ML20027D813
| ML20027D813 | |
| Person / Time | |
|---|---|
| Issue date: | 11/05/1981 |
| 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 8211100123 | |
| Download: ML20027D813 (44) | |
Text
1
,c
('/
Draf? +
ll/s/81
,, d t','+l.,,'k, %
w Proposed Regulatory Guide 1.fi9 Rev.1 ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS A.
INTRODUCTION Criterion III, " Design Control" and Criteria XI, " Test Control," of Appen-dix 3, "Q,ality Assurance Criteria for Nuclear Power 81 ants and Fuel Reproc-essing Plants," to 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities," requires that design control measures provide for verifying theadequacyofaspecificdesignfeaturebydesignreviews,byvaEiouscalcu-lational methods, or by suitable quali.fication 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, " General Orsign Criteria for Nuclear Power Plants," to 10 CFR Part 50 and S 50'.49, " Environ-mental Qualification of Electric Equipment for Nuclear Power Plants," to 10 CFR Part 50, require 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 nuclear power plants to ensure that the equipment can perform its safety function.
a e
e
.1 9 if
1%:
f
..i ;
_ _. _. I CURRAN 82-426 PDR
/,
B.
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 (IEEE), and subsequently was approved by the IEEE Standards Board on December 13, 1973.
The standard describes basic procedures for qualifying Class 1E equipment and interfaces that are to be used in nuclear power plants, including 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 that, when satisfied, will confirm the adequacy of the equip-ment design for the performance of safety functions under normal, abnormal, design-basis-event, post-design-basis-event, and containment-test conditions.
Equipment should be q'ualified to meet its performance requirements under the environmental and operating conditions in which it will be required to function and for the length of time for which its function is required. The following are examples of considerations to be
't, ken into account when deter-mining the environment for which the equipment is to be qualified:
(1) equip-ment outside containment would generally see a less severe environment than j
equipment inside containment; (2) equipment whose location is shielded from a l
radiation source would generally receive a smaller radiation dose than equip-ment at the same distance from the source but exposed to its direct radiation; t
yCopies may be obtained from the Institute of Electrical and Electronics Engineers, Inc., United Engineering Center, 345 East 47th Street, New York, New York 10017.
j 2
1
~
b 5 O
(3) equipment required to i.M tiate protective action would generally be required for a shorter period of time than instrumentation required to follow the course of an accident. The sp xific 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 (with margin) it is required to operate.
A component to be qualified in a nuclear radiation environment should be j
exposed to a fluence that simulates the total dose and dose rate, conservatively calculated, that the component should withstand prior to completion of its l
intended function. Dose rates, spectrum, and particle type should be simulated as closely as practicable unless it can be shown by analysis that damage is not l
l significantly depen:!ent on dose rates, spectrum, or particle type.
Regulatory Position C.1 of this guide calls for the qualification of addi-tional equipment which if it malfunctioned or failed when subjected to an acci-dent condition could negate the safety function of essential systems and equip-ment. The NRC staff is currently assessing the need to include the rod control system among this equipment, considering the potential adverse effect of rod control system malfunction.
~
l Regulatory Position 4.d(12) of this guide addresses qualification of equip-ment exposed to low-level radiation doses. Numerous studies that have compiled lj radiation effects data on all classes of organic compounds show that the least radiation resistant compounds have damage thresholds greater than 104 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 or operating experience for radiation exposures below I
104 rads. However, for electronic components, studies have shown failures in 3
4
_J.
n.
,.-~.
I
~
metal oxide semiconductor devices at somewhat lower doses. Therefore, radiation qualification for electronic components may have a lower exposure threshold.
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 mainte-nance and quality assurance be established, including minimizing dust accumula-tion 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,"1-dated Feb-ruary 28, 1974, are acceptable to the NRC staff for qualifying electric equip-ment for service in nuclear power plants to ensure that the equipment can per-form its safety functions subject to the following:
1.
Section 50.49, " Environmental Qualification of Electric Equipment for Nuclear Power Plants," of 10 CFR Part 50 requires that essential electric systems and equipment be qualified to perform their intended functions. Typical essential systems and equipment that mitigate accidents are listed in Appen-
~
dix 0 of this guide. Additional equipment should also be qualified for the accident conditions if its malfunction or failure due to such conditions will negate the safety function of essential systems and equipment. Examples of such additional equipment that should be considered for qualification are associated circuits as defined in Regulatory Guide 1.75, " Physical Independ-ence of Electric Systems" and the pressurizer pressure control.
4
_ _1
~
~
5 l.
2.
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.
However, the testing should be performed in its proper sequences as indicated in Section 6 of IEEE Std 323-1974.
3.
Section 5, " Principles of Qualifications" of IEEE Std 323-1974 presents various methods for qualifying equipment, including analysis. The 1
NRC will generally not accept analysis in lieu of testing. 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 if testing of the. equipment is impractical because of size limitations or by the state of the art. However, partial type test data should be provided to support the analytical assumptions and conclusions reached.
4.
Section 6.2 of IEEE Std 323-1974 requires equipment specifications l
l to define performance and environmental requirements. In defining the require-meats called for in item (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 to this guide. The calculations should account for the time dependence and spatial distribution of these vari-ables. For example, superheated steam followed by saturated steam may be a limiting condition and should be considered.
5 F
[
l, e
(a) For pressurized water Reactors (PWRs) with a dry containment, cal-culate LOCA containment environment using CONTEMPT-LT or equivalent industry codes. Additional guidance is provided in Standard Review Plan (SRP) Sec-tion 6.2.1.1.A, NUREG-0800.
(b) For PWRs with an ice condenser containment, calculate LOCA contain-ment environment using LOTIC or equivalent industry codes. Additional guidance is provided in SRP Section 6.2.1.1.B, NUREG-0800.
l (c) For boiling water reactors (BWRs) with a Mark I, II or III contain-ment, calculate LOCA environmeni, using methods of GESSAR Appendix 3B or equiva-lent industry codes. Additional guidance is provided in SRP Section 6.2.1.1.C, NUREG-0800.
(2) Since the test profiles included in Appendix A to IEEE Std 323-1974 are only representative, they should not be considered an acceptable alterna-
)
i-tive in lieu of using plant-spe ific containment temperature and pressure i
design profiles unless plant-specific analysis is provided to verify the applicability of those profiles.
b.
Tenparature and Pressure Conditions Inside Containment for Main
'I l,
Steam Line Break (hSLB).
Methods acceptable to the NRC staff for calculating the environmental parameters of an MSLB used for equipment qualification are provided below.
(1) Models that are acceptable for calculating containment parameters are listed in Position 4.a(1).
~~
(2) Since the test profiles included in Appendix A to IEEE Std 323-1974 are s.ly representative, they should not be consi.dered an acceptable alterna-tive 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.
6
. f
s c.
Effects of Chemicals Guidelines for the chemical spray are provided in SRP Section 6.5.2 (NUREG-0800), paragraph II, item (e). For plants that use domineralized water as spray solution, effects of the spray should also be considered.
d.
Radiation Conditions Inside anc Outside Containment l
The radiation environment for qualification of equipment should be based on the normally expected radiation environment over the installed life of the equipment, plus that associated with the mos't severe accident during or following which the equipment must remain functional.
It shou'd be assumed that the accident-related environmental conditions occur at the most critical point of degradation during the installed life of the equipment, which may be at the end of its installed life.
Methods acceptable to the NRC staff for establishing radiation limits for l
the qualification of equipment for BWRs and PWRs are provided in the sample calculations in Appendix B and the following:
(1) The source term to be used in determining the radiation environment l
for equipment qualification associated with a LOCA should consider the most limiting environment associated with the following:
(a) For a LOCA where the primary system cannot be restored,100% of the core activity inventory of noble gases and 50% of the core activity inven-tory cf the halogens should be assumed to be instantaneously released from the funi to the containment. Fifty percent of the cesium activity and 1% of the remaining fission product solids activity inventory in the core should be assumed to be instantaneously released from the fuel to the primary coolant and carried by the coolant to the containment sump.
7
,_,.-.,..._,,r.
e y-
6 (b) For a LOCA where the primary system integrity can be restored, 100% of the core activity inventory of the noble gases, 50% of the core activ-ity inventory of the halogens, and 50% of the core activity inventory of the cesium and 1% of the remaining fission product solids activity inventory should be assumed to be instantaneously released (after an initial time delay) and cir-culated in the primary coolant system. 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.
(2) For all other design basis accidents (e.g., non-LOCA high-energy !ine 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 the total rod activity inventory of iodine and 10% of the total activity inventory 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 fuel rods should be assumed instantaneously released to the primary coolant.
When fuel melting is postulated the activity inventory of the mel'ted 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 inventory, and 1% of the remaining fission product solids 8
1
~*"[
_ ~'"~~~'
"'~~[_
" " ~ ' ~
^
f activity inventory in these elements should be assumed to be instantaneously.
released to the primary coolant.
(3) For a limited number of accident-monitoring instrumentation channels with instrument ranges that extend well beyond that which the selected vari-ables can attain under limiting conditions as specified in Regulatory Guide 1.97,
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs conditions During and Following an Accident," the source term l
should assume an initial release to the containment that considers the fission j
l product release groups associated with grossly melted fuel. An acceptable assumption of fractional 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 l
from the containment 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-leased fission products within various regions of containment and auxiliary strt. tures.
(5) The initial distribution of activity within the containment should be based on mechanistic assumptions. For example, compartmented containments such as some BWRs, 100% of the source should be assumed to be initially con-tained in the drywell. For ice condenser containments, it should be assumed that 100% of the source is initially contained in the lower portion of the containment. The assumption of uniform distribution of activity throughout a compartmented containment at time zero may not be appropriate, 9
f
-e=.
vi.
e
-,g
--g
--w--
---,y---ws-
- y 4. a v,,, - -
ye7%
-,.w.
-,,-m w--
m
I (6) Effects of ESF systems that act to remove airborne activity and redistribute activity within containment, such as containment sprays and con-tainment ventilation and filtration systems, should be calculated using the same assumptions used in the calculation of offsite dose. See SRP Section 15.6.5 (NUREG-0800) and the related sections referenced in the Appendices to that section.
(7) Natural deposition (i.e., plateout) of airborne activity should be determined using a mechanistic model and best estimates for the model param-eters (See Ref. 3, Appendix B). The assumption of 50 percent instantaneous plateout 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 analy-i sis or experiment.
(f,) 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 calculated dose (beta plus gamma) at the equipment location. A gamma source (only) may be used for qualification testing provided analysis or tests indi-cate that the doses and dose rate produce damage similar to the damage that
{
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 a gamma test source.
Plant-specific analysis may be used to justify any reduction in dose or dose rate due to equipment location or shielding.
10
- . = =.
c -
.i l
(9) Shielded components need be qualified only to the gamma radiation dose or dose rate required provided an analysis or test shows that the sensi-tive portions of the component or equipment are not exposed to significant beta
(
radiation dose rates or that the effects of beta radiation heating and secondary radiation have no deleterious effects on component performance.
(lu) Coatings and coverings on electric equipment should be assumed to be exposed to both beta and gamma dose and dose rates in assessing their resistance l
to radiation.
Plateout activity should be assumed to remain on the equipment surface unless the effects of the removal mechanisms such as spray washoff or steam condensate flow can be justifed 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 should not generally be considered exempt from radiation qualification testing. Exemp-tion may be based on qualification by analysis supported by test data or operat-ing experience that verifies that the dose and dose rates will not degrade the operability of the equipmen't 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 inte-grated beta and gamma doses accounting for beta and gamma dose rates equal to or higher than those levels resulting from an analysis that (1) is similar in nature and scope to that included in Appendix B and (2) incorporates appro-t priate factors pertinent to the plant design (e.g., reactor types and power level, containment size).
e.
Environmental Conditions for Outside Containment (1) Equipment that is located outside containment and that could be sub-jected to high-energy pipe breaks as defined in the Standard Review Plan should.
ij u
I 2
~"
,---.,.r-,
' ' Tr;.: ~, ~ ~
L,,_
-.__,-____~~;^~~~Ts
be qualified to the conditions resulting from an accident for the duration required. The techniques to calculate the environmental conditions should employ a plant-specific model based on good engineering judgment.
(2) Equipment located in general plant areas outside containment where equipment is not subjected to a design basis accident environment should be qualified to ae normal and abnormal range of environmental conditions postu-lated to occur at the equipment location.
(3) Equipment not served by environmental support systems within the scope of this guide or served by other systems within the scope of this guide that may be secured during plant operation or shutdown should be qualified to l
the limiting environmental conditions that are postulated for that location, assuming a loss of the environmental support system.
5.
Section 6.3, " Type Test Procedures," of IEEE Std 323-1974 should be supplemented with the following:
a.
Equipment located in' a mild' environment defined in Positions 4.e.(2)
(
and (3) are not required to be qualified by test. The " Design / Purchase" speci-fications that contain a description of the functional requirements of its specific environmental location during normal and abnormal environmental condi-tions and that is supported by a certificate of compliance based on test data and analysis will generally be acceptable. A well-supported surveillance program in conjunction with a good preventive maintenance program should be provided to ensure that such equipment will function for its design life. A aild environment is an environment that would at no time be no more severe thart the environment that would occur during normal power plant operation or during anticipated operational occurrences.
l b.
Equipment located in watertight enclosures should be qualified by testing that demonstrates the adequacy of such protection.
Equipment that could be submerged should be identified and demonstrated to be qualified by l
12 i
,j_1
- '*"'*" 7'- **
- * ' ' ~ ~ ~
g__-
1 7
- --.p --w,y
l=
't, testing that demonstrates seal integrity and functional operability for the duration required. Shortened test periods and analytical extrapolatiorr should be justified.
c.
Equipment located in an area where rapid pressure changes are expected should be qualified by testing that demonstrates that, under the most adverse time-dependent relative humidity conditions (superheated steam followed by saturated steam may be a limiting condition) and c.dverse postulated pressure-transient for the equipment location, the equipment seals and vapor barriers will prevent moisture from penetrating into the equipment to the degree neces-sary to maintain equipment integrity for the length of time tha equipment func-tion is required.
d.
The temperature to which equipment is qualified when exposed to the 1
simulated environment should be defined by temperature readings close to the component being qualified which adequately characterizes tne environment.
e.
Performance characteristics of equipment should be verified before, after, and periodically during testing throughout its range of required oper-ability. Variables indicative of momentary failure, e.g., momentary opening
.of a relay contact, should be monitored continuously to ensure that spurious failures (if any) have been accounted for during testing.
For long-term test-l ing, however, continuous monitoring during periodic intervals may be used if justified.
f.
Chemical spray or.domineralized water spray should be incorporated during simulated event tasting at or near the maximum pressure and temperature conditions that would occur when the spray systems actuate.
i g.
Expected extremes in power supply voltage range and frequency should be applied appropriately during simulated event environmental testing.
h.
Cobalt-60 or Cesium-137 woula be acceptable gamma radiation sources for environmental qualification.
D l'
- .~
.-.z--.,..
l J
6.
In the absence of plant-specific margins, the suggested values in Section 6.3.1.5, " Margin," of IEEE Std 323-1974 may be used as a guide sucject to the following:
a.
Quantified margins should be applied to the design parameters dis-cussed in Position C.4 to ensure 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. The margins should (1) account for uncertainties associated with the use of analytical techniques in deriving environmental parameters, when best estimates methods are used rather than conservative licensing methods, (2) account for uncertainties associated with defining satisfactory performance (e.g., when only a few units are tested), (3) account for variations in the commer-cial production of the equipment, and (4) account for the inaccuracies in the test equipment to assure that the calculated parameters have been adequately enveloped.
b.
Some equipment may b'e required by the design to perform its safety function only 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 />), and, once its function is completed, subsequent failures are shown not to be detrimental to plant safety. Other equipment may not be required to per-form a safety function but must not fail within a short time period into the i~
event, and subsequent failures are also shown not to be detrimental to plant safety. Equipment in these categories shoulo remain functional in the acci-dent environment for a period of at least I hour in ex:ess of the time assumed in the accident analysis. For all other equipment (e.g., postaccident moni-toring, recombiners, etc.), the 10 percent time margin identified in Sec-tion 6.3.1.5 of IEEE Std 323-1974 should be used.
~
14 3 -.
1 i _..
7.
Section 6.3.3, " Aging," of IEEE Std 323-1974 should be supplemented with the following:
a.
Where synergi.stic effects have been identified (e.g., effects result-ing from dose rates in combination with other aging effects and from different sequences of applying qualification test parameters), they should be accounted for in the qualification program.
b.
The expected operating temperature of the equipment under service conditions should be accounted for in thermal aging. The Arrhenius methodology is considered an acceptable 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.
The aging acceleration rate and activation energies used during quali-fication testing and the basis upon which the rate and Ectivation energy was established should be defined, justified and documented.
e.
Periodic surveillance testing under normal service conditions is not considered an acceptable method for ongoing qualification unless the testing includes provisions for subjecting the equipment to the limiting service and environmental 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.
1 g.
The qualified life of the equipment (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 performed, the conservatisms employed in the extrapolation of data, the operating history, and the other methods that may be reasonably assumed.
All assumptions should be documented.
15 I
t
[.'. '
i l
(i) An ongoing program to review surveillance and maintenance records to i
identify age-related degradations should be established.
(j) A component maintenance and replacement schedule that includes con-i sideration of aging characteristics of the installed components should be established.
8.
' 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-1 ence End analysis based on test cata may be used where testing is precluded by the physical size of the equipment or the 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.
9.
Components that are part of equipment qualified as an asse@ly (e.g.,
a motor starter that 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 compo -
'1ould be designed to meet the performance requirements and should be qualified to meet the service conditions specified for the original components.
- 10. Section 8 of IEEE Std 323-1974 discusses documentation.
In addition to the requirements of Section 8, the documentation should include sufficient information to address the required information identified in Appendix C.
A certificate of conformance by itself is not acceptable unless it is accompanied by information on the qualification program, including test data or comparable test csta from equivalent equipment. A record of the qualification should be maintained in a central file to permit verification that each item of electric 16 i:
l
equipment is qualified for its application and meets its specified performance requirements when subjected to the conditions present when it must perform its safety function up to the end of its qualified life.
D.
IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the flRC staff's plans for using this regulatory guide.
Except in those cases in which the applicant proposes an acceptable alter-native method for complying with specified portions of the Commission's regul-tions, the methods described herein will be used in the evaluation of the quali-fication of electric equipment for all operating plants and plants which have not received an operating license subject to the following:
(1) For operating plants, sequence testing as noted in Position C.2 is not required.
(2) For plants t:1at 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 tested only for high temperature, pressure, and steam, equipment may not need to be tested again to include other service conditions such as radiation anc chemical sprays. The qualification of equipment for these service conditions may be established by analysis.
I (3) Regarding aging considerations in equipment qualification, for all plcats that are not committed to the November 1974 issue of Regulatory Guide 1.89/IEEE Std 323-1974, a specific qualified life need not be demonstrated except in the case of equipment that uses naterials that have been identified as being susceptible to significant degrar.'ation due to aging. Component main-tenance or replacement schedules should include considerations of the specific 17 j
V
9 aging characteristics of the component materials. Ongoing programs should exist at the plant to review surveillance and maintenance records to ensure that equipment which is exhibiting age-related degradation will be identified and replaced as necessary. However, the valve operators and the motors shculd be preconditioned by aging prio.r..to testing for those plants that are committed to Regulatory Guide 1.73/IEEE Std 382-1972 and Regulatory Guide 1.40/IEEE Std 334-1971.
(4) Beginning with May 23, 1980, replacement components or spare parts used to replace currently installed equipment or components should be qualf-fied according to this guide unless there are sound reasons to the contrary.
Unavailability or the fact that the component to be used as a replacement is in stock or was purchased prior to May 23, 1980, are among the factors to be considered in weighing whether there are sound reasons to the contrary.
18
[
i i
..._,y
APPENDIX A
!fETHODS FOR CALCULATINC MASS AND ENERGY n m in 4
4 0
l
APPENDIX A METHODS FOR CALCULATIFG MASS AND ENERGY P"mE l
to determine
' Acceptable methods for calculating the mass and energy re ease described in the following:
l (1) Topical Report WCAP-8312A for Westinghouse plants.
Section 6.2.1 of CISSAR System 80 PSAR for Combustion Engineering (2) plants.
(3) Appendix 6A of B-SAR-205 for Babcock & Wilcox plants.
4 (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 Appendix 6B of CISSAR System 80 PSAR for Combustion Engineering 1
(1) plants.
Section 15.1.14 of B-SAR-205 for Babcock & Wilcox plants.
(2)
Same as item (4) above fer General Electric plants.
(3)
(Although this Topical Report WCAP-88 ' for Westinghouse plants.
Topical Report is currently under review, the use of this method is (4)
Reanalysis acceptable in the interim if no entrainment is assumed.
may be required following the NRC staff review of the entrainment model as presently described.)
A-1 S
e
"***4*e De atom
APPENDIX B SAMPLE CALCULATION AND TYPE METHODOLOGY FOR' RADIATION QUALIFICATION DOSE f
e l-
~~
APPENDIX B SAMPLE CALCULATION AND TYPE METHODOLOGY F0F. RADIATION QUALIFICATION DOSE This appendix illustrates the staff model for calculating dose rates and integrated doses for equipment qualification purposes. The doses shown in Fig-ure B-1 include contributions from airborne and plateout radiation sources in the containment and coyer a period of one year following the postulated fis-sion product release. The dose values shown here are provided fer 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 dccermined 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, have been determined using models and assumptions contained in this appendix. This analysis is conservative, and factors in the important time-dependent phenomena related to the action of engineered safety features (ESFs) and natural phenomena, such as iodine plateout, as done in previous staff analyses.
Doses were calculated for a point inside the containment (at the midpoint of the containment) taking sprays and plateout mechanisms into account. The doses presented in Figure B-1 are values for a PWR plant having a containment free volume of 2.5 million cubic feet and a power rating of 4100 MWt.
e B-1 L__
- 1. 0 Basic Assumptions Used in the Analysis Gamma and beta doses and dose rates should be determined for three types of radioactive source distributions:
(1) from activity suspended in the con-tainment atmosphere, (2) from activity plated out on containment surfaces, and (3) from activity mixed in the containment sump water. A given piece of equip-ment may receive a doce contribution from any or all of these sources. The amount of dose contributed by each of these sources is determined by the loca-tion of the equipment, the time-dependent and location-dependent distribution of the source, 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 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.
Theta rulemaking activities also included an examination of fission product releases under degraded core conditions. While the final resolution of the 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 assumptions for equipment qualification.
Based on release estimates in the Rogovin Report (Ref. 1), the staff assumptions for noble gas and iodine releases still appear to be conservative.
Howevert the report estimates that the TMI-2 release contained between 40 and B-2 i
l
~
.o i
60 percent of the Cs-134 and Cs-137 core activity in the primary system water, in the containment sump water, and in auxiliary building tanks. Comparison of i
the integrated dose from the TMI-2 cesium release to the previous staff assumo-tion of "1% solids" shows that assuming "1% solids" may not result in a con-i servative estimate for the radiation exposure for equipment required to func-tion for time periods ex.:eeding thirty days. The staff feels that as a first step toward modificatica of the TID-14844 source term in the direction indi-j cated by the TMI-2 e perience, it may be prudent to factor in a cesium release i
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 activ-ity 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 consist-ent with the findings of recent source term studies reported in NUREG-0772 j
(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
'l j
I and Cs would, in the long tarm, be located primarily in the reactor water 2
and the containment sump water (PWR) or suppression pool (BWR). The staff 4
i recognizes that the revised source terms contained in this report are interim
- i
!.j values and that the conclusions from the report cited above, as well as further i
results from current research efforts in the source term area, should ultimately
> t i
i -!
B-3 s.
i form the basis for any revision of source term assumptions. Any revision of
- f the source term assumptions, such as the incorporation of additional radionu-clides, would be factored into the guide before it is issued as an effective guide.
i 2.0 Assumotions Used in Calculatina Fission Product Concentrations This section discusses the assumptions used to simulate the PWR and BWR j
containments for determining the time-dependent and location-dependent distri-bution of noble gases and fodines airborne within the containment atmosphere, activity plated out on containment surfaces, and activity in the sump water.
The staff used a computer program, TACT, to model the time-dependent behavior of iodine anc noble gases within a nuclear power plant. The TACT code or other equivalent industry codes would provide an acceptable method for l
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.plateout 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.
4 2.1 PWR Dry Centainments 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 and 50 percent of the core inventory of cesium and 1% of the remaining " solid" activity J
inventory were released from the core and carried with the primary coolant directly to the containment sump.
B-4 I.
_ _ _ _ _. l
sr 6
3 b.
The containment free volume was taken as 2.52 x 10 fg. Of this 6
3 volume, 74 percent or 1.85 x 10 ft is assumed to be directly covered by the containment sprays.
(Plants with different contain-ment free volumes should use plant-specific values.)
5 3
c.
6.6 x 10 ft of the containment free volume is assumed unsprayed, which includes regions within the main containment space under the
, containment dome and compartments below the operating floor level.
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 assumed.
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 i
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.
g.
Trace levels of hydrazine was assumed to be 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 the operation B-5
~'
1
of only one spray train 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 i.
Plateout of iodine on contaira.ent 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 value for. the overall elemental iodine plateout constant was
~1 1.23 hr The assumption that 50 percent of the ectivity is instantaneously plated out should not be used.
j The spray removal and plateout process were modeled as competing iodine removal mechanisms.
i k.
A spray removal rate constant (A) for particulate iodine concentra-tion was calculated using the staff's SPIRT program (Ref. 3). The
-1 staff calculated a value of A = 0.43 hr and allowed the removal of l
particulate iodine to continue until the airborne concentration was 4
reduced by a factor of 10. The organic' iodine concentiation in the containment atmosphere is assumed not to be affected by either the containment spray or plateout removal mechanisms.
1.
The sprays were assumed to remove elemental iodine until the instan-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 m
m.
A relatively open (not compartmented) containment was assumed, and the large release was uniformly distributed in the containment.
This is an adequate simplification for dose assessment in a PWR con-tainment and is 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 core inventory of the iodine released is associated with airborne particulate mate-rials and 2 percent of the core inventory of the iodine released formed organic compounds. The remaining 95 o percent remained as elemental iodine. For conservatism this composition was. assumed present at time t = 0.
(These assumptions concerning the iodine form are consistent with those of Regulatory Guides 1.3 and 1.4 when a plateout 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 assumed when calculating the radiation environment following other non-LOCA design basis accidents provided the safety features systems l
are automatically activated as a result of the accident.
B-7 l
I.
=-
l
=
2.2 PWR Ice Condenser Containments The assumptions and methods presented for the calculation of the radia-tion er/ironment in PWR dry containments are appropriate for use in calculat-ing the radiation environment following a design basis LOCA for ice condenser containments with the following modifications:
a.
The source should be assumed to be initially released to the lower containment compartment. The distribution of the activity should be based on the forced recirculation fan flow rates and the transfer rates through the ice beds as a function of time.
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 plateout and spray processes may be assumed provided these removal processes are evalaated using the assumptions consistent with items h through 1 in Section 2.1 and
~
plant-specific parameters.
2.3 BWR Containments The assumptions and methods prest.nted 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 BWRs with the following modifications:
B-8
- e*=r9+
S*.-
1
- =. *G
- ~ "
a.
A decontamination factor (DF) of 10 may be assumed for both the elemental and particulate iodine as the iodine activity passes 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 drywell area and the transfer of activity from these regions via containment leakage l
to the surrounding reactor building volume should be used to predict the qualification levels within the reactor building (sacondary containment).
j c.
Removal of airborne iodine in the drywell or reactor building by both the plateout and the spray process may be assumed provided the effectiveness of these competing iodine remo'al processes are v
evaluated using the assumptions consistent with items h through 1 in Section 2.1 and plant-specific parameters.
d.
The removal of airborne activity from the reactor building by opera-tion of the Standby Gas Treatment System (SGTS) may be assumed.
l 3.0 Model for Calculating the Oose Rate of Airborne and Plateout 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
.4 1
h.
.--e W
- W
- 4
- 9 m
- e g
, e m..m%
a
e f
the containment. The containment was modeled as a cylinder cf equal height and diameter. Containment shielding and internal structures were neglected because this was considered to involve a degree of complexity beyond the scope of the present work. The calculations of Reference 4 indicate that the specific internal shielding and structure would be expected to reduce the gamma doses and dose rates by factors of two or more, depending upon the specific location i
and geometry.
Because of the short range of the betas in air, the airborne beta doses 3
were calculated using an infinite medium approximation. This is shown in Reference 5 to result in only a small error. For beta dose calculations for i
i equipment located on the containment walls or on large internal structures, i
the semiinfinite beta dose model may be used.
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 evaluation.
Finally, all gamma doses were multiplied by a correction factor of 1.3 as suggested in Reference 5 to account for the omission of the contribution from the decay chains of the isotopes.
i 1
4.0 Model for Calculatino the 00.3 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 iodine B-10 2
2
.2..
m e
c, (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).
The dose calculations may assume a time-dependent iodine source. (The difference between the integrated dose assuming 50 percent of the core iodine immediately 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 l
should be calculated including an estimate of the effects of buildup.
J 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 Nuclear Regulatory 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' g
TAB E B-1 SUlftiARY TABI.E OF ESTTMATH 70R TOTE AIRBORNE GAHtfA DOSE CONTRIBUTORS IN CONTALM TO A POINT IN THE CONTAI2RfENT CENTER TI!fE AIRBORNE IODINE AIRBORNE NOBIZ GAS PI.ATIOUT IODINE TOTE DOSE (RRS)
DOSE (R)
DOSE (R)
DOSE (R)
(R) 0.00 0.03 4.82E+4 7.42E+4 1.69E+3 1.24E+5 0.06 8.57E+4 1.391+5 3.98E+3 2.29E+5 0.09 1.09E+5 1.98E+5 7.22E+3 3.14E+5 0.12 1.25E+5 2.51E+5 1.10E+4 3.87E+5 0.15 1.38E+5 3.01E+5
.1.52E+4 4.54E+5 0.18 1.47E+5 3.48E+5 1.96E+4 5.15E+5 0.21 1.55E+5 3.92E+5 2.41E+4 5.71E+5 0.25 1.64E+5 4.49E+5 3.03E+4 6.43E+5 0.38 1.87E+5 6.19E+5 5.05E+4 8.57E+5 0.50 2.03E+5 7.61E+5 6.9CE+4 1.03E+6 0.75-2.36E+5 1.03E+6 1.06E+5 1.37E+6 1.00 2.66E+5 1.26E+6 1.40E+5 1.67E+6 2.00 3.62E+5 2.04E+6 2.61E+5 2.66E+6 5.00 5.50E+5 3.56E+6 5.40E+5 4.65E+6 j
8.00 6.63E+5 4.38E+6 7.47E+5 5.79E+6 l
24.0 1.01E+6 6.26E+6 1.45E+6' 8.72E+6 60.0 1.31E+6 7.16E+6 2.10E+6 1.06E+7 96.0 1.45E+6 7.56E+6 2.39E+6 1.14E+7 l
192.
1.68E+6 8.29E+6 2.86E+6 1.28E+7 298.
1.85E+6 8.76E+6 3.19E+6 1.38E+7 394.
1.95E+6 8.85E+6 3.41E+6 1.42E+7 560.
2.07E+6 9.06E+6 3.64E+6 1.48E+7 720.
2.13E+6 9.15E+6
- 3. 76E+6 1.50E+7 888.
2.16E+6 9.191+6
- 3. 83E+6 1.52E+7 1060 2.18E+6 9.21E+6 3.87E+6 1.53E+6 1220 2.19E+6 9.21E+6 3.89E+6 1.53E+7 1390 2.20E+6 9.21E+6
- 3. 9CE+6 1.53E+7 1560 2.20E+6 9.22E+6
- 3. 91E+6 1.53E+7 1730 2.20E+6 9.22E+6 3.91E+6 1.53E+7 1900 2.20E+6 9.22E+6 3.92E+6 1.53E+7 2060 2.20E+6 9.22E+6 3.92E+6 1.53E+7 2230 2.20E+6 9.22E+6 3.92E+6 1.53E+7 2950 2.20E+6 9.23E+6 3.92E+6 1.54E+7 3670 2.20E+4 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 i
5830 2.20E+6
- 9. 25E+6 3.92E+6 1.54E+7.
6550 2.20E+4 9.26E+6 3.92E+4 1.54E+7 7270 2.20E+4 9.26E+4 3.92E+6 1.54E+7 8000 2.20E+4 9.27E+6 3.92E+6 1.54E+7 8710 2.20E+6 9.2,8E+6 3.9:E+6 1.54Z*7 l
TOTE 1.54E+7 l*
s-12 i
l
= - _.. _;-
= -
___; _ _ y
TAEE ' B-2 St22 FART TABIE OF ESTI!!ATIS FOR TOTAL AIRBORNE BETA DOSE CONTRIBUTORS IN CONTADQfENT TO A POINT IN THE CON *N N TI!!E AIRBORNE IODINE AIRBORNE NOBIE GAS TOTE DOSE (HRS)
DOSE (RADS)
- DOSE (RADS) *
(RADS)
- 0.00 0.03 1.47E+5 5.48E+5 6.95E+5 0.06 2.62 +5 9.86E+5 1.25E+6 E
0.09 3.33 +5 1.35E+5 1.68E+6 E
0.12 3.83E+5 1.65E+6 2.03E+6 0.15 4.20E+5 1.91E+6 2.33E+6 0.18 4.49E+5 2.14E+6 2.595+6 0.21 4.73E+5 2.35E+6 2.82E+6 0.25 5.00E+5 2.60E+6 3.10E+6 0.38 5.67E+5 3.30E+6 3.87E+6 I
0.50 6.15 E+5 3.86E+6 4.48E+6 0.75 7.13 E+5 4.895+6 5.60E%
1.00
- 8. 00 E+5 5.81E+6 6.61E+6 2.00
- 1. 07 E+6 9.02E+6 1.01E+7 5.00 1.58 E+6 1.65E+7 1.81E+7 8.00 1.88E+6 2.20E+7 2.39E+7 24.0
- 2. 87 E+6 4.08E+7 4.37E+7 60.0
- 3. 89 E+6 6.15E+7 6.54E+7 96.0 4.37 5+6 7.48E+7 7.92E+7 192.
5.14 Er&6 1.00E+8 1.05E+8 298.
5.64 +6 1.17E+8 1.23E+8 E
394.
5.99 +6 1.25E+8 1.31E+8 E
560.
6.34 E+6 1.34E+8 1.40E+8 720.
6.53 +6 1.39E+8 1.46E+8 E
888.
- 6. 63 E+6 1.42E+8 1.49E+8 1060 6.69 E+6 1.44E+8 1.51E+8 1220 6.73 E+6 1.45E+8 1.52E+8 1390 6.75 E+6 1.47E+8 1.54E+8 1560 6.76 E+6 1.49E+8 1.56E+8 1730
- 6. 76 E+6 1.51E+8 1.58E+8 1900 6.76 E+6 1.52E+8 1.59E+8 2060 6.76 E+6 1.54 +8 1.61E+8 E
2230 6.77 E+6 1.55E+8 1.62E+8 2950 6.77 E+6 1.62E+8 1.69E+8 3670 6.77 E+6 1.69E+8 1.76E+8 4390 6.77 E+6 1.76E+8 1.83E+8 5110 6.771+6 1.83E+8 1.90E+8 5830 6.77 E+6 1.89E+8 1.96E+8 6550 6.77 E+6 1.96E+8 2.03E+8 7270 6.77 E+6 2.03E+8 2.10E+8 8000 6.77 E+6 2.09E+8 2.16E+8 8710 6.77 E+6 2.16E+8 2.23E+8 TOTE 2.23E+8
- Dose conversion factor is based on absorption to cissue.
B-13 4:I'
=.:
. =_
V Cp im l-
^
=
Z Cp W
e
=
Y c
=
^z w
e MW CO F
c 4%
Cl
- 8 I
QC v
=
vW a
=
~
W CO
=
NO
=
CO O
m a
a GE
=
HE F
0 "S
WC
=*
e mg
=
4 8 I w
4
=
>C O a
1 e
in o E
=
u O =
C *
=
z &
=
v d
=
r
=
=
a=e
>=
=
w o
r e
=g a
=
a
=a o
=
"J c
=
cwo
=
Aw
=
e
=
-4 b
=
a e
t vs *
=
"N 7
co 3
On CD f%
W 10 @.
e i........
i.......
i........
ws m
CD tlD
=
.=
w
.=
MEbd@NSbdO OO@d B-14
...o
.-%+.~*+
- + -.,. - - -
g
REFERENCES i
1 1.
Mitchell Rogovin, George T. Frampton, Jr., et al, "Three Mile Island--
a report to the Commissioners and to the Public" NUREG/CR-1250, Volume II, Part 2.
1 2.
NUREG-0772, " Technical Basis for Estimating Fission Product Beaavior During LWR Accidents."
3.
A. K. Postma, R. R. Sherry and P. Tam, " Technological Bases for Models of Spray Washout and Airborne Contaminants in Containment Vessels," USNRC Report NUREG/CR-0009, November 1978. Available for purchase from National Technical Information Service, Springfield, Virginia 22161.
4.
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 Society, 1976. Available from tschnical libraries.
5.
M. J. Kolar and N. C. Olson, " Calculation of Accident Doses to. Ecuipment Inside Containment of Power Reactors," Vol. 22, pp. 808-809 in Transac-tions of the American Nuclear Society, 1975. Available from technical libraries.
BIBLIOGRAPHY A.K. Postma and R. Zavadoski, " Review of Organic Iodide Formation Under Acci-dent Conditions in Water Cooled Reactors," WASH-1233, October 1972, pp. 62-64.
Available for purchase from the National Technical Information Service, Springfield, Virginia 22161.
D. C. Kocher, ed., " Nuclear Decay 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.
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 libraries.
R. A. Lorenz, J. L. Collins, and A. P. Malinauskas, " Fission Product Source l
Terms for the LWR Loss-of-Coolant Accident: Summary Report," USNRC Report NUREG/CR-0091, May 1978. Available for purchase from the National Technical l
Information Service, Springfield, Virginia 22161.
l I
l B-15
~
g -
-s
.,-.,n-,.,
-.c.
., w
,n p -
. +. - - _.
y G
APPENDIX C QUALIFICATION DOCUMENTATION FOR ELECTRIC EQUIPMEh"I e
9 i
\\1
APPENDIX C QUALIFICATION DOCUMENTATION FOR ELECTRIC EQUIPMENT In order to ensure that an environmental qualf fication program conforms with General Design Criteria 1, 2, 4, and 23 of Appendix A and Sections III and II of Appendix B to 10 CFR Part 50, and to the national standards mentioned in-Part II, " Acceptance Criteria" (which includes IEEE Std 323), contained in the Standard Review Plan, Section 3.11, the following information on the qualifica-tion program.is required for electric equipment within the scope of this guide.
l 1.
Identify all electric equipment within the scope of this guide and provide the following:
a.
Type (functional designncion) b.
Manufacturer c.
Manufacturer's type number and model number The equipment should include the following, as applicable:
(1) Switchgear (2) Motor control centers (3) valve operators (4) Motors (5) Logic equipment
-(6) Cable (7) Connectors (8) Diesel generator control equipment (9) Sendors (pressure, pressure diff erencial, temperature, neutron, and other radiation)
~
(10) Limit switches (11) Heaters (12) Fans (13) Control boards (14) Instrument racks and panels (15) Electric penetrations (16) Splices (17) Terminm1 blocks 1
2.
Catergorize the equipment identiff ed in item 1 above into one of the following categories:
a.
Equipment that will experience the environmental conditions of design basis accidents for which it must function to mitigats such accidents and that will be qualified to demonstrate operability in the accident environment for the time required for accident mitigation with safety margin to failure.
b.
Equipment that will experience environmental conditions of design basis accidents through which it need not function for mitigation of such accidents but through which it must not fail in a manner detrimental to plant safety or accident mitigation and that will be qualified to demon-strate the capability to withstand any accident environment for the time during which it must not fail with safety margin to failete.
f C-1 I
. -. ]
. =.
,, =.,
J k
Equipment that will experience environmental conditions of design c.
basis accidents through which it need not function for sicisation of such accidents and whose failurs (in any mode) is daened not detrimental to plant safety or sccident mitigation and need not be qualified for any accident environment, but win be qualified for its nonaccident service environment.
d.
Equipment thae will not experiesca environ==ae=1 conditions of design basis accidents and that win be qualified to demonstrata operability under the expected. extremes of its.nonaccident servica environment. This equipment would nomaany be located outside the reactor come=4n=ene-3.
For each type of equipment in the' categories of equipment listed in item 2 above, provide separately the equipment design specification requirements, including:
The system safety function requirements.
a.
I b.
An environmental envelope as a function of time that includes an aztreme parasetars, both marimum and minimum values, expected to occur during plant shutdown, normal operation, abnormal operation, and any design basis event (including 10CA and M513), including post-event conditions.
Time required to fulfin its safety function when subjected to any c.
of the extremes of the env'ironment envelope specified above.
d.
Technical bases should be provided to justify the placement of each E7P8 of equipment,in categories 2.b and 2.c.
l 4.
Provide the qualification test plan, test setup, test procedures, and acceptance critaria. for at least one of each group of equipment of item 1.d as appropriate to the category identified in item 2 above. If any method other than type tasting was used for qualification (operating experience, analysis, combined qualifiestion, or ongoing qualification),
describe the method in safficient detail to permit evaluation of its j
adequacy.
I 5.
For each category of equipment identified in item 2 above, state the actual qualification envelope simulated during testing (defining the duration of the bestile environment and the margin in escass 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.
6.
A summary of test results that demonstrates the adequacy of the qualification program. If analysis is used for qualification, justification of an analysis assu=ptions sust be provided.
7.
Identification of the qualification documents that contain detailed supporting information, including cast data, for items 4, 5, and 6.
i i
C-2 t
i:
"*e-
s t
~
APPENDIX 0.
TYPICAL EQUIPMENT / FUNCTIONS FOR ACCIDENT MITIGATION i
. I
. - s.
APPENDIX D TYPICAL EQUIPMENT / FUNCTIONS FOR ACCIDENT MITIGATION Engineered Safeguard Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power l
Emergency Core Cooling Containment Heat Removal t
Containment Fission Product Removal Containment Combustible Gas Control Auxiliary Feedwater Containment Ventilation Containment Radiation Monitoring Control Room Habitability System-(e.g., HVAC, Ritdiation Filters)
Ventilation for Areas Containing Safety Equipment i~ ;
Component Cooling Service Water Emergency Systems to Achieve Safe Shutdown Postaccident Sampling and Monitoring 2 Radiation Monitoring 2 l
Safety-Related Display Instrumentation 2 1These systems will differ for PWRs and BWRs and for older and newer plants.
In each case the system features that allow for transfer to recirculation cooling mode and establishment of long term cooling with boron precipitation control are to be considered as part of the system to be evaluated.
e i
D-1 l
r n,
e n
s
- p e
i I
r
~
~
2More specific identification of these types of equipment can be found in the i
plant emergency procedures and in Tables 1 and 2 of Regulatory Guide 1.97, i
Categories 1 and 2.
t i
F r
i 1
N t
j l.
}
\\
I l-1 1
i i
l J
A 4
i i
4 4
i i
l i
i
.I e
e 9
4 4
t-
. i f:
D-2 1:
1i:
ho
^
w.
4 A#Y s.
VALUE/ IMPACT STATEMENT FOR REVISION 1 TO REGULATORY GUIDE 1.89. " ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS"
~
Background
Regulatory Guide 1.89, " Qualification of Class 1E Equipment for Nuclear Power Plants," is being revised to reflect current staff position on equipment qualification.
NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electric Equipment," was issued for public cannent in December 1979.
Subsequent to its issuance for canment, the Commissioners (see " Memorandum and Order", dated May 23,1980) directed the staff to use NUREG-0588 along with
" DOR Guidelines for Evaluating Qualification of Class 1E 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.
1 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 equivalent to the staff position and guidance contained in the proposed revised version of hUREG-0588 which is based on consideration of public comments and lessons learned from TMI-2 in source term definition.
1 Substantive Changes and Their Value/Imoact 1.
Regulatory Position C.2, which provided radiological source terms for equipment qualification tests, was deleted and the following positions were added:
i em+-
w - <
w wmmme gas esp m. e om,,
emm w-aw
IN RESPONSE, PLEASE REFER TO:
M811110
~
- pa moq'o UNITED STATES
~
NUCLEAR REGULATORY COMMISSION n
5, 4,I waswiuoTou. o.c. noses ACTION - Denton Cys:
Dircks t.
'% *.. + /*
November 18, 1981 Cornell Rehm omca or twa Stallo ssenstany M,inogue y Aggrawal Vollmer MEMORANDUM FOR:
William J. Dircks, Executi Director Rostoc::y for Operations Shapar DeYoung FROM:
Samuel J. Chilk, Ser.:reta Michelson 7
SUBJECT-STAFF REQUIREMENTS BRI ON SECY-81-504, EQUIPMENT Q ICATION PROGRAM PLAN, AND SECY-81-603/603A, PROPOSED RULEMAKING, " ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS," 2:00 P.M.,
TUESDAY, NOVEMBER 10, 1981, COMMISSIONERS' CONFERENCE ROOM, DC OFFICE (OPEN TO PUBLIC ATTENDANCE)
The Commission
- was briefed by staff on the subject staff papers relating to equipment qualification.
The Commission requested:
a.
that staff rewrite SECY-81-6'03 b'ased on Version 1 with the deadline for equipment qualification stated therein; b.
that the first sentence of paragraph (g) be revised based on Version 2 to require identification of electric equipment already qualified; c.
that the new draft include a statament to the effect that the rule does not cover seismic and dynamic qualification of equipment; and d.
that the new rule be made effective upon publication of the rule.
(RES)
(Subsequently, staff provided the Commission with the l
revised paper.)
Commissioner Ahearne requested that staff brief him on the various seismic and dynamic qualification requirements and how they've been applied to operating reactors over a number of years.
(NRR) (SECY SUSPENSE:
To be determined)
Reffd Cff. EDG'
~
- Commissioner Gilinsky was not present.
- y'*
.,..=....
b1L
/
h f01
.I,
f i
l 2
N The Commission requested that SECY-81-504 he redrafted based on the new version of SECY-81-603.
h; f tr d-e)
(NRR) (SECY SUSPENSE:
T /16l1%
/
commissioner Ahearne requested written responses to the questions in his memorandum of November 17, 1981. [ M )
(NRR) (SECY SUSPENSE:
TO L. Cot.e d ned}
"A/r/
e cc:
Chairman Palladino Commissioner Gilinsky Commissioner Bradford Commissioner Ahearne Commissioner Roberts Commission Staff Offices Public Document Room 9
I l
l h
. _..