ML20044G773
| ML20044G773 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/01/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9306040240 | |
| Download: ML20044G773 (77) | |
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GENuclear Energy Genero!Decric Compasy 175 Curtner hvenue. San Jose. CA 95125 June 1,1993 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Markups for Chapters 2 and 3 of the ABWR SSAR
Dear Chet:
The subject markups have been discussed with T. Cheng of the NRC.
Please distribute to T. Cheng and G. Bagchi Sincerely,
!- L' I bLggf Jack Fox Advanced Reactor Programs cc: G. W. Ehlert ,
Norman Fletcher (DOE)
T. Lo (LLNL)
N182 0300G 30 t\ ,
9306040240 930601 l PDR ADOCK 05200001 l A PDR'
4 21A6100AD
- FEV. B SJulbrd Plant TABLE 2.0-1
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ENVELOPE OF ABWR STANDARD PLANT SITE DESIGN PARAMETERS ed:
Maximum Ground Water level:
Extreme Wind:
BasicWing/ km/hr g 171km/hr l 61.0 cm below grade
- 97 Tornado:I4}
Maximum Flood (or Tsunami) level:I I 30.5 cm below grade - Maximum tornado wind speed: 483 km/hr l - Maximum Rotational Speed: 386 km/hr
~
- 45.7 m Precipitation (for Roof Design):493 cm/hr(8) - Radius: 2
- Mrximum rainfall rate: 2 - Maximum pressure drop: 0.141kg/cm 4
- Maximum snow load: 0.024 kg/cm 0.0346 kg/cm'/sec
- Rate of pressure drop:
- Missile Spectra: Per SRP 3.5.1.4 Spectrum I Soll Properties: 2 cT) l Design Temperatures: - Minimum Static Bearing Capacity: 732 kg/q9)
- Ambient - Minimum Shear Wave Velocity: 305 m/sec 1% Exceedance Wlues None at plant site
- 1.Jquefaction Potential:
- Maximem: 37.8"C dry bulb /25 C wet bulb res g from7 l (coincident),26.6"C wet bulb (non-coincident) SSE
- Minimum:-233 C Seismology: frAzn/ pt%cn q 0% Exceedance Wlues (Historicallimit) i - Maximum 46.1"C dry bulb /26.7"C wet bulb
)
" (coinddent),27.2"C wet bulb (non-coincident) - SSE Peak Ground Acceleration: f030g )
- Minimum:-40 C - SSE Response Spectrx per Reg. Guide 1.60
- SSE Time History: Envelope SSE Response Spectra (1) 50-year recurrence interval; value to be utilized for design of non-safety-related structures only.
(2) 200-year recurrence interval; value to be utilized for design for safety-related structures only.
Probable maximum flood level (PMF), as defined in ANSI /ANS 2.8, ' Determining Design Ba (3)
Flooding at Power Reactor Sites.'
(4) 10,000\000Lyear tomado recurrence interval.
(5) Free-)ield, at plant gode elevation.
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(6) Deleted N (7) L a-. 3 l . SurWa I:.lfw aiS:ia m .a::ufbiina:a=L oa,p. Al foa0 Aew probable maximum precipitation (PMP) with ratio of 5 (8) Maximum value for I hour over 2.6 km Q' minutes to i hour PMP of 0.32 as found in National Weather Source Publicat Maximum short term rate: 15.7 cm/S min.
This is the minimum shear wave velocity at low strains after the soil property uncertainties (9) have been applied.
2.02 Amendment 26
zmimio ;
ABWR ws l Standard Plant i
/3 Design adequacy is established if rioo ^
2.3 COL LICENSE INFORMATION ; (1)
C/ l response spectra are bounded by Section 3G 4 spectra (or the actual spectra considere$in
)
23.1 Envelope of Standard Plant Design design if applicable) at key locations. Thz .ite l Parameters unique response spectra used for comp ' on need not be broadened since uncertain ses in 23.1.1 Non-Seismic Design Parameters the structural frequencies hav been l ccounted for in the smooth broade ed site Compliance with the envelope of ABWR Standard e velope spectra.
Plant site non-seismic design parameters of Table 2.0-1 shall be demonstrated for design bases events. If n t, examine whether the de ions are at hd m 3 (See Subsection 2.2.1)
(2) majo resonant frequencies of th component W under usideration. If not, , ign adequacy 23.1.2 Selsmic Design Parameters ts co a med. Otherwise, pc orm analysis
+
and/or esting to demon rate that the To conttrm the seismic design adequacy of the/ accepta ce criteria gi en in design l
\tandard plant, COL applicants shall demonstrate specificatio are met, that the eight (8) site-dependent conditions specifkd in Kection 3A.1 are satisfied. In meeting these elght If the soil pro of the ' e varyvery abruptly cond,itions, the compliance with the site envelope with depth (site de endent ondition 7), a site parameters shown in Table 2.0-1 for soil properties specific SSE SSI ~ is re uired. The evaluation and seismologyis also established. / ce iteria specified above procedures and accept
\ is any deviation of If there e eight are applicable.
site-dependent conditions, a site specific evaluation isIf the soil bearing acity at the site is not required. Tve type of evaluation will vdy dependmg standard plant design
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.U on the deviatihn. If the deviation is f6r condition adequate to accomm loads1 (site-dependent condi 'on 8), the foundation
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(peak ground acceleration),2 (ground response material shallbe re ed an eplaced with better spectra), or 6 (sicar wave velocity), a site specific material to achie the requir d bearing capacity.
SSE soil structSte interactiory analysis (SSI) is Alternatively, the L applicanthtay perform a site j icquired. The calcklated site unique responses are demonstrate t the site has an '
specific analysis compared to the sith-envelop / responses defined in adequate beari capacity against he site unique Section 3G.4 to confirlo the seismic design adequacy loads.
of the standard plantheepding to the following ,
The site- ependent conditions 3 qquefaction l procedures and acceptancepriteria. potential) nd 4 (fault displacement otential)
The Seismic Catego I tructures including the require si specificinvestigation.
lJ RPV and its internal components that are included ,
in the SSI analysis modeh A s' e specific evaluation is required if the
/ em est depths of Seismic Category I bubdings (1) Design adequ/cy is establi ed if maximum devi e from those from the standard plant de, sign in terms f force, moment, structural are bounded by a e Section 3G.4 (si -dependent condition 5). The evaluation
- or a pr dure and acceptance criteria are the same'as responses the actual seismic lhads considered ose defined above for the site specific SSE SSI in design ' app 5 cable) at key !=tiAae Aainis-
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. (2) If not, calculate resulting SSE stresses. Design 2.3.2 Standard Review Plan Site Charscteristics adequacy is confirmed if combined stresses due to SSE and other appropriate loads ariwithin Identification and description of all differences
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design code allowable limits. from SRP Section II Acceptance Criteria for site l c ,
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(') For Seismic Category I equipment and piping characteristics (as augmented by Table 2.1,1) shall be provided. Where such differences catst, the ,
i whose seismic input is in the form of floor response s
evaluation shall discuss how the alternate site spectra: 3 characteristicis acceptable. In addition, the COL
! 23-I J Amendment 27
'1 '
(D To confirm seismic design adequacy of the standard plant, COL' applicants referencing the-ABWR design shall demonstrate that O the site-specific conditions meet the following site envelope-parameters considered in the standardized design.
(1) SSE Ground Motion
'The site-specific SSE ground response spectra of 5% damping at !
plant grade in the free-field are' enveloped by the design :
ground spectra shown in Figs. 3.7-1 and 3.7-2 of the horizontal :
and vertical components, respectively, which are based on RG '!
1.60 (Revision 1 December 1973) anchored to 0.39 peak ground acceleration. When the site-specific control ground motion is- -i determined to locate at the rock outcrop or a hypothetical rock -
outcrop according to SRP 3.7.1 guidelines (e.g., shallow soil :
site), the site-specific soil free-surface motion through soil layer amplification shall be calculated and the resulting ground surface response spectra shall be bounded by the design !
ground spectra. !
4 (2) Bearing Capacity
- The site soil static bearing capacity at the foundation-level -f of the reactor and control buildings is 15 ksf minimum.
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1 23A610MD MN nm a Standard Plant _
COL applicants will provide the onsite applicant will provide / address the following: meteorologpcal measurements programs.
)
2.3.2.1 Site Location and Description 2.3.2.9 Short-Term Dispersion Estimates for Accidental Atmospheric Releases COL applicants will provide site-specific information to site location, includir.g political COL applicants will provide site-specific subdivisions, natural and ms::-made features, short-term dispersion estimates for NRC review to population, highways, railways, waterways, and otherensure that the envelope values (Tables 15.6-3, significant features of the area. 15.6-7.15.6-13,15.6-14 and 15.6-18) of relative concentrations are not exceeded.
2J.2.2 Exclusion Area Authority and Control 2J.2.10 Long-Term Diffusion Estimates COL applicants will provide site specific information related to activities that may be COL applicants will provide annual average permitted within the designated exclusion area. atmospheric dispersion values for routine releases for NRC review.
2.3.2.3 Population Distribution 2.3.2.11 Ilydrologie Descriptica COL applicants will provide population data for the site environs. COL applicants will provide a detailed descripdon of all major hydrologic features on or in the vicinity 2.3.2.4 Identification of Potential Hazards in Site of the site. They will also provide a specific Vicinity description of the site and all safety-related elevations, structures, exterior a-m equipment, COL applicants will provide information with p)
( respect to industrial, military, and transportation facilities and routes to establish the presence and and systems from the standpoint of hydrology mahmions.
magnitude of potential external hazards.
2J.2.12 Floods 2J.2.5 Evaluatice of Potential Accidents COL applicants will provide site specific information related to historical flooding and the COL applicants will identify potential accident potential flooding at the plant site, including flood situations in the vicinity of the plant and the bases for history, flood design considerations, and effects of which these potential accidents were or were not localintense pr~4pimh accommodated in the design.
I 2.3.3.13 Probable Maximum Flood om Streams and 23.2.6 External Impnet Hazards Rivers COL applicants will provide a review and COL applicants will provide site-spccific evaluation of the effects on the protection criteria of information related to determining design basis some externalimpact hazards, such as general flooding at power reactor sites and the extent of aviation or nearby @h -
flood protection required for those safety related systems, hw s and components.
2J.2.7 Local Meteorology 2.3.2.14 Ice Effects COL applicants will provide local meteorology for NRC review COL applicants will demonstrate that safety-related facilities and water supply are not affected byice floodmg or blockage.
('; 2.3.2.8 Onsite Meteorological Measurements 2J.2.15 CoolingWater chienets end Reservoirs !
Program COL applicants will provide the basis for the 2.3-2 Amendment 26
.Nb Cenp.
'S ussimso ABWR nev a Standard Plant 6 cation of the ground motion response spectra [
)
hydraulic design of channels and reservoirs used to
\J transport and impound plant cooling and for 23.2.23 Surface Faulting l protection of safety.related structures.
COL applicants will develop site-specific 23.2.16 Channel Division geological data to ensure that no potential exists for surface faulting at the site.
COL applicants will provide site-specific information related to channel diversion. 2.3.2.24 Stability of Subsurface Material and l Foundation 23.2.17 Flooding Protection Requirements
+- Jwb @
r COL applicants will develop and submit to tae)
COL applicans will provide site-specific NRC site-specific geotechnical data to demonstrate M iaformation related to flooding proteetion that they are comparable to the design assumptions requirements. concerning the soil-deposit depths, the soil profile and properties, and the ground water level.
23.2.18 Cooling Water Supply Particular attention should be paid to the assumptions for the depth of embedment in the case COL applicants will identify natural events that of rock and the three cases of soil. deposit deghs for may reduce or limit the available cooling water which fiaed values of depths are assumed. COL supply and ensure that an adequate water supply willapplicants will demonstrate that the envelope of' exist to operate or shutdown the plant as required. structural response with fixed soil depth will cover completely the cases for which the soil deposit 2J.2.19 Accidental Release of Uquid Effluents in depths and properties are different from those Ground and Surface Waters -
Jessumed in theKMR.
~
COL applicants will provide information on the l
~ "( #) ability of the surface water environment to disperse, 2J.2.25 Site and Facilities dilute, or concentrate archral rele== Effects of COL appHesen will provide a detailed description these releases on existing and known future use of of the site conditions and_ geologic features, ad surface water / resources shall also be pronded.
(Semonstrate the site charactenstics are enveloped by f
the 03r peak lukuutal ground acceleration for the 2J.2.20 Technical Specifications and Emergency QSE./The description williractude site topographical Operation Requirement features and the location of various Seismic Category I structures and appurtenances (pipelines, ch* Is COL applicants will establish the technical etc.) with respect to the source of normal and specifications and emergency procedures required toemergency coolingwater, implement flood protection for safety.related facilities and prodde assurance of an adequate water 2J.2.26 FleidInvatigations supply to shutdown and cool the reactor.
The type, quantity, extent, and purpose of all field 2.3.2.21 Basic ^
Cr'@l and Selamle Information exploration will be provided by COL applicants.
Logs of all borings and test pits should be provided.
COL applicants will provide site specific Results of geophpical surveys should be presented in informatioa related to regioaal and site tables and profiles. Records of field plate load tests, physiography, geomorphology, stratigraphy, lithologyfield permeability tests, and other special field tests yh and tectonics. (e.g., bore. hold extensometer or pressuremeter tests) should be ghen.
23.2.22 Vibratory Ground Motloa 2.3.2.27 Laboratorylavestigations COL applicants will develop site. specific j '] geological, scismological, and perWest data and 'Be number and type of laboratory tests and the
'J -
ubmn these cata to GENRC for renew. A nesi5 A location of samples should be provid ata should be comparable to the design basis applicant in tabular form. The results of laboratory ssumptions regardingjhe SSE. including 2.33 Amendment 26
- m ._m_,.._ _ , . -- a ai
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h,' compare.thesite-specific-SSEgroundresponse' design ground. spectra according to Subsection 2.3.1.2.spectra'to-the 1
O P COL applicants will. provide'infornation concerning'the ~
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g.propertiesandstabilityofsite-specificsoilsandrocks'under
' both static and dynamic conditions including the vibratory j
( ground motions associated with the site-specific SSE. ,
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c.g 2wnomo ABWR ne.. o Standard Plant tests on disturbed and undisturbed soil and rock cohesion and angle of internal friction, and dynamic samples obtained from field investigations should soil propertiesj rd I; 4= _c . cage;2 xdy.!c:e -
also be provided. . d ccfe g .
2.3.2.31 Liquefaction Poten SSs prcu d n -
M g l 23.2.28 Subsurface Conditions COL applicants mus demonstrate that no COL applicants willinvestigate and define the liquefaction potential exists for soils under and subsurface conditions and provide the engineering around all Seismic Category I structures, including classifications and descriptions of soil and rock Category I buried pipelines and electrical supporting the foundations. The information should COL applicants willjustify the selection of the soil ducts. l include the history of soil deposition and erosion, properties used in the liquefaction potential past and present goundwater. levels, glacial or other evaluation (e.g., laberatory tests, field tests, and preloading influences, rock weathering, and any rock published data), the magnitude and duration of the or soil characteristics that may present a hazard to earthquake and the number of cycles of earthquakes.
plant safety. Profiles through the Seismic Category I l structures will be provided that show generalized 2.3.2.32 Response of Soll and Rock to Dynamic Ioading l
subsurface features beneath these structmes.
4-.- - Jn. serb &
l 2.3.2.29 Excavation and Backfilling for Foundation [ COL applicants must establish and documen Construction tite-specific geoterbical properties to demonstrate their comparability with the conditions used for the COL applicants will provide site specific Winic Ann enveloce demibd in Annendir thickness and properties of soil (if any) between the base of the foundation and the underlying rock. The 23.233 Maxim m Soll Bearing Pressure configuration, along with detailed longitudinal d--lnser6 $
sections and cross sections of other safety.related COL applicants win proviac the site.specirk structures of the plant, including the ultimate heat aximum soil pressure along with supporting sink and Seismic Category I buried pipes and elec- calculations and will comnare thens with a!!ownbl.
trical ducts, should be provided. COL applicants l wiB provide data concerning the extent (horizontally
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2.3.234 Earth Pnssures sde M g c, l and vertically) of all Seismic Category I excavations, fills, and slopes. The locations, elevations, and i grades for excavated slopes should be described and COL applicants will provide a discussion anB shown on plot plans and typical cross-sections. COL evaluation of static and dynamic lateral earth applicants'submittals should disccss, as appropriate, pressures and hydrostatic groundwater pressures excavating and dewatering methods, excavation acting on plant acilitiesp ic aca m.ccrf :c s.
de- s t*4 - m mecHE: 6:!p t depths below grade, field inspection and testing of A-excavations, protection of foundation excavations .he fmheABWn - ' tc ddrese-all4acilities from deterioration during construction, and the -c? " E th W N Sa%2 7 tel d e2.
foundation dental fillwork. The sources, quantities, and static and dynamic engineering properties of 2.3.2.35 Soll Properties for Seismic Analysis of borrow materials willbe described. The compaction Buried Pipes requirements; results of test fills, and fill properties, -fv such as moisture content, density, permeability, COL applicants ~ provide and justify the soil compressibility, and gradation should be provided. properties used r the seismic analysis of seismic Category I buried pipes and electrical conduits.
Effech of l 23.230 3Groundwater yM 23.236 Static and rde-nec=$c.
e Stability of Facilities COL applicants will analyze he goundwater Perform Sino Mqmn/ddim #f condition for the specific site and dems :,:=:: hs ' COL applicants willkaW< all safety-related p= bill:y rkh4h: /.E?"; W wg.mL facilities it i; :.:= c4-- -- , te Mernte d#
Ti- i.uac: ea d.;;!d indudithe effect of the"- ? + at.2 2 i -- - ' ?.E?!". E:!; Sn c;d to ^-
.ae+ent groundwaterlevetYn such site geotechnical adacas :" p!::: P = _ t% i: ?.BV'". =p-R_.
properties as total and effective unit weights, W , udyr- Jinclude foundation rebound,
) (
2M Amendment 26
_ . , , . . - - . . .~
4)
.j COL applicants will determine dynamic soil properties of1the -)
& site in terms of shear: modulus and material damping as function W' I of shear strain. These strain-dependent properties will-be used
' - in the determination of'the site-specific SSE ground motion.
j J
COL applicants will demonstrac.e that the site hass a minimum i static bearing capacity of 15 ket at the foundation-level of-
@4 the reactor and control buildings. For other safety-related plant facilities. COL applicants will demonstrate that the-foundation material has adequate bearing capacity to withstand .i L the site-specific loads.
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M - 23A6100AD au s Standard Plant
. settlement, differential settlement, and bearing capacity,that == ne --,-,a rnr assian In=ds of grm. and clant raeintien. Assumptions made in ,
l stability analyses wiE be confirmed by as-built data. 1 l 2J.237 Subsurface Instammentation Instrumentation, if any, proposed for the surveillance of the performance of the foundations for safety-related structures will be described by COL applicants. The type, location, and purpose of each instrument and significant details of installation methods will be provided. For example, the location '
and the installation procedures for permapent :
benchmarks and markers required for monitoring '
the settlement of Category I structures should be ,
described. In the case of safety-related water-control !
structures (such as damt, slopes, canals), the details ,
- of installing instrumentation such as piezometers, slope indicators, and settlement plates should be described. A schedule for i=<t="~ag and reading all
- instruments and for interpreting the data will be i presented. Limiting values for co'stinued safety should be identified.
l 23.238 StabilityofSlopes
- i COL applicants will provide information about :
the static and dynamic stability of all soil and rock slopes, the failure of which could adversely affect the safety of the plant. The staff will evaluate the stability of all slopes at the site, using the .;
state-of the-art procedures available at the time of t
application.
h l 23.239 Em%ts and Dams COL applicants should provide information about the static and dynamic stability of all embankments and dams that impound water required for sadifsperation (and shutdown) of the ABWR for reviewby the NRC if embankments and dams are used. W .
2.3.3 CRAC 2 Computer Code, Calculations Compliance with acceptance criteria, data input and analysis of Subsection 2.2.2 for the ,
determination of ABWR site acceptability for severe accidents shallbe demonstrated. :
r Sh Amendment 26 ;
AMM 23461004u nrw a Standard Plant
- r 3.3 WIND ANDTORNADO LOADINGS Reference 1. Reference 2 is used to obtain the effective wind pressures for cases which Refer-ABWR Standard Plant structures which are ence 1 does not cover. Since the Seismic Cat-Seismic Category I are designed for tornado and egory I structures are not slender or flexible, extreme wind phenomena. vortex-shedding analysis is not required and the above wind loading is applied as a static load.
3.3.1 Wind IAadings s- Arp>ncl .[g,g 4 3.3.2 Tornado IAadings %cb cong g g ' 4t 33.1.1 Design Wind Velocity mdw"Ste be 1d ft '
33.2.1 Applicable Design Parameters found in a dg I\0 3Ha 3w g Seismic Category I structures are, designed to g withstand a design wind velocity of de mph at an The design basis tornado is described by the [3' elevation of 33 feet above grade with a recur- following parameters:
ryI I rence interval of 40Q years. See Subsection 3.3.3.lgfor interface re,quirement. (1) A maximum tornado wind speed of 300 mph at a an d 3.3 5.3 So radius of 150 feet from the center of the 33.1.2 Determination of Applied Forces tornado; The design wind velocity is converted to (2) A maximum translational velocity of 60 mph; velocity pressure in accordance with Reference 1 using the formula: (3) A maximum tangential velocity of 240 mph, based on the translational velocity of 60 ;
q = 0.00256 zK (IV)' mph; z
where K =the velocity pressure exposure (4) A maximum atmospheric pressure drop of 2.00 coefficient which depends upon the psi with a rate of the pressure change of type of exposure and height (z) 1.2 psi per second; and above ground per Table 6 of i
Reference 1. (5) The spectrum of tornado-generated missiles and their pertinent characteristics as given I = the importance in Subsection 3.5.1.4 Meur.h@n the type ofg, factor xpere; which appropr: ate depends values of I are listed in Table See Subsection 3.3.3.2 for COL license 3.3-1, information.
tsa V = design wind velocity of +30 mph, and 33.2.2 Determination of Forces on Structures q#
= velocity pressure in psf The procedures of transforming the tornado
~. - -_
- s loading into effective loads and the distribu-The velocity pressure (q7 ) distribution with' tion across the structures are in accordance
[ height for exposure type \;G-ansly of Reference I with Reference 4. The procedure for transform-
\ are given in Table 3.3-2.
- - m-
% ing the tornado-generated m^ssile impact into an effective or equivalent static load on struc-The design wind pressures and forces for tures is given in Subsection 3.5.3.1. The load-buildings, components and cladding, and other ing combinations of the individual tornado load-structures at various heights above the ground ing components and the load factors are in accor-are obtained, in accordance with Table 4 of dance with Reference 4.
Reference 1 by multiplying the velocity pressure by the appropriate pressure coefficients and gust The reactor building and control building are factors. Gust factors are in accordance with not vented structures. The exposed exterior Table 8 of Reference 1. Appropriate pressure roofs and walls of these structures are designed coefficients are in accordance with Figures 2, for the 2.00 psi pressure drop. Tornado dampers 3a,3b,4, and Tables 9 and 11 through 16 of Amendment 23 3.3-1
f ABTM 23461oore i Standard Plant REV.H are provided on all air intake and exhaust 3. Deleted openings. These dampers are designed to withstand a negative 4,44 psi pressure.
2.o ,
33.23 EITect of Failure of Structures or Components Not Designed for Tornado leads 4. Bechtel Topical Report BC-TOP-3-A, Revision 3, Tornado and Extreme Wind Design Criteria All safety-related system and components are for Nuclear Power Plants.
protected within tornado-resistant structures.
for COL license l See Subsection 3.3.3.
information requirements.
3.3.3 COL License Information 333.1 Site-Specific Design Basis Wind The site-specific design basis wind shall not exceed the design basis wind given in Table 2.0-1 (See Subsection 2.2.1).
3.33.2 Site-Specific Design Basis Tornado The site-specific design basis tornado shall not exceed the design basis tornado given in Table 2.0-1 (See Subsection 2.2.1). l 33.3.3 ERM d &%du f 3.3.3 Effect of Remainder of Plant Strue. P 6+ s+cubres, sgshyns and tures, Systems, and Components not Designed for Tornado Loads <* '"P e n c o-fs n94- desio3ned 4ac coind
}Ohd5 All remainder of plant structures, systems, All f ernagnder of p} ant structures and components not designed for tornado loads 2 shall be analyzed for the site-specific loadings gshrns, and cornponeM5 vso4 to ensure that their mode of failure will not cle,93ned Eor M d loaE5 SboII b6 '
effect the ability of the Scismic Category I ABWR Standard Plant structures, systems, and compo- ""
fb 6 g#D 4 g*gg i
nents to perform their intended safety functions. or shnN (See Subsection 33.2.3) trnPrka.n c e. Sac ft>r c.hecl<ed 4bok Nett mofE *p 3.3.4 References 4 o.d ure W Lg M j gc4 4 a}fy'rf) j i'
- ANst/Asc6 7-Vg
- 1. %NSI-StandardA58,1, Minimum Design Loads yO_yg\akeh OroCNT**5 s YW ,
- y for Buildings and Other Structures, o g- i"f,m fQc '
C u nrmit t ee-Ar-58vir-A m e r i c a n--N e f ie n a % co rnfoOf
-St Noveandards-Institute.%
m 6er .17, t 99o , est g .Qgeks'ot15 _
- 2. ASCE Paper No. 3269, Wind Forces on Structures, Transactions of the American Society of Civil Engineers, Vol.126, Part II.
I Amendment 27 3.3-2
k 23A6100AE Standard Plant REV A Table 3.31 Importance Factor (I) for Wind Imads Non s a6% - Re tu4, d 56h Rt \chth
-EXPOSLTREC- -EAMiRED--
1.00 -le+
1.11 N O E S:
- 1. These values of (I) are based on Table 5 of Reference 1 but-are-modified to4,n-c, the iseyer-return-period-of-the-de'siga wind velseisNenus-the 50 poi- i cturn-period 4a-si: ef Referesee-4w -
- 2. Exposure categories are as defined in t
.Se_ction 6.5.3 of Reference 1.
I l
Amendment 1 3.3-3
__ . _ _ -- . ~ . _
ABM - us6 oors .i Standard Plant REVIA j
Table 3.3 2 VELOCITY PRESSURE DISTRIBUTION AND GUST FACTORS ATVARIOUS HEIGHTS !
EXPOSURE TYPE C . l Height qz Gh Windward Side Wall Roof Leeward l Zone (ft/sec) Wall Pressure Suction Suction _ Wall Suction :
z (ft) 0.8G qz 0.7Ggqh 0.5G qh !
4 0.7Ggqh 4 (Ib/ft-) (Ib/ft') (Ib/ft ) (ib/ft') l i
0-15 34.6 \ 132 37 58 58 42 :!
20 37.6 1.29 39 57 57 41 I 25 40.2 1.27 41 56 . 56 40 l 30 42.4 1 26 43 56 56 40- ;
40 45.9 13.3 45 54 54 39 j 50 48.9 47 54 54 38 l 1.21\ I 60 51.5 1.20 \ 49 53 53 38 70 53.6 1.19 51 53 53 38 80 55.8 1.18 52 52 37 :'
90 58.0 1.17 '52 52 37' .
100 59.7 1.16 55 51 51 37 l 120 62.7 1.15 58 .
51 51 36 140 65.8 1.14 60 50 50 36 ,
160 68.4 1.13 62 50 50 -36 l I
-t .
EXPOSURE ED ,
i Height qz Ga Windward 'de Wall Roof Leeward !
Zone (ft/sec) Wall Pressure S 'on Suction Wall Suction . .i z (ft) 0.7 0.7G 0.8Ggqz ' (Ib/fgqh 0.5Ggqh (Ib/ft ) (lb/ft ). t) (Ib/ft )-
0-15 56.2 1.15 52 58 58 42 20 59.4 1.14 54 58 58 41 ,
25 61.8 1.13 56 57 57 41 !
30 64.1 1.12 57 57 57 41 40 68 3 1.11 61 56 40 -l 50 71.1 1.10 63 56 40 :
60 73.9 1.09 64 55 55 39 70 763 1.08 66 55 55 39 i
80 78.1 1.08 67 55 55 39 l 90 80.0 1.07 68 54 54 39 100 81.9 1.07 70 54 '54 39 d 120 - 84.7 1.06 72 54 54. 38 ,
140 87.5 1.05 74 53 53 +
160 89.8 1.05 75 53 53 g 4 .
qh " 9z @ z = 123.69 ft (roof height above grade) for the reactor building {
t
.@A j!
t ~
Amendment 1 3.34 p- ,_r
23A61[X)All Standard Plant arx n i SECTION 3.7 !
I CONTENTS (Continued)
Section Title Ea..gg i 3.73.17 Methods for Seismic Analysis of Above-Ground Tanks 3.7-24 3.7.4 Seismic Instrumentation 3.7-24.1 Leeda.;cn and kst tri ~ lion efSusa Indrusslakb.1 3.7.4.1 Cc r"h^a "M ".C Regtrtato.y Gmdc112-L 3.7-24.1
%e Zadramdsknt. CyrMf and Chraclerak 3.7.4.2 beentkwen4Descr4ption of 1strumentation % 3.7-24.1 3.7.4.3 Control Room Operator Notification 3.7-24.1 3.7.4.4 Comparison of Measured and Predicted Responses 3.7-24.2 3.7.4.5 In-service Surveillance 3.7-24.2 3.7.5 COL License Information 3.7-26 3.7.5.1 Seismic Design Parameters 3.7-26 3.7.5.2 Pre-Earthquake Planning and Post-Earthquake Actions 3.7-26 3.7.6 References 3.7-26 4
3.7-vi Amendment l
Nf c ? "; h = Q, an-ihte&dmmthate McGr- GL MkN hub */cc8*pyen"c"/JTiechpn Vrices( g3A6100AE -
n, n y Standard Plant w7---~~ \
_. _.h (7) Either the tank top head is located at an .7.4.2 Location and Description of \
elevation higher than the slosh height above I trymentatji n ,4 y g (
the top of the fluid or else is designed for JnM (g) (
pressures resulting from fluid sloshing against iarid ; ira; %;j c c;ck.g;. J-FFAprs'- \
this head. im d E 'h: fis MX Mfroduces a record ofthe time-v rying acceleration at the sensor location / The i (8) At the point of attachment, the tank shellis triaxia acceleration sensor unit contain threcf - '
accelero eters muunted in an orthigonal ray (tb q- '
designed to withstand the seismic forces imposed by the attached piping. An horizonta and one vertical) in alignmen with the ,,po6 appropriate analysis is performed to verify y plant majo axes assumed in seismic an ysis. The-this design. f'
'd ensor has a dynamic ra ge of @Il zero tg"J.c., .0 g to 1.0g) and a frequency (9) The tank foundation is designed to accom- range c ~ . ". The THA system /s triggered try ,
modate the seismic forces imposed on it. the accelerome r signals. The tril[ser is actuated These forces include the hydrodynamic fluid whenever a thre old acceleration'" =: m:~ 9 .uce 5 pressures imposed on the base of the tank as r any of the hree axes. The f(ei g we11 as the tank shMI longitudinal 0.04 ; msb.E g^
initial setpoint ^ 4 M Ym y be changedg compressive and tensile forces resulting from significant plant ope ting data ave been obtained M.g which indicate that a ferent tpoint would provide (6 /xt better THA system ope tion g g (10) In addition to the above, a consideration is The instrumentatio nstalled is capable of gd given to prevent buckling of tank walls and roof, failure of connecting piping, and sliding on-line digital recording all three components of l
the ground motion he digitized rate of the h of the tank.
recorder is at least ' - sam es per secondjmd'the g~,
3 bandwidth is at least rom 0. Hz to 50 HzY. he 4 7.4 Seismic Instrumentation pre event memory the instr ent is sufficient to f lyu C.
^
record the onset of e earthquak The instrument \ r*
\3.7.4.1 Comparison with NRC Regulatory Guide is also capable of onverting the ecorded (digital)k ec 11-l 12 signal into the tandardized C V and the 5% \
The seismic instrumentation prpgram damped respon spectrum.
desen ed in the following subsections ets the intent s ' Regulatory Guide 1.12,in t at seismic The ins umentation system i capable of tion is provided so that he control routinely ca rating the response spectr check of instrume reom oper tor can be immediat'ely informed 0.2g. Al o, the CAV of 0.16g-sec ould be {
through the es nt indicators whe fesponse spectra . calibrate with a copy of the October,198 Whittier, level and the c ulative abso te velocity (CAV) Califor a earthquake or an equivalent ca 'bration in the free field ex eds the s tdown level and can record rovided for this purpose by the manu acture take the necessary ahions. / of th instrumentation. In the event that an etual
\/ [ eart quake has been recorded at the plant sit the ab e calibration will be performed to demonst te Seismic instru ' tation at locations oit e tructures as required G 1.12 is not provided} t t the system was functioning properly at the ti since the free fieldInstr tation is sufficient to the earthquake.
provide informati/n required gdetermine whethe 3.7.43 Control Room Operator Notification yhnt shutdown n neede' Seism instrumentation a locations on Activation of the seismic trigger causes an equipme , piping, and supports as rehed by RG audible and visual annunciation in the control room to alert the plant operator thatOfel carthquake has y 1.12 is ot provided since experience as shown that ta obtained at these locations are bscured;: occurred. T ratory motion associated with norm plant) Om by o cration.
A 4-- yf 3.7-211 Amendment
Insert a i
3.7.4.1 Location and Description of Seismic Instrumentation State-of-the-art solid-state digital instrumentation that will enable the prompt processing of the data at the plant site should be used. A triaxial time-history accelerometer should be provided at each of the following locations:
One at the finished grade in the free-field Three in the reactor building: one located on the foundation mat at elevation -8.2 m, one at floor elevation 12.3 m and one at the operating -
floor at elevation 31.7 m.
Two in the control building; one on the foundation mat at elevation -8.2 and one at elevation 7.9 m.
3.7.4.2 Seismic Instrumentation Operability and Characteristics The seismic instrumentatica should operate during all modes of plant operation, including periods of plant shutdown. The maintenance and repair procedures should provide for keeping the maximum number of instruments in service during plant operation and shutdown.
The design should include provisions for in-service testing. The instruments should be capable of periodic channel checks during normal plant operation and the capability for in-place functional testing. The instrumentation on the foundation and at elevations within the same building or structure should be interconnected for common starting and common timing, and the instrumentation should contain provisions for an external remote alarm to indicate actuation. .
The pre-event memory of the instrumentation should be sufficient to record'the onset of the earthquake. It should operate continuously during the period in which the earthquake exceeds the seismic trigger threshold and for a minimum of 5 seconds beyond the last trigger level signal. The instrument should be capable of a minimum of 25 minutes of continuous recording. The acceleration sensors should have a dynamic range of 1000:1 zero to peak (i.e. 0.001g to 1.0g) and the frequency range should be 0.20 Hz to 50 Hz.
The seismic instrumentation system is triggered by the accelerometer signals.
The actuating level should be adjustable for a minimum of 0.005g to 0.02g The trigger is actuated whenever the acceleration exceeds 0.01g. The initial setpoint may be change (but shall not exceed 0.029) once sufficient plant operating data have been obtained which indicate that a different setpoint would provide better system operation. 1 The instrumentation should be capable af on-line digital recording all components accelerometer signals. The digitized rate of the recorder should be at least 200 samples per second, the frequency band width should be at least from 0.20 Hz to 50 Hz, and the dynamic range should be 1000:1. The '
Insert a (continued) instrumentation should be capable of using the recorded signal to calculate the standardized cumulative absolute velocity (CAV) and the 5 % of critically e damped response spectrum.
The instruments should be capable of having routine channel checks, functional tests, and calibrations. The CAV shutdown threshold of 0.16g-seconds should be calibrated with the October, 1987 Whittier California earthquake record or anGFequivalent calibration record provided for this purpose by the manufacture of the instrumentation. In the event that an earthquake is recorded at the plant site, all calibrations including that of the CAV will be performed to demonstrate that the system was functioning properly at the time'of the earthquake.
G
[
l l
1 ABWR useimie '
Standard Plant nry A SECTION3.8 CONTENTS Section Title Page 3.8.1 Concrete Containment 3.8-1 3.8.1.1 Description of the Containment 3.8-1 3.8.1.1.1 Concrete Containment 3.8-1 33.1.1.2 Containment Liner Plate 3.8-2 3.8.1.2 Applicable Codes, Standards, and Specifications 3.8-2 3.8.L2.1 Regulations 3.8-2 3.8.1.2.2 Construction Codes of Practice 3.8-2 3.8.1.23 General Design Criteria, Regulatory Guides, 3.8-2 and Industry Standards 3.8.1.2.4 Containment Boundary 3.8-2 3.8.13 Loads and Load Combinations 3.8-3 3.8.13.1 Normal Loads 3.8-3 3.8.13.2 Preoperational Testing Loads 3.8-4 3.8.13 3 Severe Emironmental Loads 3.8-4 3.8.13.4 Extreme Emironmental Loads 3.8-4 3.8.13.3 Abnormal Plant Loads 3.8-4 3.8.13.6 Load Combinations for the Containment 3.8-5 Structure and Liner Plate 3.8.1.4 Design and Analysis Procedures 3.8-5 3.8.L4.1 Containment Cylindrical Wall, Top Slab 3.8-5 and Foundation Mat 3.8.1.4.1.1 Analytical Methods 3.8-5 33.L4.1.1.1 Nonaxisymmetrical Loads 3.8-5 3.8-il Amendment i
..ABM 2346ioore -
Standard Plant -REV A j
SECTION 3.8 CONTENTS (Continued)
Section Title P_agt 3.8.1.4.1.1.2 Axisymmetricalleads 33-6 l 3.8.1.4.1.1.3 Major Penetrations 33-7 3.8.1.4.1.1.4 Variation of Physical Material Properties 3.8 3.8.1.4.1.2- Design Methods 33-7 3.8.1.4.1.3 Concrete Cracking Considerations 3.8-7 3.8.1.4.1.4 Corrosion Prevention 3.8-8 ,
3.8.1.4.2 Ultimate Capacity of the Containment 3.8-9 >
3.8.1.5 Structural Acceptance Criteria 3&9 -]
i 3.8.1.6 Material Quality Control, and Special ;
Construction Techniques 33 9 !
3.8.1.6.1 Concrete 33-9 !
a 3.8.1.6.2 Reinforcing Steel 3.8-10 3.8.1.6.3 Splices of Reinforcing Steel 3.8-10
)
3.8.1.6.4 ,1iner Plate and Appurtenances 3 & 10 >
3.8.1.6.5 Ouality control 3 & 11 ,
3.8.1.6.6 Welding Methods and Acceptance Criteria for Containment Vessel Liner and Appurtances 3.8-11 1
3.8.1.7 Testing and Inservice Inspection Requirements 3.8-11 3.8.L7.1 StructuralIntegrity Pressure Test 3.8 ;
, 3.8.1.7.2 Preoperational and Inservice Integrated Leak Rate Test 3 & 11 :
3.8.2 Steel Comnonents of the Reinforced ,
Concrete Containment 3 & 11 !
3.8.2.1 Description of the Contamment 3 & 11 ,
33.2.1.1 Description of Penetrations 3.8-11 '!
3.8-iii :,
i Amendmen 2?'
zu++' n - _, _ ,
1 ABWR umwan Standard Plant REV A SECTION 3.8 CONTENTS (Continued)
Section Tille Paec 3.8.2.1.1.1 Personnel Air lects 3.8-11 3.8.2.1.1.2 Equipment Hatch 3.8-12 3.8.2.1.13 Other Penetrations 3.8-12 3.8.2.1.1.4 DrywellHead 3.8-12 3.8.2.13 Boundaries 3.8-13 3.8.2.2 Applicable Codes, Standards, and Specifications 3.8-13 3.8.2.2.1 Codes and Standards 3.8-13 3.8.2.2.2 Code Classifications 3.8-13 3.8.2.23 Code Compliance 3.8-13 3.8.23 Loads and Load Combinations 3.8-13 3.8.2.4 Design and Analysis Procedures 3.8-13 3.8.2.4.1 Description 3.8-13 3A2.4.1.1 Personnel Air Locks 3.8-13 3.8.2.4.1.2 Equipment Hatches 3.8-13 3.8.2.4.13 Penetrations 3.8-14 3.8.2.4.1.4 Drywell Heads 3.8-14 3.8.2.5 Structural Acceptance Criteria 3.8-14 3.8.2.6 Materials, Quality Control, and Special 3.8-14 Construction Techniques 3.8.2.7 Testing and Inservice Inspection Requirements 3.8 15 l
3.8-iv Amendmen' m
23A6100AE Standard Plant gev 4 SECTION3.8 CONTENTS (Continued)
Section lille Page 3.8.2.7.1 Welding Methods and Acceptance Criteria 3&l5 3.83 Concrete and Steel Internal Structures of the Concrete Containment 3.8-16 3.83.1 Description of the Internal Structures 3.8-16 3.83.1.1 Diaphragm Floor 3.8-16 3.83.1.2 Reactor Pedestal 3&l6 3.83.13 Reactor Shield Wall 3.8-17 333.1.4 Drywell Equipment and Pipe Support Structure 3.8-17 3.83.1.5 Other Internal Structures 3.8-17 3.83.1.5.1 Miscellaneous Platforms 3.8-17 3.83.1.5.2 Lower Drywell Equipment Tunnel 3&l7 3.83.1.53 lower Drywell Personnel Tunnel 3.8-17 3.83.2 Applicable Codes, Standards, and Specifications 3.8-17 3.833 Loads and Load Combinations 3.8-18 3&v Amendment
i ABWR mums :
Standard Plant nev. n .
i SECTION 3.8 -
CONTENTS (Continued) !
Section Title East ]
3333.1 Load Definitions - 3.8-18 ,
3.833.2 Load Combinations 3.8-18 .-
3.83.4 Design and Analysis Procedures 3.8-18 l 3.83.4.1 Diaphragm Floor 3.8-18 3.83.4.2 Reactor Pedestal ' 3.8-18 3.83.43 Reactor Shield Wall 3&l9 : !
3.83.4.4 Drywell Equipment and Pipe Support Structure 3.8-19 ' l A
3.83.4.5 Other Internal Structures ' 3E-19 3.83.5 Structural Acceptance Criteria 3.8-19 7
3.83.5.1- Drywell Equipment and Pipe Support Structures 3.8-19 ,
e 3.83.5.2 OtherInternalStructures . 3.8-19 .l t
3.83.6 Materials, Quality Control, and Special-Construction Techniques 3 & 19 3.83.6.1 Diaphragm Floor : 3.8-19 i
3.83.6.2 Reactor Pedestal . 3.8-19 ;
333.63 Reactor Shield Wall . 3&l9.1 - .;
3.83.6.4 Drywell Equipment and Pipe Support Structure 3 & l9.1=
Other InternalStructures 3.8-20 3.83.6.5 l 3 3-20 3.83.7 Testing and Inserdce Inspection Requirements .
'3.83.8 Welding Methods and Acceptance Criteria for l Structural and Building Steel : 3 & 20 j 3.8.4 Other Seismic Catenorv I Structures 3.8-20. l I
3.8-vi ,
t 3
Amendment ,
9 .. -..-- .
ABWR -mu arv. n Standard Plant SECTION 3.8 CONTENTS (Continued)
Section Title Page 3.8.4.1 Description of the Structures 3.8-20 3.8.4.1.1 Reactor Building Structure 3.8-20 3.8.4.1.2 Control Building 3.8-21 3.8.4.1.4 Seismic Category I Cable Trays, Cable Tray Supports, Conduit and Conduit Supports 3.8.4.1.5 Seismic Category I HVAC Ducts and Supports 3.8-21.1 3.8.4.2 Applicable Codes, Standards, and Specifications 3.8-21.1 3.8.4.2.1 Reactor Building 3.8-21.1 3.8.4.2.2 Control Building 3.8-22 314.23 Radwast e Building Substructure 3.8-23 3.8.4.2.4 Seismic Category I Cable Trays, Cable Tray Supports, Condra, and Conduit Supports 3.8-23 3.8.4.2.5 Seismic Category I HVAC Ducts and Supports 3 3-23 3.84.2.6 Welding and Weld Acceptance Criteria 3.8-23 3.8.4.2.6.1 Welding of Electrical Cable TYray and Conduit Supports 3.8-23 3.8.4.~2.6.2 Welding of Heating Ventilation and Air Conditioning Supports 3.8-23 3.8.4.2.63 Welding of Refuel Cavity and Spent Fuel Pool Liner 3.8-23 3.8.43 Loads and Load Combinations 3.8-23 3.8.43.1 Reactor Building 3.8 23 3.8.43.L1 Loads and Notations 3.8-23 3.8.43.1.2 load Combinations for Concrete Members 3.8-24 3.8.43.13 load Combinations for Steel Members 3.8 25 3.8.43.2 Control Building and Radwaste Building Substructures 3.8-25 3.8-vii Amendmen
ABWR mame anv. n Standard Plant SECTION 3.8 CONTENTS (Continued)
Sectioil Title Page 3.8.433 Seismic Category I Cable Trays, Cable Tray Supports, and Conduit Supports 3 3-26 3.8.43.4 Seismic Category I HVAC Ducts and Supports 3.8-26 3.8.4.4 Design and Analysis Procedures 3&26 3.8.4.4.1 Reactor Building, Control Building and Radwaste Building Substructures 3.8-26 3.8.4.4.2 Seismic Category I Cable Trays, Cable Tray Supports, and Conduit Supports 3.8 26 3.8.4.4.2.1 Seismic Category I HVAC Ducts and Supports 3.8-26 3.8.4.5 Structural Acceptance Criteria 3 & 26 3.8.4.5.1 Reactor Building ?. 8-26 33.4.5.1.1 General Criteria 3.8-26 3.8.4.5.1.2 Materials Criteria 3.8-27 3.8.4.5.2 Control Building 3 & 27 3.8.4.53 Radwaste Building Substructure 3 & 27 3.8.4.5.4 Seismic Category I Cable Tray and Conduit Support 3 & 27 3.8.4.5.5 Seismic Category I HVAC Ducts and Supports 3.8-27 3.8.5 Foundations 3.8-27 3.8.5.1 Description of the Foundations 3.8-27 3.8.5.2 Applicable Codes, Standards and Specifications 3.8-27 3.8.53 Loads and Load Combinations 3.8-27.1 3.8.5.4 Design and Analysis Procedures 3 & 27.1 ;
3.8.5.5 Structural Acceptance Criteria 3.8-28 3.8.5.6 Materials, Quality Control, and Special Construction Techniques 3.8-28 3.8-viii Amendment ;
ABWR mame Standard Plant anv n SECTION 3.8 CONTENTS (Continued)
Section Title Page 3.8.5.7 Testing and Insenice Inspection Requirements 3.8-28 3.8.6 COL License Information 3.8-28 3.8.6.1 Foundation Waterproofing 3.8-28 3.8.6.2 Site Specific Physical Properties and Foundation Settlement 3.8-28 Section Iife Page 3.8-sili.1 Amendment
9 MM 23A6100AE l' Standard Plant REv. n SECTION 3.8 j TABLES Table Title P_ age 3.8-1 Load Combinations, Load Factors, and Acceptance .;
Criteria for the Reinforced Concrete Containment 3.8-29 y 3&2 Major Allowable Stresses in Concrete and .
Reinforcing Steel 3.8-30 !
3.8-3 Stress Intensity Limits 3.8-31 3&4 Codes, Standards, Specifications, and Regulations
, Used in the Design and Construction of Seismic Category I Internal Structures of the Containment 3.8-32 3&S Load Combinations, Load Factors, and Acceptance .
Criteria for the Reinforced Concrete Structures .i Inside the Containment 3 & 34 -!
3.8-6 Load Combinations, Load Factors, and Acceptance Criteria for Steel Structures inside the Containment 3.8-35 ,
3.8-7 Load Combinations for Foundation Design -3 & 36 3.8-8 Welding ~ Activities and Weld Examination Requirements ,
for Containment Vessel 3 & 36.1- .!
ILLUSTRATIONS !
Figure Iills Page j 3&1 Reactor Building Arrangement Floor B2P 3&37 Elevation -1700mm 3!
3&2 Reactor Building Arrangement Floor B3F - 3&37 -' ;j Elevation -8200mm 3 & 38. .;
i 3&3 Reactor Building Arrangement Floor B1F ;
Elevation 4800mm - 3.8-39 ,
3.8-4 Reactor Building Arrangement Floor IF , .,
Elevation 12300mm 3.8-40
]
3.8-5 Reactor Building Arrangement Floor 2F ]
Elevation 18100mm 3 & 41 I
3&ix -]
1 Amendment .!
1
'I
- )
1
ABWR MMMAE Standard Plant arv n SECTION 3.8 ILLUSTRATIONS (Continued)
Figure Iille Page 3.8-6 Reactor Building Arrangement Floor 3F Elevation 23500mm 3.8-42 3.8-7 Reactor Building Arrangement Floor 4F Elevation 31700mm 3.8-43 3.8-8 Reactor Building Arrangement-Elevation 38200mm 3.8-44 3.8-9 Typical Section of Containment Liner Plate and Anchor 3.8-45 3.8-10 Deleted 3.8-46 3.8-11 Deleted 3.8-47 3.8-12 Deleted 3.8-48 3.8-13 Deleted 3.8-49 3.8-14 Deleted 3.8-50 3.8-15 RB Containment Upper Drywell Equipment Hatch 3.8-51 3.8-16 RB Containment Drywell Head 3 & 52 3.8-17 RB-RCCV Internal Structures Nomenclature 3.8-53 3.8-18 RB-RCCV Configuration 3.8-54 3.8-19 Soil Pressure on Reactor Building 3.8-55 3.8-20 Annual Temperature Profile of Suppression Pool Water During Normal Operations of a Typical Plant in Southern States 3.8-56 3.8-x Amendmer.t
ABWR ==u Standard Plant uv. n 3.8.1 CONCRETECONTAINMENT interference with the containment wall reinforce-ment.
The containment structure is designed to house the primary nuclear system and is part of the containment system whose functional requirement is to confine the potential release of radioac- The containment wall is a right, circular cyl-tive materialin the event of LOCA. This subsec- inder,2 m (6 feet,7 inches) thick, with an tion describes the concrete containment struc- inside radius of 14.5 m (47 feet,7 inches) and ture. Steel components of the containment that has a height of 29.5 m (96 feet,9 inches) mea-resist pressure and are not backed by structural sured from the top of the foundation mat to the concrete are discussed in Subsection 3.8.2. A bottom of the containment top stab. The main detailed description of the containment system is reinforcement in the wall consists of inside and presented in Subsection 6.2. outside layers of hoop and vertical reinforce-ment and radial bars for shear reinforcement.
3.!L1.1 Description of the Containment 3.8.1.1.1 Concrete Containment The containment is shown in the summary report Reinforcement is placed at major discontinu-contained in Section 3H.1. This report contains ities in the wall including the intersection of a description of the containment, figures, loads, the wall and foundation mat, the vicinity of the load combinations, concrete stresses, rein- wall intersection with the top slab, around forcement stresses, and liner strains for the major piping penetrations, the upper drywell concrete containment vessel, equipment hatch and personnel airlock, the lower drywell equipment hatch and personnel airlock The structural system is a low-leakage rein- tunnels, and suppression chamber access hatch.
forced concrete structure with an internal steel liner in the drywell and suppression chamber to serve as a leak-tight membrane. The containment is a cylindrical shell structure which is divided by the diaphragm floor and the reactor pedestal The containment top slab is nominally 2.2 m into an upper drywell chamber, a lower drywell (7 feet,2 inches) thick. The slab thickness is chamber and a suppression chamber. The top slab increased to 2.4 m (7 feet,11 inches) beneath of the containment is'an integral part of the the fuel pool, steam dryer and steam separator fuel pool with the pool girders rigidly connected pool, and around the drywell head opening.
to the containment top slab and the reactor building walls. The reactor h9? ding floors that The containment top slab main reinforcement abut the containment are integrated structurally consists of a top and bottom layer of reinforce-with the concrete containment. The containment ment. The top layer of reinforcement is ar-foundation mat is continuous with the reactor ranged in a rectangular grid. The bottom layer building foundation mat. The containment wall, of reinforcement is arranged in a rectangular top slab, reactor building floor slabs and grid and then is bent near the containment wall foundation mat are constructed of cast-in-place, into a radial pattern to avoid interference with conventionally reinforced concrete. the containment wall vertical reinforcement.
Hoop reinforcement is provided in the area of The containment foundation mat is 5.5m (18 the drywell head opening.
feet) thick. The foundation mat reinforcement consists of a top layer of reinforcement, a bottom layer of reinforcement, and vertical shear reinforcement. The bottom layer of reinforcement is arranged in a rectangular grid. The top layer of reinforcement is arranged in a rectangular grid at the center of the mat and then radiates outward in a polar pattern in order to avoid Amendment 38-1
ABWR = =n Standard Plant anv.n 3.8.1.1.2 Containment Liner Plate Boiler and Pressure Vessel Code, Division 2, Section 111, Subsection CC.
The internal surface of the containment is lined with welded steel plate to form a leaktight 3.8.1.2.3 General Design Criteria, Regulatory barrier. The liner plate is fabricated from Guides, and Industry Standards carbon steel except that stainless steel plate or clad is used on wetted surfaces of the suppres- (1) 10 CFR 50, Appendix A " General Design '
sion chamber. Criteria for Nuclear Power Plants" Criteria 1, 2, 4, 16 a n d 5 0. Conformance is The liner plate is stiffened by use of discussed in Section 3.1.
structural sections and plates to carry the design loads and to anchor the liner plate to the (2) U.S. Nuclear Regulatory Commission (NRC) l concrete, as shown Figure 3.8-9. The liner plate Regulatory Guides. Regulatory Guide 1.136.
is thickened locally and additional anchorage is Materials, Construction and Testing of provided at major structural attachments such as Concrete Containment.
penetration sleeves, structural beam brackets, the RPV pedestal and the SRV quencher support (3) Industry Standards connection to the basemat, and the diaphragm Nationally recognized industry standards floor connection to the containment wall. such as those published by the American
- Society for testing and Materials (ASTM) and The erection of the liner is performed using the American National Standards Institute standard construction procedures. The (ANSI) as referenced by the Applicable containment wall liner and top slab liner are Codes, Standards, and Regulations are used.
used as a form for concrete placement. The liner on the bottom of the suppression chamber and 3.8.1.2.4 Containment Boundary lower drywell is placed after the foundation mat concrete is in place. The jurisdictional boundary for application of Section 111, Division 2 of the ASME code to 3.8.1.2 Applicable Codes, Standards, and the concrete containment is shown in Figure Specifications 311.1-2. The boundary extends to the: l The design, fabrication, construction, (1) Outside diameter of the containment wall testing, and inservice inspection of the contain- from the foundation mat to the containment ,
ment conforms to the applicable codes, standards, top stab.
specifications, and regulations listed below, except where specifically stated otherwise. (2) The foundation mat within the outside diameter of the containment wall.
3.8.1.2,1 Regulations (3) The containment top slab from the drywell (1) Code of Federal Regulations, Title 10, head opening to the outside diameter of the Energy, Part 50, " Licensing of Production containment wall.
and Utilization Facilities."
(4) The intersection of the RPV pedestal on top (2) Code of Federal Regulations (CFR), Title 10 of the basemat.
- Energy, Part 100, Reactor Site Criteria, (10 CFR 100), including Appendix A thereto, (5) The intersection of the diaphragm floor with
" Seismic and Geologic Siting Criteria for the containment wall.
Nuclear Power Plants."
The concrete containment pressure boundary is 3.8.1.2.2 Construction Codes of Practice limited to the cylindrical wall of the drywell and wetwell, and the drywell top slab.
American Society of Mechanical Engineers (ASME) 3.8-2 Amendment
ABM 23xetooxe Standard Plant REv A (3) To - Thermal effects and loads during They are included in ASME code jurisdiction normal operating, startup or shutdown boundary for design, material, fabrication, conditions, including liner plate expansion, inspection, testing, stamping, etc., requirements equipment and pipe reactions, and thermal of the code. However, the fuel pool girders and gradients based on the most critical any other structural components which are transient or steady state thermal gradient.
integral with the containment structure are treated the same as the containment only as far (4) Ro - Pipe reactions during normal as loads and loading combinations are concerned operating or shutdown conditions based on in the design. Similaryly, the reactor building the most critical transient or steady state floor slabs that are integrated with the conditions.
containment are not included in the ASME code jurisdictional boundaries but are treated the (5) Po Pressure loads resulting from the same as the containment only as far as loads and pressure difference between the interior and load combinations are concerned. exterior of the containment, considering both interior pressure changes because of The reactor pedestal and diaphragm floor slab, heating or cooling and exterior atmospheric which partitition the containment into drywell pressure variations, and suppression chamber, are not part of the containment boundary. The reactor pedestal, (6) Construction Loads - Construction loads are steel structures filled with concrete, and the loads which are applied to the containment diaphragm floor slab are designed according to from start to completion of construction.
codes given in Sections 3.8.3 and 3.8.4, The definitions for D, L and To given respectively, above are applicable, but are based on actual construction methods and/or Those portions of the structure outside the conditions.
indicated code jurisdictional boundary will be designed, analyzed and constructed as indicated (7) SRV - Safety / Relief Valve Loads.
in Sections 3.8.3, 3.8.4 a n d 3.8.5. The Oscillatory dynamic pressure loadings analytical models will include both the resulting from discharge of safety / relief containment and reactor building and therefore valves (SRVs) into the suppression pool.
will provide continuity in the analysis. The development of these loads is with the methods described in Appendix 3B of this 3.8.13 Loads and Load Combinations SSAR. The reactor building vibration (RVB) dynamic effects shall be included in the The containment is analyzed and designed for load combinations. The number and all credibic conditions of loading including combinations of valves that will open during normal loads, preoperational testing loads, loads a reactor vessel pressure transient are as during severe environmental conditions, loads follows:
during extreme environmental conditions and loads during abnormal plant conditions. (a) G3 - Design pressure load on the suppression pool boundary resulting from 3.8.13.1 Nonnal Loads discharge of one safety / relief valve into the suppression pool. First (1) D - Dead load of the structure and equipment actuation and subsequent actuation shall plus any other permanent loads, including be considered.
vertical and lateral pressure's of liquids.
(b) G2 - Design pressure load on the (2) L - Live loads, including any moveable suppression pool boundary resulting from equipment loads and other loads which vary discharge of two adjacent safety / relief in intensity and occurrence, such as forces valves, first actuation, into the exerted by the lateral pressure of soil. suppression pool.
Amendment 384
ABM 234sicaan Standard Plant nev it (c) G AR - Design pressure load on the LOCA (LBL), based upon the calculated peak suppression pool boundary resulting from pressure with an appropriate margin.
discharge of all SRVs, first actuation, into the suppression pool. (5) P. - Design accident pressure load within th'e containment generated by an intermediate (d) ADS - Design pressure load on the break LOCA (IBL).
suppression pool boundary resulting from the SRV automatic depressurization (6) P - Design accident pressure load within system (ADS) discharge into the th'e containment generated by a small break suppression pool. LOCA (SBL).
3.8.13.2 Preoperational Testing Imads (7) Y - Local effects on the containment due to a LOCA. The local effects shall include the (1) P, - Test loads are loads which are following:
applied during the structural integrity test. (a) Y - Load on the containment generated by the reaction of a ruptured high ener.
(2) T, - Thermal effects and loads during the gy pipe during the postulated event of structural integrity test. the DBA. The time-dependent nature of the load and the ability of the contain-3.8.133 Severt Environmental Loads ment to deform beyond yield shall be considered in establishing the structu-I W - Loads generated by the design wind ral capacity necessary to resist the specified for the plant site as defined in effects of Y,.
Section 3.3.
(b) Y. - Load on the containment generated 3.8.13.4 Extreme Environmental Loads by the jet impingement from a ruptured high energy pipe during the postulated (1) E' - Safe shutdown earthquake (SSE) loads as event of the DBA. The time-dependent defined in Section 3.7. nature of the load and the ability of the containment to deform beyond yield (2) W' - Loads generated by the tornado shall be considered in establishing the specified in Section 3.3. structural capacity necessary to resist the effects of Y..
3 3.8.13.5 Abnormal Plant Loads (c) Y - The load on the containment re-(1) F - Hydrostatic load due to post-LOCA su3ing from the impact of a ruptured fiboding of the containment to a level of high energy pipe during the DBA. The 131 feet above the top of the foundation mat type of impact, for example plastic or for fuel recovery subsequent to a design elastic, together with the ability of basis accident. the containment to deform beyond yield shall be considered in establishing the (2) R - Pipe reactions (including R ) from structural capacity necessary to resist th*ermal conditions generated by a 00CA. the impact.
(3) T - Thermal effects (including T ) and (8) CO - An oscillatory dynamic loading-lo* ads generated by a LOCA.~ (condensation oscillation), on the suppression pool boundary due to steam (4) P - Design accident pressure load within condensation at the vent exits during the th"e containment generated by large break period of high steam mass flow through the vents following a LOCA.
(9) CHUG - An oscillatory dynamic loading 3.8-4 Amendmem
ABM 23AsiooAn Standard Plant REV.A 38.1.4.1.1.3 Major Penetrations (CECAP), described in Appendix 3C, is used to determine the extent of concrete cracking at The major penetrations in the containment wall these sections, and the concrete and rebar include the upper drywell equipment and personnel stresses and liner plate strains. The CECAP hatches, the lower drywell equipment and program models a single element of unit height, personnel tunnels and hatches, the suppression unit width, and depth equal to the thickness of chamber access hatch, and the main steam and the wall or slab. The calculations used in feedwater pipe penetrations. The state of stress CECAP assume that the concrete is isotropic and and behavior of the containment wall around these linear clastic but with zero tensile strength, openings is determined by the use of analytical CECAP also can calculate the reduced thermal numerical techniques. The analysis of the area forces and moments due to concrete cracking.
around the penetrations consists of a non-linear However, the redistribution of forces and three dimensional finite element analysis with moments are not calculated. To account for the boundaries extending to a region where the concrete cracking effects on the redistribution discontinuity effects of the opening are of forces and moments, an interactive procedure negligible. described in Subsection 3.8.1.4.1.3 is used.
Displacements compatible with the global The input data for the CECAP program consist analysis of the containment are applied at these of the membrane forces, shear forces and bending boundaries. The stresses and strains in the moments calculated by the STARDYNE analysis.
reinforcement, concrete and liner plate are The areas of the reinforcing steel in terms of obtained from the local finite element mode. The steel area to concrete cross section ratio are analysis considers concrete cracking and thermal based on a trial design shown in Apper. dix 3H.
strains. When the calculated stresses in the steel and in the concrete meet the acceptance criteria, the 3 1 1.4.1.1.4 Variation of Physical Material trial design is adequate; otherwise, the trial Properties design is revised. The evaluation of contain-ment structural adequacy is shown in Subsection In the design analysis of the containment 3.8.1.5.
consideration is given to effects of possible variations in the physical properties of 3 1 1.4.1.3 Concrete Cracking Considerations materials on the analytical results. A high initial value of the concrete modulus of The membrane forces, shear forces and bending elasticity is used in the analysis, which will moments in the containment structure subjected conservatively determine the upper bound effects to loads are obtained by applying the STARDYNE from thermal stresses in the structure. The computer program to the finite element model variation of the soil modulus of elasticity and that was developed on the basis of an uncracked Poisson's ratio of concrete is considered in the section. This model is called the uncracked analysis in order to constitute the most severe model. In sizing the reinforcing steel or in structural responses. calculating the rebar stresses, the concrete is not relied upon for resisting tension. Thus, 3.8.1.4.1.2 Design Methods those portions of structures which are either in membrane tension or in flexural tension are The design of the containment structure is cracked to transfer loads from concrete to based on the membrane forces, shear forces and rebar. The CECAP program, described in Appendix bending rroments for the load combinations defined 3C, is used for calculating the extent of in Subsection 3.8.1.3.6. The membrane forces, concrete cracking and the stresses in the shear forces and bending moments in selected concrete and in the steel. Because of concrete sections are obtained by the computer program cracking that leads to stiffness changes, the STARDYNE as described in Section 3.8.1.4.1.1. distribution of forces and moments is different The selected sections are shown in Figure 3.8-14. from those calculated from the uneracked model described above. To determine the effects due The Concrete Element Cracking Analysis Program to concrete cracking a revised finite element Amendment 3 B-7
ABWR mmu Standard Plant Rev n model called the " cracked model', that takes into The suppression pool contains air-saturated, account the concrete cracking, was developed for stagnant, high purity water and is designed for applying the STARDYNE computer program. The a 60. year life. The amount of corrosion is
" cracked model" is the same as the original based on the annual temperature profile of noneracked model except that the element suppression pool water for a typical plant in thicknesses for the diaphragm floor (D/F) slab southern states under normal operation (Figure and the top slab are reduced to reflect concrete 3.8-20). The following conditions can cause the cracking due to bending while the element pool temperature to rise above normal:
thicknesses for the reactor buildir.g (R/B) floor slab and walls are reduced to reflect the tension (1) Reactor core isolation mode: pool, cracks. In this model, the thicknesses for the temperature can rise 17"C above normal RPV pedestal and reactor shield wall are not for a total of 165 days during the 60-year reduced because they are steel structures (with life; and concrete fill). The basemat and the columns are predominately subjected to compression, so no (2) Suppression pool cooling mode: pool cracking is assumed, temperature can rise 17 C t.bove normal for l a total of 540 days during the 60-year lifetime.
The corrosion allowance for type 304L stain-less steel in air-saturated water for any oxygen level and temperatures up to 316 C for 60 A comparison indicates that the forces and/or years is 0.12 mm. The major concern has bending moments from the uncracked model are involved the air / water interface area where controlling in some of the regions of the pitting is most likely to occur. The 0.12 m I structure and the results from the cracked model co:rosion allowance is a small fraction of the are controlling in other regions. stainless steel thickness which will be a nominal 2.5 mm if clad carbon steel plate is l '
The procedures for the design and analysis of used.
the liner plate and its anchorage system are in accordance with the provisions of the ASME Code Water used to fill the suppression pool is Section til Div. 2 Subarticle CC-3600. The liner either condensate or demineralized. No chem-plate analysis considers deviations in geometry icals are added to the suppression pool water, due to fabrication and ' erection tolerances, and variations of the assumed physical properties of Observations made on suppression pool water the liner and anchor material. The strains and quality over a period of several years indicate stresses in the liner and its anchors are within that periodic pool cleaning such as by allowable limits defined by the ASME Code, underwater vacuuming will be required as well as Section Ill, Div. 2. Subarticle CC-3720. the use of the suppression pool cleanup system to maintain water quality standards. The 3.8.1.4.1.4 Corrosion Prevention suppression pool cleanup system (Subsection 9.5.9) also acts to maintain purity levels.
Type 304L stainless steel or clad carbon steel plate will be used for the containment An ultrasonic thickness measurement program liner in the wetted areas of the suppression pool will be performed to detect any general as protection against any potential pitting and corrosion at underwater positions. A visual co.trosion on all wetted surfaces and at the water examination for local pitting on the underwater
-t o-air interface area. The advantage of portions of the steel containment will be made cladding is greatly reduced pool maintenance / at refueling outages using underwater lighting operating costs over the life of the plant and and short focus binoculars. This covers 10% of crud removal and burial costs over the life of the surface at the first refueling outage after the plant. the start of commercial operation, 5%
384 Amendment
ABWR uu -
any n Standard Plant additional surface approximately two to five 3.8.1.6.1 Concrete years later and 5% at five-year intervals there-after. If pits are detected at any examination, All concrete materials are approved prior to representative ones are ultrasonically tested and start of construction on the basis of their the depth of those large enough for measurement characteristics in test comparisons using ASTM will be determined. Appropriate repairs can be standard methods. Concrete aggregates and made as required. cement, conforming to the acceptance criteria of the specifications, are obtained from approved 3.8.1.4.2 Ultimate Capacity of the Containment sources. Concrete properties are determined by laboratory tests. Concrete admixtures are used An analysis is performed to determine the to minimize the mixing water requirements and ultimate capacity of the containment. The increase workability. The specified compressive results of this analysis are summarized in strength of concrete at 28 days, or earlier, is:
Chapter 19.
Specified Strength 3.8.1.5 Structural Acceptance Criteria Structure f'c (twD For evaluation of the adequacy of the contai'2- Containment 4000 ment structural design, the major allowable Foundation Mat 4000 l
stresses of concrete and reinforcing steel for service load combinations and factored load All structural concrete is batched and placed combinations according to ASME III, Div. 2 are in accordance with Subarticle CC-2200 and shown in Table 3.8-2. Article CC-4000 of the ASME Code,Section III, Division 2.
The allowable shear stresses are extracted from ASME III, Div. 2. The allowable values are (a) Cement low when only the orthogonal reinforcement system is considered in resisting tangential shear. Cement is Type II conforming to the However, the draft of the ASME Code,Section III, Specification for Portland Cement (ASTM C Div. 2. revision has adopted higher allowable 150). The cement contains no more than values (500 psi). Because this draft has passed 0.60 percent by weight of alkalies the ASME Ill design committee, the revised calculated as sodium oxide plus 0.658 allowable shear stresses are used for the potassium oxide. Certified copies of containment structural adequacy evaluation to material test reports showing the chemical justify that no diagonal reinforcement is to be composition and physical properties are used in resisting lateral shear loads. obtained for each load of cement delivered.
For sites where concrete may come into con-tact with soils having more than 0.20% wa-ter soluble sulfate (as SO4 ) of ground-3.8.1.6 Material, Quality Control, and Special water with a sulfate concentration exceed-Construction Techniques. ing 1500 ppm, only Type V cement shall be used unless other suitable means are Materials used in construction of the con- employed to prevent sulfate attack and tainment are in accordance with Regulatory Guide concrete deterioration.
1.1.36 and the ASME Code,Section III, Division
- 2. Article CC4000. fpecifications covering all (b) Aggregates materials are in sufficient detail to assure that the structural design requirements of the work All aggregates conform to the Specification are met. for Concrete Aggregates (ASTM C 33).
l l
1 3.8-9 Amendmen.
.g ABWR 2mme Standard Plant Rev.n (c) Water meeting the design requirements specified for concrete are used.
Water and ice for mixing is clean, with a total solids content of not more than 2000 3.8.1.6.2 Reinforcing Steel ppm as measured by ASTM D 1888. The mixing water, including that contained as free Reinforcing bars for concrete are deformed water in aggregate, contains not more than bars meeting requirements of the Specification 250 ppm of chlorides as Cl as determined by for Deformed and Plain Billet Steel Bars for ASTM D 512. Chloride ions contained in the Concrete Reinforcement (ASTM A 615, Grade 60).
aggregate are included in calculating the Mill test reports, in accordance with ASTM A total chloride ion content of the mixing 615, are obtained from the reinforcing steel water. The chloride content contributed by supplier to substantiate specification the aggregate is determined in accordance requirements.
with ASTM D 1411.
The test procedures are in accordance with t (d) Admixtures - ASTM A 370, and acceptance standards are in accordance with ASTM A 615.
The concrete may also contain an air-en-training admixture and/or t. water-reducing 3.8.1.6.3 Splices of Reinforcing Steel admixture. The air-entraining admixture is in accordance with the Specification of Air Sleeves for reinforcing steel mechanical Entraining Admixtures for Concrete (ASTM C splices conform to ASTM A 513, A519 or A 576 260). It is capable of entraining 3 to 6 Grades 1008 through 1030. Certified copies of percent air, is completely water soluble, material test reports indicating chemical compo-and is completely dissolved when it enters sition and physical properties are furnished by the batch. Superplasticizers, entraining the manufacturer for each sleeve lot.
from 1.5 to 4.5 percent air, may be used in concrete mixes (f' = 5000 psi, maximum) for Placing and splicing of reinforcing bars is l congested areas to improve workability and in accordance with Article CC-4300 and Subarti- t prevent the formation of voids around rein- cle CC-3530 of the ASME Code,Section III, forcement. The water-reducing admixture Division 2.
conforms to the standard specification for Chemical Admixtures for Concrete (ASTM C 3.8.1.6.4 Liner Plate and Appurtenances 494), Types A and D. Type A is used when average ambient temperature for the daylight The materials conform to all applicable period is below 700F. Type D is used when requirements of the ASME Code,Section III, average ambient air temperature for the day- Division 2.
light period is 700F and above. Pozzo-lans, if used, conform to Specification for Steel plate is tested at the mill in full Fly Ash and Raw or Calcined Natural Pozzo- conformance to the applicable ASTM specifica-lans for Use in Portland Cement Concrete tions, and certified mill test reports are sup-(ASTM C 618) except that the loss on igni- plied for review and approval. The plate is tion shall be limited to 6 percent. A d mix- visually examined for laminations and pitting.
tures containing more than 1 percent by Identity of the plate is maintained throughout weight chloride ions are not used. fabrication.
(c) Concrete Mix Design Dimensional tolerances for the crection of the liner plate and appurtenances are detailed Concrete mixes are designed in accordance in the Construction Specification based on the with ACI 211.1, Recommended Practice for structure geometry, liner stability, concrete Selecting Proportions for Normal and Heavy strength, and the construction methods to be Weight Concrete, using materials qualified used.
and accepted for this work. Only mixes Amendment 3.8-10
MM 23A6100AE am n
~
l Standard Plant 3.8.1.6.5 Quality Control Subsubarticle CC-6230 of the ASME Code, Section :
III, Division 2. See Subsection 3.8.6.3 for COL i Quality control procedures are established in license information. l the Construction Specification and implemented [
during construction and inspection. The 3.8.1.7.2 Preoperational and Inservice l t
Construction Specification covers the Integrated leak Rate Test fabrication, furnishing, and installation of each structural item and specifies the inspection and . Preoperational and inservice integrated leak ,
documentation requirements to ensure that the rate testing is discussed in Subsection 6.2.1.
requirements of the ASME Code,Section III, ,
Division 2, and the applicable Regulatory Guides 3.8.2 STEEL COMPONENTS OF THE :
are met. REINFORCED CONCRETE l CONTAINMENT !
3.8.1.6.6 Welding Methods and Acceptance Criteria for Containment Vessel Liner and 3.8.2.1 Description of the Containment Appurtances The ABWR has a reinforced concrete.
Welding methods and acceptance criteria for containment vessel (RCCV) as described in .!
the containment vessel liner and appurtance are Subsection 3.8.1. This section will describe .;
the same as those for the steel components of the the following steel components of the concretc ;
concrete containment vessel (i.e., personnel air containment vessel:
locks, equipment batches, penetrations, and drywell head) given in Subsection 3.8.2.7.1. (1) Personnel Airlocks 3.8.1.7 Testing and Inservice Inspection (2) Equipment Hatches, I
Requirements (3) Penetrations j 3.8.1.7.1 Structural Integrity Pressure Test (4) Drywell Head A structural integrity test of the l i
containment structure will be performed by the 3.8.2.1.1. Description of Penetrations
)
COL applicant in accordance with Article CC-6000 of the ASME Code,Section III, Division 2 and The penetrations through the RCCV include the Regulatory Guide 1.136, after completion of the ' following.
j containment construction. The' test is conducted at 115% of the design pressure condition of 45 3.8.2.1.1.1 Personnel Airlacks i
psig in both the drywell and suppression chamber, '
Two personnel air locks with an inside simultaneously.
differential pressure A pressure condition test of for 25 the psigdesign _ diameter sufficient to provide 6 ft., 8 in. . ,
between the drywell and the suppression chamber high by 3 ft., 6 in., wide minimum clearance ;
i is also performed where the drywell pressure is above the floor at the door way are provided.
greater than the suppression chamber pressure. One of these air locks provides access to the l upper dry well and the other provides access to l During these tests the suppression chamber and the lower drywell via the access tunnel, i spent fuel pool are filled with water to the.
normal operational water level. Deflection and Lock and swing of the doors is by manual and concrete crack measurements are made to determine automatic means. The locks extend radically that the actual structural response is within the outward from the RCCV into the reactor building limits predicted by the design analysis. and are supported by the RCCV only. The minimum clear horizontal distance not impaired by the In addition to the' deflection and crack door swing is 6 ft.
measurements, the first prototype containment structure is instrumented for the measurement of Each personnel air lock has two pressure-strains in accordance with the provisions of seated doors interlocked to prevent simultaneous 3.8-11 Amendment
ABWR 23461004e Standard Plant nrv. a opening of both doors and to ensure that one door is completely closed before the opposite door can be opened. The design is such that the interlocking is not defeated by postulated malfunctions of the electrical system. Signals and controls that indicate the operational status of the doors are provided. Provision is 3
3 8-II I Amendment
/ DIN 23A6100AE nov, n Standard Plant made to permit temporary bypassing of the door 3.8.2.1.1.3 Other Penetrations interlock system during plant cold shutdown. The door operation is designed and constructed so The RCCV penetrations are categorized into either door may be operated from inside the two basic types. These types differ with containment vessel, inside the lock, or from respect to whether the penetration is subjected "
outside the containment vessel. to a hot or cold operational environment.
The lock is equipped with a digital readout The cold penetrations pass through the RCCV pressure transducer system to read inside and wall and are embedded directly in it. The hot outside pressures. Quick acting valves are penetrations do not come in direct contact with provided to equalize the pressure in the air lock the RCCV wall but are provided with a thermal when personnel enter or leave the containment sleeve which is attached to the RCCV wall. The vessel. The personnel air locks have a double thermal sleeve is attached to the process pipe scaled flange with provisions to pressure test at distance from the RCCV wall to minimize the space between the seals of the flange. conductive heat transfer to the RCCV wall.
Besides piping penetrations, several electri-cal penetrations also exist. A description of the various penetrations is given Chapter 8.
3.8.2.1.1.2 Equipment flatch 3.8.2.1.1.4 Drywell liesd Three equipment hatches are provided. Two A 10300 mm in diameter opening in the RCCV or l drywell equipment hatches, one of these serves upper drywell top slab over the RPV is covered the upper drywell and the other serves the lower with a removable steel toruspherical drywell I drywell via the access tunnel, The third head which is part of the pressure boundary.
l equipment hatch is 20000 mm provides personnel The drywell head is designed for removal during and equipment access to the suppression chamber reactor refueling and for replacement prior to airspace. reactor operation using the reactor building crane. One pair of mating flanges is anchored The equipment hatch covers have a double in the drywell top slab and the other is welded sealed flange with provisions to pressure test integrally with the drywell head. Provisions the space between the seals of the flange. A are made for testing the flange seals without means for removing and handling the equipment pressurizing the drywell. Figure 19F.3-1 shows l hatch cover is provided. The hoisting equipment the drywell head.
and hoisting guides are arranged to minimize con-tact between the doors and seals during opening and closing. The equipment hatch includes the electric-motorized hoist with pushbutton control stations, lifting slings, hoist supports, hoist-ing guides, access platforms, and ladders for ac-cess to the dogged position of the door and hoist, latches, seats, dogging devices, and tools required for operation and maintenance of the hatch.
The equipment hatches and covers are entirely supported by the RCCV. Figure 3.8-15 show gen- ,
eral details of the equipment hatch and cover.
l 3.8-12 Amendment
N 23A6100AE )
Standard Plant am n <
?
(7) bar and machine steel (A576, carbon content not less than 0.3%); and (8) clad (SA-240 type 304L).
The structural steel materials located beyond the containment vessel boundaries are as follows: ,
(1) carbon steel (A36 or SA-36) and (2) stainless steel extruded shapes (SA-479).
The. materials meet requirements as specified .
in Subarticle NE-2000 of ASME Code Section III.
The lowest service metal temperature is 300 F. ,
3.8.2.7 Testing and Inservice Inspection Requirements Leakage of the containment vessel, including
- - the steel components is described in Subsection 3.8.1.7.
3.8.2.7.1 Welding Methods and Acceptance Criteria Welding activities shall be performed in accordance with requirements of Section III of t the ASME Code. The required nondestructive -
examination and acceptance criteria are provided in Table 3.8-8. ,
t
.' 3 3.8.2.7.2 Shop Testing Requhements I The shop tests of the personnel air locks in- -
clude operational testing and an' overpressure test. After completion of the personnel air -
locks tests' (including all latching mechanisms and interlocks) each-lock is given,an ,
operational test . consisting:of repeated - !
operating of-each door and mechanism to determine whether a!! parts are operating. L, smoothly without binding or other deiects. . All ;
defects encountered are corrected and retested.
The process of testing, correcting defects, and retesting are continued until no defects are' detectable. .
.I Amendment 3M. .{
i
.. A
ABWR 2346 oorn !
Standard Plant anv n l
For the operational test, the personnel air Figures 3.8-17 and 3.8-18 and Figures 1.2-2 locks are pressurized with air to the maximum through 1.2-13 show an overview of the permissible code test pressure. All welds and containment including the internal structures.
seals are observed for visual signs of distress j or noticeable leakage. The lock pressure is then The summary report contained in Appendix 311.1 reduced to design pressure and a thick bubble contains the figures for the reactor pedestal solution is applied to all welds and seals and and the diagphram slab. Including but not observed for bubbles or dry flaking as indica- limited to structural steel details, rein-tions of leaks. All leaks and questionable areas forcement details, loads, load combinations, are clearly marked for identification and concrete stresses, reinforcement stresses, liner subsequent repair. stresses, and structural shell stresses.
During the overpressure testing, the inner 3.83.1.1 Diaphragm Floor door are blocked with holddown devices to prevent unseating of the seals. The internal pressure of The diaphragm floor serves as a barrier the lock is reduced to atmospheric pressure and between the drywell and the suppression all leaks are repaired. Afterward, the lock is chamber. It is a reinforced concrete circular again pressurized to the design pressure with air slab, with an outside diameter of 14.5 m (47 ft, and all areas suspected or known to have leaked 7 in), and a thickness of 1.2 m (3 ft,11 in).
during the previous test are retested by the bubble technique. This procedure is repeated The diaphragm floor is supported by the reac-until no leaks are discernible, tor pedestal and the containment wall. The con-nection of the diaphragm floor to the contain-3.8.3 Concrete and SteelInternal ment wall is a fixed support. The diaphragm Structures of the Concrete floor connection to the reactor pedestal is a Containtnent hinged support. The r'.iaphragm floor is pene-trated by 18,508 mm (20 in) diameter sleeves 3.83.1 Description of the Internal Structures for the SRV lines.
The functions of the containment internal structures include: support of the reactor vessel radiation shielding, support of piping and A 1/4 inch thick, carbon steel liner plate is equipment, and formation of the pressure provided on the bottom of the diaphragm floor, suppression boundary. The containment internal and is anchored to it. The liner plate serves structures are constructed of reinforced concrete as a form during construction and prevents the and structural steel. The containment internal bypass flow of steam from the upper drywell to structures include the following: the suppression chamber air space during a LOCA.
L (a) Diaphragm floor 3.83.1.2 Reactor Pedestal (b) Reactor pedestal A composite steel and concrete pedestal pro-vides support for the reactor pressure vessel, (c) Reactor shield wall the reactor shield wall, the diaphragm floor, access tunnels, horizontal vents, and the lower (d) Drywell and equipment pipe support structure drywell access platforms. The pedestal consists of two concentric steel shells tied together by (e) Miscellaneous platforms _
vertical steel diaphragms. The regions formed by the steel shells and the vertical diaphragms, (f) Lower drywell equipment tunnel except the vents and the vent channels, are filled with concrete. There are ten drywell (g) Lower drywell personnel tunnel connecting vent (DCV) channels connecting the upper drywell to the lower drywell and the !
(h) Reactor shield wall stabilizer horizontal vents. ,
l l
3.8-16 !
Amendment I
. ., q 23A6100AE I Standard Plant nrw n The wetted portion of the exterior surface of the reactor pedestal steel shell in the suppression chamber is clad with stainless steel to provide corrosion protection. The extent of the cladding and the reactor pedestal configuration is provided in Figure 1.2-2.
Amendment 3116.1
MN 23A6100AE REv. n Standard Plant _
The design and analysis is based on the elas- loads. All safety-related items which thel ;
tic method. All loads are resisted by the integ- inelastic beam deformations may effect are q ral action of the inner and outer steel shells. , evaluated to verify that no required _ safety i The concrete placed in the annulus between the function would be compromised.J ~ ~
j inner and outer shells acts to distribute loads between the steel shells, and provides stability 3.83.4.5 Other internal Structures to the compression elements of the pedestal.
The design and analysis procedures used for 3.83.43 Reactor Shield Wall other internal structures are similar to those used for the drywell equipment and pipe st.pport The design and analysis procedures used for structure as described in Subsection 3.83.4.4.
the reactor shield wall are similar to those used for the reactor pedestal described in Subsection 3.83.5 Structural Acceptance Criteria 3.8.3.4.2.
3.83.5.1 Drywell Equipment and Pipe Support 3.83.4.4 Drywell Equipment and Pipe Support Structure Structure .
_b The structural acceptance criteria for the 4 The drywell equipmert and pipe support struc- DEPSS are in accordance with ANSI /AISC-N690. If b l ture (DEPSS) is designed using the AISC working ; the frame type pipe support requirements of st r e ss m ethods for ste el s afe ty-r ela t e d { Subsection 3.733.4 can not be met at some pipe structur_e_s_ for nuclear _ facilities (ANSI /AISC- I supports, amplification factors will be N690).jThe DEPSS is designed to support the developed for the ARS for the piping group.
deadweight of equipment and piping. Since there ,
~
- are no safety-related equipment supported on the { 3.83.5.2 Other Internal Structures IG'5d DEPSS. The DEPSS is segregated into two The structural acceptance criteria for
- o. sections.
internal concrete or steel structures are in One section supports the drywell cooling accordance with ACI-349 and ANSI /AISC-N690, system and related chilled water piping, HVAC respectively.
duct work, and work platforms. This system is classified as nonsafety.related. 3.83.6 Materials, Quality Control, and Special Construction Techniques The other section provides a support point for f pipe hangers per Subsection 3.7.3.3.4, and - 3 .83.6.1 Diaphragm Floor provides a support point for pipe whip restraints. This section is classified as The materials, quality control, and construc-safety-related. The structural group will tion techniques used for the diaphragm floor and provide amplification factors to the piping liner plate are the same as those used for the group for those safety-related pipe support containment wall and liner plate in Subsection points that do not met either the stiffness orj 3.8.1.6.
displacement criteria found in Subsection ,
3.7.3.3.4. Those beams and columns supporting 3.83.6.2 Reactor Pedestal pipe supports will carry piping dynamic loads while remaining clastic. Those beams supporting The materials conform to all applicable pipe whip restraints allow inelastic deformations requirements of ANSI /AISC N690 and ACI 349 and due to pipe rupture loads. All safety-related comply with the following:
items which the inelastic beam deformations may Ikm Soccification effect are evaluated to verify that no required safety function would be compromised. The beams Inner and outer shells ASTM A441 or A572
} and columns supporting pipe supports will carry Internal stiffeners ASTM A441 or A572 1, piping dynamic loads while remaining elastic.
i Those beams supporting pipe whip restraints allow k inelastic deformations due to pipe ru_pture Concrete fill f' c = 4000 psi n.-- - -
~ _ _ _ _ _ _
l 3 8-19 I Amendment
insert a The DEPSS is designed to support the deadweight of nonsafety-related equipment and support safety-related and nonsafety-related piping. The nonsafety-related equipment are the drywell cooling coils and fans. The safety-related items include safety-relief valves, mainsteam isolation valves, ECCS isolation valves, and feedwater check valves. In addition the DEPSS provides access platforms such that all of these pieces of equipment ;
can be accessed, inspected, and removed from the drywell if neccessary. The DEPSS is a l 2 level, 3D space frame, consisting of columns, radial beams, circumferential beams, and steel grating. The DEPSS is shown in figures 1.2-3,1.2-3a,1.2-13a,1.2-13b, and 1.2-13c.
1 l
The DEPSS provides piping systems within the drywell a stable platform for pipe suppon, and pipe whip restraints. It is designed in accordance with ANSI /AISC-N690.' In
, addition, the criteria given in 3.7.3.3.4 is applied to the DEPSS. If the criteria can not be met, the COL applicant will generate the ARS at piping attachment points considering the DEPSS as part of structure using the dynamic analysis methods described in subsection i l
3.7.2, or will analyze the piping systems treating the DEPSS as part ofpipe support.
Thcs.,s oeams and columns supporting pipe supports will carry piping dynamic loads without buckling and while remaining elastic. Those beams and columns supporting pipe i whip restraints allow inelastic deformations due to pipe rupture loads. All safety-related j items which the inelastic beam deformations may effect are evaluated to verify that no required safety function would be compromised.
insert b The structural acceptance criteria for the DEPSS design are in accordance with ANSI /AISC-N690.
l l
23A6100AE Standard Plant nev. n 3.83.63 Reactor Shield Wall The materials conform to all applicable requirements of ANSI /ASIC Nf 90 and ACI 349 and comply with the following:
11rm Snecit' cation Inner and outer shells ASTM A441 or A572 Internal stiffeners ASTM A441 or A572 Concrete fill f' c = 4000 psi Stainless Steel Clad SA-240 Type 304 L 3.83.6.4 Drywell Equipment and Pipe Support Structure The materials conform to all applicable requirements of ANSI /AISC N690 and comply with the following:
111m Specification Structural steel and ASTM A36 or A441 connections or A572 3&l9.1 )
Amendment
l ABWR u-n i Standard Plant uvn !
l High strength structural ASTM A572 or A441 3.83.8 Welding Methods and Acceptance Criteria steel plates for Structural and Building Steel Bolts, studs, and nuts ASTM A325 or A490 Welding activities shall be accomplished in (dia. 2_3/4 ") accordance with written procedures and shall meet the requirements of the American Institute ,
Bolts, studs, and nuts ASTM A307 of Steel Construction (AISC) Manual of Steel (dia. 5. 3/4 ") Construction. The visual acceptance criteria shall be as defined in American Welding Society 3.83.6.5 Other laternal Structures (AWS) Structural Welding Code D1.1 and Nuclear Construction Issue Group (NCIG) Standard, The materials conform to all applicable Visual Weld Acceptance Criteria for Structural requirements of ANSI /AISC N690 and comply with Welding at Nuclear Plants, NCIG-01.
the following:
3.8A OTHER SEISMIC CATEGORY I 11ral Specification STRUCTURES Miscellaneous platforms Same as Section Other Seismic Category I structures which 3.83.6.4 constitute the ABWR Standard Plant are the reactor building, control building and radwaste Lower drywell equipment ASTM A516 Grade 70 building substructure. Figure 1.2-1 shows the tunnel SA-240 Type 304 L spatial relationship of these buildings. The only other structure in close proximity to these Lower dryweli personnel ASTM A516 Grade 70 structures is the turbine building. They are tunnel SA-240 Type 304 L structurally separated from the other ABWR Standard Plant buildings.
Reactor shield wall stabilizer The Seismic Category I structure within the
--tube sections ASTM A501 ABWR Standard Plaut, other than the containment structures, that contains high-energy pipes are
--plates ASTM A36 the reactor building and control building. The steam tunnel walls protect the reactor building lower drywell floor fill A material other and control building from potential impact by material than limestone rupture of the high-energy pipes. This building concrete is designed to accommodate the guard pipe support forces.
3.83.7 Testing and Inservice Inspection Requirements The reactor building, steam tunnel, residual heat removal (RHR) system, reactor water cleanup A formal program of testing and inservice in- (RWCU) system, and reactor core isolation cool-spection is not planned for the internal struc- ing (RCIC) system rooms are designed to handle tures except the diaphragm floor, reactor pedes- the consequences of high energy pipe breaks.
tal, and lower drywell access tunnels. The other The RHR, RCIC, and RWCU rooms are designed for internal structures are not directly related to differential compartment pressures, with the the functioning of the containment system; associated temperature rise and jet force, therefore, no testing or inspection is performed. Steam generated in the RHR compartment from the postulated pipe break exits to the steam tunnel Testing and inservice inspection of the dia- through blowout panels. The steam tunnel is phragm floor, reactor pedestal and lower drywell vented to the turbine building through the access tunnels are discussed in Subsection seismie interface restraint structure (SIRS).
3.8.1.7. The steam tunnel, which contains several pipe-lines (e.g., main steam, feedwater, RHR), is al-so designed for a compartment differential pres-sure with the associated temperature changes and jet force.
Amendment 3&20
ABWR mame nuv. n Standard Plant Seismic Category I masonry walls are not used in the design. The ABWR Standard Plant does not contain seismic Category I pipelines buried in soil.
3.8.4.1 Description of the Structures 3.8.4.1.1 Reactor Building Structure The reactor building (RB) is constructed of reinforced concrete with a steel frame roof. The RB has four stories above the ground level and three stories below. Its shape is a rectangle of 59 meters in the E-W direction,56 meters in the N-S direction, and a height of about 57.9 meters from the top of the basemat.
3.8-20.1 Amendraent
ABM 2u61 min Standard Plant REv n The Reinforced Concrete Containment Vessel The CB is a shear wall structure designed to (RCCV) in the center of the RB encloses the accomodate all seismic loads with its walls.
Reactor Pressure Vessel (RPV). The dCCV supports Therefore, frame members such as beams or the upper pool and is integrated with the RB columns are designed to accomodate deformations structure from the basemat up through the of the walls in case of earthquake conditions.
elevation of the RCCV top slab. The interior floors of the RB are also integrated with the The summary report for the control building RCCV wall. The RB has slabs and beams which join is in Appendix 3H.2. This report contains a the exterior wall. Columns support the floor description of the control building, the loads, slabs and beams. The fuel pool girders are load combinations, reinforcement stresses, and integrated with the RCCV top slab and with RB concrete reinforcement details for the base mat, wall. columns. The RB is a shear wall structure seismic walss, steam tunnel, and floors.
designed to accommodate all seismic loads with its walls. Therefore, frame members such as 3.8.4.1.3 Radwaste Building Substructure beams or columns are designed to accommodate deformations of the walls in case of earthquake The radwaste building substructure (RWB) is conditions. shown in Section 1.2.
The summary report for the reactor building is The radwaste building is a reinforced in Appendix 3H.1. This report contains a concrete structure 60.4m by 41.2m by 29.5m description of the reactor building, the loads, high. The building consists of a below grade load combinations, reinforcement stresses, and substructure consisting of walls and slabs of concrete stresses at locations of interest. In reinforced concrete,1.2m thick, forming a rigid addition, the report contains reinforcement box structure which serves as a container to details for the basemat, seismic walls, and fuel hold radioactive waste in case of an accident.
pool girders. This substructure is located below grade to increase shielding capability and to maximize 3.8.4.1.2 Control Building safety. It is supported on a separate foundation mat whose top is 13.8m below grade. l The control building (CB) is located between In addition, a reinforced concrete super-the reactor building and the turbine building. structure 15.7m high extends above grade and l It is shown in Section 1.2. houses the balance of the radwaste equipment.
The CB houses the essential electrical, The radwaste building substructure houses the control and instrumentation equipment, the high and low conducivity tanks, clean-up phase control room for the reactor and turbine separators, spend resin storage tanks, a buildings, the CB HVAC equipment, RB cooling concentrated waste storage tank, distillate tank water pumps and heat-exchangers, the essential and associated filters, and pumps for the switchgear, essential battery rooms, and the radioactive liquid and solid waste treatment steam tunnel. systems.
The CB is a Seismic Category I structure that Although the radwaste superstructure is not a houses control equipment and operation personnel Seismie Category I structure, its major and is designed to provide missile and tornado structural concrete walls and slabs are designed protection. The CB is constructed of reinforced to resist Seismic Category I loads.
concrete with a steel roof. The CB has two stories above the ground level and four stories The summary report for the radwaste building below. Its shape is a rectangle of 56 m (183 is in Appendix 3H.3. This report contains a feet,8 inches) in the E-W direction,24 m (78 description of the radwaste building, the loads, feet, 9 inches) in the N-S direction, and a load combinations, reinforcement stresses, and height of about 22 m (72 feet 2 inches) from the concrete stresses at locations of interest, in top of the base mat. addition, the report contains reinforcement detailas for the basemat, seismic walls, and floors.
Amendment 3A21
ABWR uuixxu Standard Plant REV H 3.8.4.1.4 Seismic Category I Cable Tray, Cable (6) NRC publications TID 7024 and TID 25021, Tray Supports, and Conduit Supports Nuclear Reactors and Earthquakes and Summary of Current Seismic Design Practice for Electrical cables are carried on continuous Nuclear Reactor Facilities; horizontal and vertical runs of steel trays supported at intervals by structural steel (7) The inservice inspection requirements for frames. The tray locations and elevations are the fuel pool liners in the Reactor Building predetermined based on the requirements of the are in conformance with ASME Code Section electrical cable network. Generally, several III, Division 2.
trays of different sizes are grouped together and connected to a common support. (8) NRC Regulatory Guides:
The support frame spacing is determined by (a) Regulatory Guide 1.10, Mechanical allowable tray spans, which are governed by (Cadweld) Splices in Reinforcing Bars of rigidity and stress. The frames may be Category I Concrete Structures; ceiling-supported, or wall-supported, or a combination of both. Various type of frames form (b) Regulatory Guide 1.15, Testing of a support system with transverse and longitudinal Reinforcing Bars for Category I Concrete bracing to the nearest wall or ceiling to take Structures; the scismic loads.
(c) Regulatory Guide 1.28, Quality Assurance 3.8.4.1.5 Seismic Category I IIVAC Ducts and Program Requirements (Design and Supports Construction);
HVAC ducts are supported at intervals by structural steel frames The duct locations and elevations are predetermined based on the requirements of the HVAC system.
3.8.4.2 Applicable Codes, Standards, and Specifications 3.8.4.2.1 Reactor Building The major portion of the reactor building, is not subjected to the abnormal and severe acci-dent conditions associated with a containment. A listing of applicable documents follows:
(1) ACI 349, Code Requirements for Nuclear Safety Related Concrete Structures; (2) AISC, Specification for Design, Fabrication and Erection of Structural Steel for B uildin gs; (3) ASME Boiler and Pressure Vessel Code Section III, Subsection NE, Division 1, Class MC (for design of main steam tunnel embedment ;
piping anchorage in the RB and CB only); j (4) AWS Structural Welding Code, AWS D1.1; (5) AWS Structural Welding Code, AWS D12.1; 3.8-21.1 Amendmen' I
e
ABWR ummie Standard Plant nev. n (d) Regulatory Guide 1.29, Seismic Design Classification; (c) Regulatory Guide 1.31, Control of (c) Deleted Stainless Steel Welding; (f) Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel; (f) ANSI N45.4, Leakage-Rate Testing of Containment Structures for Nuclear (g) Regulatory Guide 1.55, Concrete Reactors; Placement in Category I Structures; (g) ANSI N101.2, Protective Coatings (h) Regulatory Guide 1.60, Design Response (Paints) for Light Water Nuclear Spectra for Seismic Design of Nuclear Reactor Containment Facilities; and Power Plants; (h) ANSI N101.4, Quality Assurance for (i) Regulatory Guide 1.61, Quality Assurance Protective Coatings Applied to Nuclear Requirements for the Design of Nuclear Facilities; Power Plants; (10) Steel Structures Painting Council Standards (k) Regulatory Guide 1.69, Concrete Radiation-Shields for Nuclear Power (a) SSPC-PA-1, Shop, Field and Maintenance Plants. Painting; (1) Regulatory Guide 1.76, Design Basis (b) SSPC-PA-2, Measurement of Paint Film Tornado. Thickness with Magnetic Gages; (m) Regulatory Guide 1.142, Safety-Related (c) SSPC-SP-1, Solvent Cleaning; Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and (d) SSPC-SP-5, White Metal Blast Cleaning; Containment); and (e) SSPC-SP-6, Commercial Blast Cleaning; (n) Regulatory Guide 1.94, Quality Assurance and Requirements wr Installation, Inspec-tion, and Testing of Structural Concrete (f) SSPC-SP-10, Near-White Blast Cleaning; and Structural Steel During the Con-struction Phase of Nuclear Power Plants. (11) ACI-ASCE Committec 326, Shear and Diagonal Tension, ACI Manual of C.mcrete Practice, (9) ANSI: Part 2; (a) ANSI A58.1, Building Code Requirements (12) Applicable ASTM Specifications for for Minimum Design Loads in Building and Matcrials and Standards; and Other Structures; ,
(13) AASHTO Standard Specifications for Highway I (b) ANSI N5.12, Protective Coatings (Paint) Bridges for truck loading area. l for the Nuclear Industry; 3.8.4.2.2 Control Building (c) NOA-1, Quality Assurance Program Requirements for Nuclear Facilities and P.efer to Subsection 3.8.4.2.1.
NOA-1A, Adenda to ANSI /ASME NOA-1; Add NRC Rules and Regulations Title 10, Chap-(d) Deleted ter 1, Code of Federal Regulations, Part 73.2 and 73.55.
1 l
Amendmen' 33-22
ABWR m imse Standard Plant nwn 3.8.4.23 Radwaste Building Substructure 3.8.4.2.63 Welding of Refuel Cavity and Spent Fuel Pool Liners The radwaste building substructure shall be designed using the same codes and standards as Welding activities shall be accomplished in the reactor building. Refer to Subsection accordance with the AWS Structural Welding Code, 3.8.4.2.1 for a complete list. D1.1. The welded seams of the liner plate shall be spot radiographed where accessible, liquid In addition, the non-Seismic Category I penetrant and vacuum box examined after reinforced concrete portion of the fabrication to ensure that the liner does not superstructure is designed according to the leak. The acceptance criteria for these seismic provisions of Section 2314 of the examinations shall meet the acceptance criteria uniform building code. stated in subsection NE-5200 of Section III of the ASME Code.
3.8.4.2.4 Seismic Category I Cable Trays, Cable Tray supports, Conduit and Conduit Supports 3.8.43 Loads and Load Combinations '
% '(1) All codes, standards, and specifications / 3.8.43.1 Reactor Building MSdY applicable to the building structures shall' a% also apply to cable tray and conduit The temperature and pressure loads caused by a LOCA do not occur on the reactor building.
{ supports. -J The reactor building ventilation system is (2) AISI SG-673, Specification for the Design of designed to keep the building within operating Cold-formed Steel Structural Members. design conditions.
(3) NEMA, Fittings and Supports for Conduit and 3.8.43.1.1 Loads and Notations Cable Assemblies.
Loads and notations are as follows:
3.8.4.2.5 Seismic Category i IIVAC Ducts and s dead load of structure plus any D
Supports
=
other permanent load
% All codes, standards, and specifications trasst applicable to the building structures shall also L = conventional floor or roof live loads, movable equipment loads, b ' -
apply to HVAC ducts and supports.
___ and other variable loads such as 3.8.4.2.6 Welding and Weld Acceptance Criteria eonstruction 1oads. The following live loads are used:
3.8.4.2.6.1 Welding of Electrical Cable Tray and Condult Supports Concrete floors and slabs (including roofs) - 200 psf.
Welding activities shall be accomplished in Stairs, stair platforms, grating accordance with the AWS Structural Welding Code, floors, and platforms - 100 psf.
D1.1. The weld visual acceptance criteria shall Concrete roofs, live or snow load be as defined in AWS Structural Welding Code D1.1 (not concurrent) - 50 psf.
and NCIG-01. Construction live load on floor framing in addition to dead 3.8.4.2.6.2 Welding of fleating Ventilation and weight of floor - 50 psf *,
Air Conditioning Supports ~
R = pipe reactions during normal Welding activities shall be accomplished in operating or shutdown conditions accordance with the AWS Structural Welding Code, based on the most critica1 D1.1. The weld visual acceptance criteria shall transient or st eady-st a t c be as defined in AWS Structural Welding Code D1.1 eondition.
and NCIG-01.
l l
Amendment I
u
]
insert a .]
i (1) ANSI /AISC-N690," Specification for Design,. Fabrication and Erection -f of Steel Safety-Related Structures for Nuclear Facility" l I
insert b (1) ASME/ ANSI AG-1, " Code on Nuclear Air and Gas Treatment" l (2) ANSI /AISC-N690, " Specification for Design, Fabrication _and Erection l
of Steel Safety-Related Structures for Nuclear Facility" :
i n
'i
[
.+
i i
i i
r i
I i
a t
9 t
l
.t
-e 4
ld M NAM 00AE any. n Standard Plant R = pipe reactions under thermal conditions generated by the postulated break and including R
o Y
= equivalent static load on a structure generated by the reaction on the broken high-energy pipe during the postulated break and including a calculated dynamic factor to account fro the dynamic nature of the load.
Y. = jet impingement equivalent static I load on a structure generated by the postulated break and including a calculated dynamic factor to account for the dynamic nature of the load.
Y = missile impact equivalent static
- load on a structure generated by or during the postulated break, like pipe whipping, a nd including a. calculated dynamic factor to account for the dynamic nature of the load.
W =
wind force (Subsection 33.1)
If the actual constmction live load is greater than this value a design check of the structures nill be made.
Amendment 3&B1
GVR BA61%AE '
Standard Plant RE u 1.5 psi negative pressure differ- T = thermal effects (including T )
ential, or (2) 3 psi negative which may occur during a desi$n pressure differential. accident at 165 F maximum 30 minutes after LOCA P = internal negative pressure of 3.0 psig due to tornado; accident U = for concrete structures, the section pressure at main steam tunnel piping strength required to resist design embedment loads based on the strength design method described in ACI 318.
B = uplift forces created by the rise of the ground water table H = loads caused by static or seismic earth pressures.
F = internal pressures resulting from flooding of compartments For structural steel, S is the required sec-tion strength based on the elastic design me-E' = safe shutdown earthquake (SSE) loads thods and the allowable stresses defined in Part as defined in Section 3.7 1 of the AISC Specification for the Design, Fabrication and Erection of Structural Steel for Buildings.
T = thermal effects - 1oad effects 3.8.4.3.1.2 Load Combinations for Concrete induced by normal thermal gradients Members existing through the reactor building wall and roof. Both summer For the load combinations in this subsection, and winter operating conditions are where any load reduces the effects of other considered. In all cases the loads, the corresponding coefficient for that conditions are considered of long load shall be taken as 0.9 if it can be enough duration to result in a demonstrated that the load is always present or straight line temperature gradient. occurs simultaneously with the other loads.
The temperatures are as follows: Otherwise, the coefficient for that load shall be taken as zero.
(1) Summer operation:
(1) Normal operating conditions - The strength (a) air temperature inside building - design method is used and the following load 120 F combinations are satisfied:
(b) exterior temperature - 115 F U = L4 D + 1.7 L + 1.3 T + 1.7 R +
(2) Winter operation: 1.7 H + 1.4 B (a) air temperature inside building -
70 F (b) exterior temperature -(-) 40 F U =1.4 D + 1.7 L + 1.3 T + 1.7 R +
(3) Winter shutdown 1.7 H + 1.7 W t
(a) air temperature inside building - For fluid pressure F, replace 1.7 H by 1.7 F 50 F in the last two of the three equations (b) exterior temperature - (-) 40 F above.
For all cases as-constructed temperature is (2) Abnormal / extreme environmental conditions -
60 F The strength design method is used and the 3.B-24 Amendment
~
ABWR mmon any a Etandard Plant following load combinations are satisfied: (2) Abnormal / extreme environmental conditions -
The elastic working stress design method is U= D+L+T +R +H+B used and the following load combinations are satisfied:
U= D+L+T +R + H + E' 1.6 S = D + L + T +R + E' U= D+L+T +R +H 1.6 S = D + L + T +R o
+W.t U= D+L+T +R +H+W t
1.6 S = D + L + T +Ro +W+H t U= D+L+T+R* + *1.5 P* +H 1.6 S = D + L + T +R +P.
a_ a a U= D+L+T +R + 1.25 P +H 8
+ 1.25 E + "(Y + Y.3 + Y ")" 1.6 S = D + L + T +R + 1.0 E + .
P*+ ( Y .I + "Y ' +ay"),
U= D+L+T +R +P+H*
+ E' + (Y' + Y. % Y" } 1.6 S = D + L + T a +R a
+ E' + P a
3
+ (Y.3 + Y ' +y*),
3.8.4.3.1.3 Load Combinations for Steel Members in all these load combinations, both cases of L having its full value or being completely (1) Normal operating conditions - the elastic absent are checked, working stress design method is used for the following load combinations: 3.8.4.3.2 Control Ilullding and Radwaste Building Substructure S=D+L.
Refer to the loads, notations, and combina-tions established in Subsection 3.8.4.3.1, except that fluid pressure F, accident pressure S=D+L+W P , and pipe break loads Y , Y., Y do n$t exist and the live loads ar'c asI follEws:
Since thermal stresses due to T and R are present and are secondary and seli-limitin$ All concrete lloors - 400 psf in nature, the following combinations are also satisfied: Stairs, stair platforms, grating floors, and platforms - 100 psf 1.5 S = D + L + T +R.
Roof live or snow load (non concurrent) - 50 psf 1.5 S = D + L + T + W. Construction live load on floor framing in addition to dead weight of floor - 50 psf' In all these load conditions, both cases of L having its full value or being T = thermal effects. As-constructed completely absent are checked. temperature is 600 F. The temperatures inside the building are as follows:
Amendment 7 3.8-25
r 1 ABWR msicau Standard Plant uv. n Operating Conditions Appendix 3H.3. This summary report contains a section detailing safety factors against Controi room, summer 75 F 0 sliding, over turning, and floatation.
winter 700 F HVAC room, summer 950 F 3.8.4.4.2 Seismic Category I Cable Trays, winter 600 F Cable Tray Supports, and Conduit Supports Other areas, summer 750 F D d_
winter 750 F Cable trays, conduits and their supports will 6 be analyzed using the computer codes listed in Shutdown condition Appendix 3C. The seismic design of the cable trays, conduits and their supports will be Control room, summer 80 F 0 performed by one of the methods discussed in i 500 F Subsection 3.7.3.
- winter HVAC room, summer 1040 F 1 winter 500 F 3.8.4.4.2 Seismic Category I HVAC Ducts and Other areas, summer 900 F Supports inot d winter 50 0F HVAC ducts and supports will be analyzed 3.8.433 Seismic Category 1 Cable Trays, Cable using the computer codes listed in Appendix 3C.
Tray Supports, and Conduit Supports The seismic design of the HVAC ducts and
- supports shall be performed by one of the
")
~
~.. __
s Loads and load combinations for the tray and methods described in Subsection 3.7.3.
conduit supports shall use the same criteria as g
c h for the building structures where the supports are located. __
3.8.4.5 Structural Acceptance Criteria 3.8.4.5.1 Reactor Building 3.8.43.4 Seismic Category I HVAC Ducts and Supports 3.8.4.5.1.1 General Criteria
'% I Loads and load combinations for HVAC ducts ~ The first criterion is that the reactor 6aDF and supports shall use the same as for the! building shall provide biological shielding for cl building structures where the supports are plant personnel and the public outside of the r site boundary. This criterion dictates the minimum wall and roof thicknesses.
3.8.4.4 Design and Analysis Procedures The second criterion is that the reactor 3.8.4.4.1 Reactor Building, Control Building, building shall protect the reinforced concrete and Radwaste Building Substructure containment from environmental hazards such as tornado and other site proximity-generated The reactor building, control building and missiles. The shielding thicknesses are radwaste building substructure will be designed sufficient for this purpose.
in accordance with ACI-349 for concrete struc-tures and ANSI /AISC-N690 for steel structures. The reactor building provides a means for collection of fission product leakage from the The reactor building, control building, and reinforced concrete containment following an radwaste building substructure will be analyzed accident.
using the computer codes listed in Appendix 3C.
The reactor building SGTS is designed to keep The foundation for category I structures is the compartments surrounding the reinforced contained in the summary reports for their concrete containment at a negative pressure even respective buildings. The reactor building after a LOCA. In order to achieve a maximum ,
foundation is contained in Appendiz 3H.1, the in-leakage rate of 50% per day under a pressure control building foundation is in Appendix 3H.2, and the radwaste building foundation is in l
3.8-26 Amendment i
insert c. !
Loads used in dynamic analysis for tray and conduit supports are the following:
-i D + L = 75 lbs/ linear foot used for 18 inch tray, 50 lbs/ linear. foot l used for 12 inch and narrower tray Dynamic Loads - SSE plus other RBV dynamic loads I
Load' combinations used in dynamic analysis for the tray and conduit supports are the following:
D+L ,
where D, L, SSE, and RBV are defined in subsection 3.8.4.3.1.1.
i insert d .;
Loads and load combinations used for dynamic analysis for HVAC l ducts and supports are the following: {
I D + L + Po .!
where D, L, SSE, and RBV are defined in subsection 3.8.4.3.1.1,.and Po'is the internal pressure of the HVAC duct.
insert e All seismic category 1 cable trays and conduit supports are designed by ,
one of the methods discussed in subsection 3.7.3 or by design by rule methods as approved by the NRC. Design by rule methods will be based on documented performance of conduit and cable trays.during'prio'r qualification tests or analysis or exposure to natural seismic disturbances. If.an-analysis is performed it will use one of the codes listed in appendix 3C. 1 insert f All seismic category 1 HVAC duct and duct supports are designed by.
ene of the methods discussed in subsection 3.7.3 or by design by rule methods as approved by the NRC. Design by rule methods will be based on documented performance of HVAC ducts during prior qualification tests ox analysis or exposure to natural seismic disturbances. If an analysis is performed it will use one of the codes listed in appendix 3C. ,
}
h i
- i I
t D
m
ABWR 23umn nev.n Standard Plant f
l differential of 6 mm of water, the reinforcing 3.8.5.1 Description of the Foundations steel is designed to remain clastic during the SSE load combinations. The foundations of the reactor building and control building are reinforced concrete mnt 3.8.4.5.1.2 Materials Criteria foundations.
Refer to the materials criteria established in These two foundation mats are separated from 3.8.5 for the strength and materials requirements each other by a separation gap of 2 meters (6 for the reinforced concrete reactor building. feet,6 inches) wide to minimize the structural interaction between the buildings.
3.8.4.5.2 Control ituilding The reactor building foundation is a Structural acceptance criteria are defined in rectangular reinforced concrete mat 56 m (183 feet,8 inches) by 59 m (193 feet,5 inches) and l the ANSI /AISC-N690 and ACI M9 Codes. In no case does the allowable stress exceed 0.9 F where 5.5m (18 feet) thick. The foundation mat is F is the minimum specified yield strc5s. The constructed of cast-in-place conventionally dlsign criteria preclude excessive deformation of reinforced concrete. It supports the reactor the building. The clearances between adjacent building, the containment structure, the reactor buildings are sufficient to prevent impact during pedestal, and other internal structures. The a seismic event. The tornado load analysis for top of the foundation mat is approximately 20.2 this building is the same as the analysis for the m(66 feet,3 inches) below gtade.
reactor building.
The containment structure foundation, defined .
3.8.4.5.3 Radwaste lluilding Substructure as within the perimeter or the exterior surface '
of the containment structure, is integral with Structural acceptance criteria are defined in the reactor building foundation. The containment foundation mat details are discussed l ANSI /AISC-N690 and ACI 349 Codes. In no case in Subsection 3.8.1.1.1.
does the allowable stress exceed 0.9F where #
F is the minimum specified yield stress. The design criteria preclude excessive deformation of The control building foundation is the building. The clearance between adjacent rectangular reinforced concrete mat 24 m by 56 m buildings are sufficient to prevent impact during by 5.5 m. The top of the foundation mat is 13.2 a seismic event. m below grade.
3.8.4.5.4 Seismic Category 1 Cable Trays and The radwaste building ioundation is a rectangular reinforced concrete mat 60.4m by Conduit Supports f g ogg M m n iec4ed 6 ey on 41.2m and 2.5m thick. The top of the radwaste Structural acceptanceAcriteriagre defined in building mat is 13.8m below grade. The ANSI /AISC-N690 Code. in no case does t$ foundation mat is constructed of cast-in-place allowable stress exceed ).~ F where F is C conventionally reinforced concrete. It supports Y the radwaste building structure.
the minimum specified yield tTress.
o.Ct 3.8.4.5.5 Seismic Category i IIVAC Duct and The foundation for category 1 structures is Supports contained in the summary reports for their K -. _ -
respective buildings. The reactor building mg Structural acceptance criteria are defined in foundation is contained in Appendiz 3H.1, the ANSI /AISC-N690 Code. In no case does the control building foundation is in Appendix 3H.2, 3g allowable stress exceed 0.9 F where F is and the radwaste building foundation is in Y Appendix 311.3. This summary report contains a the minimum specified yield stIess.]- section detailing safety factors against 3.8.5 Foundations sliding, over turning, and floatation.
This section describes foundations for all 3.8.5.2 Applicable Codes, Standards and seismic Category I structures of the ABWR Specifications Standard Plant.
3.8 27 Amendment i i
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The structural acceptance criteria for HVAC' ducts if the analysis option I is selected will be in accordance with ANSI /ASME AG-1 Code. The HVAC L
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3.8.5.3 Loads and Load Combinations The loads and load combinations for the containment foundation mat are given in S u bse c t io n 3.8.1.3. The loads and load combinations for the other seismic category I structure foundations are given in Subsection 3.8.4.3.
The loads and load combinations for all seismic Category I foundations examined to check against sliding and overturning due to carth-quakes, winds and tornados, and against flo-tation due to floods are listed in Table 3.8-7.
3.8.5.4 Design and Analysis Procedures The foundations of seismic category I structures are analyzed using well established methods where the transfer of loads from the foundation mat to the supporting foundation media is determined by clastic methods.
Bearing walls and columns carry all the vertical loads from the structure to the foundation mat. Lateral loads are transferred to shear walls by the roof and floor diaphragms.
The shear walls then transmit the loads to the foundation mat.
The design of the mat foundations for the structures of the plant involves primarily determining shear and moments in the reinforced concrete and determining the 3.8-27.1 Amendmcni
MWR 234siv>xe >
Standard Plant uv. n interaction of the substructure with the The calculated and allowable factors of.
underlying foundation medium. For a mat safety of the ABWR structures for overturning, foundation supported on soil or rock, the sliding, and flotation are shown in Appendix 3H pertinent aspects in the design are to maintain for each foundation mat.
the bearing pressures within allowable limits, particularly due to overturning forces, and to 3.8.5.6 Materials, Quality Control, and '
ensure that there is adequate frictional and Special Construction Techniques passive resistance to prevent sliding of the structure when subjected to lateral loads. The foundations of seismic Category I structures are constructed of reinforced '
The design loads considered in analysis of the concrete using proven methods common to heavy foundations are the worst resulting forces from industrial construction. For further discussion the superstructures and loads directly applied to see Subsections 3,8,1.6 and 3.8.4.6.
the foundation mat due to static and dynamic load combinations. 3.8.5.7 Testing and Inservice Inspection Requirements The capability of the foundation to transfer shear with waterproofing will be evaluated. See A formal program of testing and inservice Subsection 3.8.6.1 for COL license information inspection is not planned and is not required requirements. for the seismic Category I structures of the ABWR.
The standard ABWR design is developed using a range of soil conditions as detailed in Appendix 3.8.6 COL License Information 3A. The variations of physical properties of the site-specific subgrade materials will be 3.8.6.1 Foundation Waterprooling determined (see Subsection 3.8.6.2). Settlement of the foundations, differential settlement The capability of foundations to transfer between foundations for the site-specific shear loads where foundation waterproofing is foundations medium will be calculated and used will be evaluated (see Subsection 3.8.5.4).
safety-related systems (i.e., piping, conduit, etc.) will be designed for the calculated 3.8.6.2 Site Specific Physical Properties and settlement of the foundations. The effect of the Foundation Settlement site-specific subgrade stiffness and calculated settlement on the design of the seismic Category Physical properties of the site specific I structures and foundations will be evaluated. subgrade medium shall be determined and the See Subsection 3.8.6.2 for COL license settlement of foundations and structures information requirements. including seismic Category I will be evaluated see Subsection 3.8.5.4).
A detailed description of the analytical and design methods for foundation mats including the 3.8.6.3 Structural Integrity Pressure Result containment foundation, is included in Appendix 3H. Each COL applicant will perform the structural integrity test (SIT) of the ABWR containment in 3.8.5.5 Structural Acceptance Criteria accordance with Subsection 3.8.1.7.1. Addition-ally, the first ABWR containment is considered The main structural criteria for the contain- as a prototype and its SIT performed accord.
ment portion of the foundation are adequate ingly. The details of the test and the instru-strength to resist loads and sufficient stiffness mentation as required for such a test will be to protect the containment liner from excessive provided by the first COL applicant for NRC strain. The acceptance criteria for the review and approval.
containment portion of the foundation mat are presented in Subsection 3.8.1.5. The structural acceptance criteria for the reactor building foundations are described in Subsection 3.8.4.5.
3.R-28 Amendment J
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- 13A6100AE REV.C Standard Plant Table 3,8-8 WELDING ACTIVITIES AND WELD EXAMINATION REQUIREMENTS FOR ,
CONTAINMENT VESSEL (I)(2)(3) ,
Component Weld Type NDE Requirements ,
Containment Category A, butt welds (Long1) RT r Containment Category B, butt welds (Circ.) RT ,
Containment Category C, butt welds RT -
Containment Category C, nonbun welds UT or MT or PT 7 Containment Category D, bun welds RT i Containment Category D, nonbuu welds UT or MT or PT Containment Structural attachment welds a) Butt welds RT:
b) Nonbun welds UT or MT orIYr SpecialWelds Weld metalcladding PT t
3 NOTES:
(1) Le required confirmation that facility welding acevities are in compliance with the requirements will include ' ,
the following third-pany verifications: ,
(a) Facility welding specifications and procedures meet the applicable ASME Code requirements:
(b) Facility welding activities are performed in accordance with the applicable ASME Code requirements; . j (c) Welding activities related records are prepared, evaluated and maintained in accordance with the ASME -
Code requirements; ,
(d) Welding processes used to weld dissimilar base metal and welding filler metal combinations are compatible for the intended applications; (e) The facility has established procedures for qualifications of welders and welding operators in accordance !
with the applicable ASME Code requirements; (f) Approved procedures are available and used for preheating and post heating of welds, and those procedures -
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Table 3.8-8 WELDING ACTIVITIES AND WELD EXAMINATION REQUIREMENTS FOR ;
CONTAINMENT VESSEL (1)(2)(3) (Continued)
(2) Radiographic film will be reviewed and accepted by the licensee's nondestructive examination (NDE), level III examiner prior to final acceptance. ,
(3) The NDE requirements for containment vessels will be as stated in subarticle NE-5300 of Section III of the ASME Code. ,
LEGEND: i i
RT - Radiographic Examination MT - Magnetic Particle Examination I'T - Liquid Penetrant Examination LEGEND (Continued):
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