ML20034F586
| ML20034F586 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 02/25/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9303040028 | |
| Download: ML20034F586 (16) | |
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GE Nuclear Energy
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February 25,1993 Docket No. STN 52-001
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Chet Posiusny, Senior Project Manager i
Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation l
Subject:
Submittal Supporting Accelerated AllWR Review Schedule - Resolution of.
Outstanding Items of Section 3.11
Dear Chet:
Enclosed are SSAR markups of selected portions of Section 3.11 supporting the l
resolution of outstanding items.
it should be noted that this markup includes Pages 31.3-10 and 31.3-16 of an earlier submittal and the proprietary affidavit under which they vrere originally issued is applicable.
r Please provide copies of this transmittal to Butch Burton.
Sincerely, f
JL%
J k Fox Advanced Reactor Programs cc: Norman Fletcher (DOE)
Bernie Genetti (GE)
J1 % 39 Z &$h 9303040028 930225 PDR ADOCK 05200001 PDR
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A / af 3 MN nA6xcre
;tandard Plant m,3 Environmental parameters include temperature, analytical techniques in the derivation of pressure, relative humidity, and neutron dose environmental parameters, the number of units rate and integrated dose. Radiation dose for tested, production tolerances, and test gamma and beta data for both normal and accident equipment inaccuracies conditions will be provided by the COL applicant in accordance with the requirements in Subsection The environmental conditions shown in the 12.2.3.1. The radiation requirements are site Appendix 31 tables are upper-bound envelopes specific documentation owing to the need to model used to establish the environmental design and r
specific equipment which is applicant quali-fication bases of safety-related l determined. The HVAC detailed modeling and the equipment. The upper bound envelopes indicate evolving considerations in the area of accident that the zone data reflects the worse case source terms are expected to generate expected environment produced by a compendium of l significantly differing radiation requirments. accident conditions. Estimated chemical Where applicable, these parameters are given in environmental conditions are also reported is terms of a time. based profile.
Appendix 31.
The magnitude and 60. year frequency of occur-3.11.2 Qualification Tests and Analyses rence of significant deviations from normal plant environments in the zones have insignificant Safety-related electrical equipment that is effects on equipment total thermal normal aging located in a harsh environment is qualified by or accident aging. Abnormal conditions are test or other methods as described in IEEE 323 overshadowed by the normal or accident conditions in the Appendix 31 tables.
Margin is defined as the difference between the most severe specified service conditions of the plant and the conditions used for qualification.
Margins shall be inc!uded in the qualification parameters to account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance. The environmental conditions shown in the Appendix 31 tables do not include margins.
Some mechanical and electrical equipment may be required by the design to perform an intended safety function between minutes of the occurrence of the event but less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the event. Such equipment shall be shown to remain functional in the accident environment for a um period of at leasgbosr'in excess of the time i
ygg assumed in the acendent analysis unless a time margin of less thangt hour can be justified.
i Such justification wil include for each piece of equipment: (1) consideration of a spectrum of breaks; (2) the potential need for the equipment later in the event or during recovery operations; (3) detemination that failure of the equipment after performance of its safety function will not be detrimental to plant safety or mislead the operator; and (5) determination that the margin apphed to the minimum operability time, when combined with other test margins, will account for the uncertainties accociated with the use of 3.11 11.1 Amendment 24
.I
Ar S&3 ABWR
%ston Standard Plant ac. n applicable locations. Alternatively, actual such locations will produce the maximum critical midti-support excitation effeas may be taken responses of Ihe components. In-equipmcnt into account by performing a multi. support response spectra from time. history response will be excitation analys2s.
generated and be in accordance with tbe requirement specified in Paragraph 3K.S.2(a)(6).
)
(b) When determining stresses, the effects of relative seismic support movements will be 3K.4.2.4.2 Qualification Determination considered. When these effects are consi-dered significant, they may be obtained by The equipment type will be considered qualified performing a static structural analysis of the by demonstrating that the equipment Nrformance system, including anchor movements. Such will meet or exceed its specified values for the most effects (which are secondary) will be severe environment or sequence of environments combined with primary (inertial) effects specified during the qualified life. An important l
using the SRSS, step in this process will be tbc determination that the qualification to the requirements adequately 3 K.4.2.4.1.4.6.3 Time History Analysis envelops the equipment applications.
Time history analysis will be performed when 3 K.4.2.5 Combined Qualification conditions arise invalidating the response spectrum method of analpis due to nonlinear phenomena, or Equipment may be qualified by type test, when generation of in-equipment response spectra analysis, previous operating crperience, or any or a more exact result is desired. To integrate or combination of these three methods.
differentiate, the analysis will be done by an applicable numerical integration technique. The 3K.4.2.6 On-going Qualification largest time step used in the analysis mill be 1/10 of the period of the highest significant mode of Some equipment may have a qualified life less vibration of the equipment. The dynamicinput will than the design life of a nuclear power generating be the time history motion at the equ;pment support station. The qualified life may be extended by location. For equipment supported at several installing additional equipment of the same type in locations, the responses will be determined by locations where senice conditions equal or exceed simultaneous excitations using appropriate time those of the equipment to be qualified, remosing history input at each support location. The scaled them after a planned period less than the presiously time interval will be varied as per Paragraph qualified life and subjecting them to a type test 3K.5.2(a)(6). If the equipment frequency is within qualification program. This test would include the range of the supporting structure, then a time additional accelerated age conditioning, dynamic, interval will be chosen such that the peak of the and DBE tests. Completion of this type test enends response spectrum shall be at the equipment the qualified life of the installed equipment by the resonance frequency. The total time interval range length of time simulated during age conditioning.
will be provided with the time history.
This procedure may be repeated until the qualified life equals the required installed life of the 3 K.4.2.4.1.4.6.4 Generation of In-Egalpment equipment or the equipment is to be replaced r
Spectra before its qualiSed life is exceeded.
As a part of the dynamic qualification of 3K.4.2.7 Margins g,y equipment,in-equipenent response spectra may be y'g g generated to quaEfy components of the equipment Margin is defined as the difference between the dynamically. In-equipment response spears will be most severe specified service conditions of the plant obtained at critical locations of the components from and the conditions used for qualification. Margins time. history analysis of the equipment or, where will be included in the quali6 cation parameters to appropriate methods are available, by response account for normal variations in commercial spectra analysis. The in-equipment qualification plan production of equipment, reasonable errors in shall identify the locations at which in-equipment defining satisfactory performance.
response spectra will be generated and will prose that the in-equipment response spectra generated at L -12 Amendment
P'* 3o43 ABWR Standsyd Plant
- [
Wr;pns will be applied :s the specified service conditions regardless of the qualification method selected. The specific (quantified) margms ap; ~;ed will be documented for each phase of the qu ifi-cation. The levels of margin provided in 1 ible 3K.4 2 are considered appropriate for r ost applications. Other margins may be used if jus:Jed as adequate for the situation. h all cases the margins will be documented. Negative factors will be applied when lowering the value of the service condition increases the severity. The application of margin to the age-conditioning of equipment will only consider, and conservatively account for, any uncertsir. ties in the process of xceleration.
Some mechanical and electrical equipment may be required by design to perform an intended safety function between minutes of the occurrence of the event but less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the event. Such equipment will be shown to remain functional in the accident environment for period,4 t;least one hour in excess of the time auumed in the accident nalysis unless a time margin of less than one hour can be justified. Such justification will include for each piece of equipment:(1) conuderation of a spearum of breaks:(2) the potential need for the equipment later in the event or during recovery operations; (3) a determination that failure of the equipment after performance of its safety function will not be detrimental to plant safety or mislead the operator; and (5) determination that the margin applied to the minimum operability, ti ne, when combined with other test margins, will account for the uncertainties associated with the use of analytical tech.%ues in the derivation of environmental parameters, the number of units tested, production tolerances, and test equipment inaccuracies.
7 3K W Amendment r=
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kid 2
.ABWR 7
23A610GAE Standard Plant m,
and permitted by 10CFR50.49(f) (Reference 1).
3.11A Loss or Heating, Ventilating, and Air' i
Equipment type test is the preferred method of Cmditbning l
qualification.
To ensure that loss of heating, ventilating, Safety-related mechanical equipment that is and air conditioning (HVAC) systeu does not located in a harsh environment,is qualified by adversely affect the operability of safety-analysis of materials data which are generally related controls' and clectrical equipment in based on test and operating experience.
buildings and areas served by safety-related l
HVAC systems, the HVAC systems serving these The qualification methodology is described in.,, areas meet the single-failure criterion.
p c
N'g*).) $eport on GE'sCenvironmental qualificationdetail_in the/SRC approved licensin systems including the detailed safety evalu-l
~
4#h'rcsses co'mpliance with the app %a_ iso add-
/nAgM prograngReference 23. This ations. The loss of ventilation calculations
'l licable portions oft are based on maximum heat loads and consider I
the General Design Criteria of 10CFR50, Appendix operation of all operable equipment regardless j
A, and the Quality Assurance Criteria of 10CFR50, of safety classification.
Appendix B. Additionally, the% describes Apfendfx l
conformance to NUREG-0588 (Reference 3), and 3.11.5 Estimated Chemical and Radiation Regulatory Guides and IEEE Standards referenced Environment 1
in Section 3.11 of NUREG-0800 (Standard Review l
Plani 3.11.5.1 Chemical Environment
(
Mild environment is that which, during or Equipment located in the containment drywell j
after a design basis event (DBE, as defined in and wetwell is potentially subject to water l
Reference 2), would at no time be significantly spray modes of the RHR system. In addition, more severe than that which exists during normal, equipment in the lower portions of the contain-I test and abnormal events.
ment is potentially subject to submergence. The chemical composition and resulting pH to which safety-related equipment is exposed during j
normal operation and design basis accident conditions is reported in Appendix 31.
l The COL applicant will require vendors of Sampling stations are provided for periodic I
equipment located in a mild environment to submit analysis of reactor water, refueling and fuel a certificate of compliance certifying that the storage pool water, and suppression pool water i
equipment b.ss been cualified to assure its to assure compliance with operational limits of I
required saf ety-related function in its the plant technical specifications.
I applicable environment. This equipment is
}
qualified for dynamic loads as addressed in 3.11.5.2 Radiation Environment j
l Sections 3.9 and 3.10. Further, a surveillance i
and maintenance program will be developed to Safety-related systems and components are l
ensure equipment operability during its designed designed to perform their safety-related i
I life. (See Subsection 3.11.6).
function when exposed to the normal operational i
a radiation levels and accident radiation levels.
l 3.11.3 Qualification Test Results i
l Electronic equipment subject to radiatioa offN The results of qualification tests for exposure in excess of 1000 R and mechanical 3.//, J-f safety-related equipment will be documented, equipment in excess of 10,000 R will be maintained, and reported as mentioned in qualified in accordarice with Reference 1.
.j Subsection 3.11.6.
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d 1!
Amendment 24 3.11 2 I
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- ABM mst $ ~*
- Standard Plant nry a The normal operational exposure is based on the Non-safety-related control systems subjected radiation sources provided in Chapter 12.
to adverse environments will be evaluated for safety implications to safety-related protective Radiation sources associated with e DBA and functions, and equipment wetting and flooding developed in accordance with NUREG-0588 above the flood level will be addressed in (Reference 3) are provided in Chapter 15.
accordance with Subsection 3.11.1.
Integrated doses associated with normal plan 3.11.7 References operation and the design basis accident condition for various plant compartments are described in (1) Code of Federal Regulations, Title 10, Appendix 31.
Chapter I, Part 50, Paragraph 50.49, Environmental Qualification of Electric 3.11.6 COL License Information Equipment Important to Safety for Nuclear Power Plant.
3.11.6.1 Environmental Qualification Document D e /c fe /
(2) encral Electric environmcotal Qualification The EOD shall be prepared summarizing the Program,NEDE-24326-1 P, Proprietary qualification results for all safety-related ocument, January 19R3.
equipment. The EOD shall include the 11.llowing:
(3) Interim Staff Position on Environmental (1) The test environmental parameters and the Qualification of Safety-Related Electrical methodology used in qualify the equipment Equipment, NUREG-Of 88.
l located in mild and harsh environments shall be identified.
(2) A summary of environmental conditions and qualified conditions for the saf.ty-related equipment located in a harsh environment zone shall b presented in the system com-ponent evaluation work (SCEW) sheets as described in Table 1-1 of GE's environmental qualification program (Reference 2). The SCEW sheets shall be compiled in the EOD.
(3) Equipment gama and beta radiation dose data for both normal and accident conditions will be provided in accordance with the re quire m ent s of Subsection 12.2.3.1.
3.11.6.2 Environrnental Qualification Records The results of the qualification tests shall be recorded and maintained in an auditable file.
i I
3.11.63 Surveillance, Maintenance arid Experience Information The COL applicant will require vendor equipment certificates of qualification compliance and will develop a surveillance and maintenance program in accordance with Subsection 2
3.11.2.
3 11-3 Amendment 24 l
oi
ty 1 of a
'ABWR 6me Standard Plant erv n Environmental parameters include temperature, analytical techniques in the derivation of pressure, relative humidity, and neutron dose environmental parameters, the number of units rate and integrated dose. Radiatios dose for tested, production tolerances, and test gamma and beta data for both normal and accident equipment inaccuracies conditions will be provided by the COL applicant in accordance with the requirements in Subsection The environmental conditions shown in the 12.2.3.1. The radiation requirements are site Appendix 31 tables are upper-bound envelopes specific documentation owing to the need to model used to establish the environ = cia.! design and specific equipment which is applicant quali-fication bases of safety-related l determined. The HVAC detailed modeling and the equipment. The upper bound envelopes indicate evolving considerations in the area of accident that the zone data reflects the worse case source terms are expected to generate expected emironment produced by a compeniium of cpfst l significantly differing radiation requirments. accident conditions. Estimated che nical pyr.q Where applicable, these parameters are given in environmental conditions are also reported is terms of a time-based profile.
Appendix 31.
CEF MU The magnitude and 60-year frequency of occur-3.11.2 Qualification Tests and Analyses F M PM rence of significant deviations from normal plant f
environments in the zones have insignificant Safety-related electrical equipment that is effects on equipment total thermal normal aging located in a harsh environment is qualified by or accident aging. Abnormal conditions are test or other methods as described in IEEE 323 overshadowed by the normal or accident conditions in the Appendix 31 tables.
Margin is defined as the difference between the most severe specified service conditions of the plant and the conditions used for qualification.
Margins shall be included in the qualification parameters to account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance. The environmental conditions shown in the Appendix 31 tables do not include margins.
Some mechanical and electrical equipment may be required by the design to perform an intended safety function between mirutes of the occurrence of the event but less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the event. Such equipment shall be shown to remain functional in the accident environment for a period of at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in excess of Ibc time assumed in the accident analysis unless a time margin of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can be justified.
Such justification will include for each piece of equipment: (1) consideration of a spectrura of breaks; (2) the potential need for the equipment later in the event or during recovery operations; (3) detemination that failure of the equipment after performance of ita safety function will not be detrimental to plant safety or mis! cad the operator; and (5) determination that the margin applied to the minimum operability time, when combined with other test margins, will account for the uncertainties accociated with the use of Amencment 24 3 11-11.1
' d ceneral saane company MkN PROPRIETARYINFORMATION 23A6100AE l
Standard Plant -
cu m.
4 i
Table 31.3 9 i
Radiation Environment Conditions Inside Primary Containment Vessel Plant Normal Operating Conditions (b) Radiation environment II Number Plant Zone / Typical Operating dose rate Integrated done Equipment Gamma Beta Neutgon and Neutron fluence (R/h) (R/h) (N/cm*-sec)
Gamma Beta Neug
(
(R)
' (R). - (N/cm ) '
- i b-1 Upper drywell area
'-if 1x10 l
l
[ Fig's.1.2-3/ 5.1-3]
I3 b-2 Upper area oflower 2x10 5x10 dryweii l
[I~y(s.1.2-3a/ 5.1-3]
]
I b-3 Lower area oflower 1x10 3x10 drywell l
5
[ Fi g's. 1.2-3 b /
11.2-2]
i f
b-4 Wetwell area (sup.
8x10 2x10 pression pool and air space)
[ F i g 's. 1.2 - 3 c /
6.2-39, 7.6-11]
r Notes:
(1) Operating dose rate is at 100% ratedponer and awayfrom the raduttson source.
[
(2) in*egrated dose means the integrated mlue over 60,) tars.
(3) The gamma and beta doses willbe provided m acceptable r""on source terms become definf by the applicant referencing the ABWR design in accordesce with the requarements ofSubsecdon 12.2.3.1.
l 1
P i
l 4
i Amendment 21 31310
ABM ussimie Standard Plant Rev s The normal operational exposure is based on the Non-safety-related control systems subjected radiation sources provided in Chapter 12.
to adverse environments will be evaluated for safety implications to safety related protective Radiation sources associated with the DBA and functions, and equipment wetting and flooding developed in accordance with NUREG-0588 above the flood level will be addressed in (Reference 3) are provided in Chapter 15.
ac:ordance with Subsection 3.11.1.
Integrated doses associated with normal plant 3.11.7 References operation and the design basis accident condition for various plant compartments are described in (1) Code of Federal Regulations, Title 10, appendix 31.
Chapter I, Part 50, Paragraph 50.49, Environmental Qualification of Electric 3.11.6 COL License Informr. tion Equipment important to Safety for Nuclear Power Plant.
3.11.6.1 Environmental Qualification Document e/g[g c/
(2) General Electric t.nvironmentai Quaiihcatic &
The EOD shall be prepared summarizing the Program,NEDE-24326-1-P, Proprietary qualification results for all safety-related Document, January 1983.
equipment. The EOD shall include the following:
(3) Interim Staff Position on Environmental (1) The test environmental parameters and the Qualification of Safety-Related Electrical methodology used to qualify the equipment Equipment, NUREG-0588.
l located in mild and harsh environments shall be identified.
(2) A summary of environmental conditions and qualified conditions for the safety-related equipment located in a harsh environment zone shall be presented in the system com-ponent evaluation work (SCEW) sheets as described in Table I-I of GE's environmental qualification program (Reference 2). The SCEW theets shall be compiled in the EOD.
(3) Equipment gama and beta radiation dose data N
for both normal and accident conditions will 7,11 3 - 2 be provided in accordance with the requirements of Subsection 12.2.3.1.
3.11.6.2 Environmental Qualification Records The results of the qualification tests shall be recorded and maintained in an auditable file.
3.11.6.3 Surveillance, Maintenance and Experience Information The COL applicant will require vendor equipment certificates of qualification compliance and will develop a surveillance and maintenance program in accordance with Subsection 3.11.2.
3.11-3 Amendment 24
9 f
General Dectric Company i
~ AB.M PacPareTravinroRurTios 23461ooie-l Standard Plant ca-m wa I
Table 31.3-15 j
Thermodynamic Environment Conditions Inside Reactor Building (Secondary Containment)
Plant Accident Conditions (a) Pressure, temperature and relative humidity i
Plant Zonefrypical Equipment 100 100' 66 66 j
Control rod drive hydraulic Temperature ( C} )
system (scram etc. of hydrau-Pressure (Kg/cm g 0.035 0.035 0.035 0.
lic control unit) [ Fig's.1,2-4 Humidity (%)
Steam Steam 100 90 max
/4.6-8 Time (2) 1(b) 6(h) 12(h) 100(day)
{
171 100 66 66 Temperature ( C} )
MS isolation valve (1) -
Pressure (Kg/cm g 0.035 0.035
- 0.035 0
j MS drain isolation valve Nitrogen line isolation valve Humidity (%)
Steam Steam 100 90 max
-i (1),(4)
Time (2) 1(b) 6(h) 12(h) 100(day)
Process water line i
isolation valve (1),(4).
[
[ Fig's.1.2-2,1.2-3,1.2-3a, 5.1-3]
171 100 66 66 Temperature ( g )
Feedwater isolation valve (1)
Pressure (Kg/cm g 0.035 0.035 0.035 0
l
[ Fig's.1.2-2,1.2-3,1,2-3a/
5.1-3]
Humidity (%)
Steam Steam 100 90 Max.
he(2) 1(h) 6(h) 12(h) 100(day)
Temperature (# ) )
C 171 100 66 66 RCICinjection valve (1), check Pressure (Kg/cm g 0.035 0.035 0.035 O~
l valve (inside MS tunnel), steam line isolation valve [ Fig's.
Humidity (%)
Steam Steam 100 90 Max..
1.2-2,1.2-3,1.2-3a/5.4-8).
. h c(2) 1(h) 6(h) 12(h).
100(day)
I 100(3) 66 66 RCIC(valve except isolation Temperature ( g )
Pressure (Kg/cm g 0.035 0.035 0
valve, assemblies, cable, -
turbine) [ Fig's. l.2-4/
Humidity (%)
Steam 100 90 Max.
5.4-8]
Time (2) 6(h) 12(b) 100(day)
.l 100 66 66 l
Temperature ( g )
RCIC turbine electric control 5
Pressure (Kg/cm g 0.035 0.035 0
system (3),(6)[ Fig's. L2-5/
5.4-8]
Humidity (%)
Steam 100 90 Max.
i Time (2) 6(h) 12(h) 100(day) 100 66 66 l
RHR (LPFI, cooling system at Temperature ( g )
Pressure (Kg/cm g 0.035 0.035 0
S/D, cottainment cooling. Ser-vice water system) valve, pump Humidity (%)
Steam 100 90 Max.
(motor, seal cooler) instrument Time (2) 6(h)_
12(h) 100(day) i control electric equipment (in-i cluding cable and sources of j
electricity)[ Fig's.1.2 4/
l 5.4-10]
l 3t3-16 i
Amendment 21 l
i
P lr{S' J
NRC Ouestion: The staff noted in the DSER that the integrated gamma accident dose is in primary containment for the ABWR is given as 6 x 107 rads, which is less than the typical value of about 2 x 108 rads quoted in the safety analysis reports of several operating reactors (e.g. Perry: 2.7 x 108 ; River Bend : 1.7 x 108 rads; Clinton: 2 x 10 rads' Nine Mile Point: 1.4 x 108 rads). It is not clear 8
why the ABWR integrated gamma accident dose is lower than the corresponding doses quoted for several operating reactors. GE's position which was provided in Section 5.3.2.1.5 of SSAR Amendment 15, did not adequately address this issue.
To resolve this issue GE must fully explain why the ABWR integrated gamma accident dose is lower than the corresponding doses quoted for several operating reactors. This is Open item 3.11.3-3.
oggg 3.H.3-3 Reply: The value of 6 x 107 rads originally reported in the ABWR SSAR is not the total integrated gamma dose for the primary containment for U.S. application but is the total integrated gamma dose as stipulated for the Kashiwaski 6/7 reactors being built and licensed under Japanese regulations. The difference between this value and the quoted existing U.S. reactors is one of philosophical approach between the two countries. To examine this difference we will compare the above ABWR calculation to that of the original TVA STRIDE design (BWR 6) i which is shown in the following table.
TVA Stride Primary Containment Integrated Gamma Dose Source Dose (rads carbon) 100 % Noble Gases + daughters 2.54 x 107 I
airt>ome 50% Halogens + daughters airbome 7.55 x 107 25% Halogens + caughters plateout Wall Plateout 7.04x 108 Ecuipment Plateout 1.20 x 108 i
Total of Plateout 1.90 x 107 Total Dose 1.20 x 108 Noble Gas Dose In the ABWR, a preliminary calculation for the noble gases has been done and is shown in the attached figures. In all three figures we see that the integrated drywell dose approaches 1.2 x 107 rads in each of the three major primary containment volumes. In Stride, a single volume was used i
to contain all the fission product release whereas in ABWR there are three separately distinct volumes which are separated by meters of concrete. In ABWR it is possible that for a short time, on the order of hours, all the noble gases would be contained in a single volume, this would certainly not be the case for a 100 day evaluation. This short period containment has not been considered in the attached figures but will be in the final. 2 s f s' evaluation. Therefore if a single volume (forcing all the release into a single volume for 100 days) were considered, the ABWR dose would increase an estimated factor of 2 to approximated 2.4 x 107 rads which is similar to Stride. Halogen Airt5orne Dose The airbome halogen dose is similar to the noble gas dose and for ABWR is roughly estimated at approximately 4 x 107 rads, again dividing the fission product release between three compartments. In a similar fashion if a single compartment were considered the dose would be approximately 8 x 107 rads which is similar to tho Stride value of 7.55 x 107 rads. Halogen Plateout Dose It is at this point that the K-6/7 analysis and the standard U.S. analysis differ. The standard U.S. analysis as is shown in the above table also considers an additional 25% halogen plated out onto the containment surfaces. The K-6/7 analysis does not. No estimate exists yet as to what this factor will be on ABWR since it needs to be determined if the release 6 would be divided betwet n the three volumes equally or by some mechanistic algorithm or whether all the release will be concentrated in a single volume. Nevertheless, it is not likely that ABWR will vary significantly from other analyses. Conclusion The above discussion has attempted to describe the original basis and differences between existing plant analyses and the value originally found in the ABWR SSAR. As is shown, when the ABWR evaluation is compete, it is not expected that the final ABWR will vary significantly from current plants.
Lower Drywell Integrated Dose from Noble Gas at various Leakages 1.0E+08 No Leakage g = B 1.0E+07 . 0.4%/dey f /darI / / I 1.0E+06 0 5 10 15 20 25 30 35 40 45 Time in days J 2s
1 e Upper Drywell Integrated Dose ,from Noble Gas at various Leakages 1.0E+08 No Leakege - mumE m ~ B 1.0E+07. o.4slaey x / 12%/ day f f 0 5 10 15 20 25 30 35 40 45 Time in days e 4-g .. ~
a. a_ a g .k x e Wetwell Integrated Dose fnpas' Noble Gas at various Leakages 1.0E+08 No Leekage -.J m B 1.0E+07 o w o.y f' i.woyg ../ [ 1.0E+06 0 5 10 15 20 25 30 35 40 45 Time in days c: h w N i .. _ -. ~ -}}