need.
one would need to have It would seem to me that 17 so me thing 18 some inf ormation of that. kind in order to know b
19 about resource allocation.
this rule Dr. Kerr, by presenting MB. AGGARWAL:
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20 it vill give us to the public for theit comment 21 and going l.
At this 22 also some idea. as to what in f act they vill f ee you is go along with us to have it time all we are asking 23 i.
24 published f or public comment..
l of your responsibilitics, ER. KIRE:
Well, one 25
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i ALDERSoN REPoRDNG COMP ANY. INo.
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1 according to what IrNdhere, is some sort of rational 3
<i, 2 eff ort to estimate the risk and the benefit associated with da 3 the rule change.
What I see in here is pretty cursory.
It
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4 s.a r s it is going to cost a lot, which could to me mean
's S o million.
I ' don 't know which order i
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- a 6 of m'aquitude ve.are talking about.
It says the public 7 saf et:r vill be improved.
Now, I think f or an organiza' tion ' as sophisticated J
8 9 as the Nuclear Regulatory Coccission nov is one could expect to a little more quantita'tive description.
I would think in 11 allocating your own resources you.vould vant a more I
12 quantitative idea of what is being accomplished.
13 MR. AGGARWALi I think, Dr. Kerr,'you have a. valid
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o' t to you, sir, that all nuclear W
14 point.
But may I point u
these 15 p6ver plants at this time are required to meet 18 requirements under the Commission order.
HR. KERRa* Well, the Commission could be exactly 17 1a right and it could be wrong.
If it is wrong it vill never
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is know it until its staff tells it that it is because the and the time 20 technica.1 capability in the Commission 21 available is not such that they can make decisions like this 22 unless they have input from people who are experts and that t
l 23 is you guys.
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- 53. HINTZE4 The requirements. tha t are included iI s
24 25 this are no different from what has been required since 1971
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f ALDERSoN REPORTING CoWPANY. INC.
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