ML20049H267
| ML20049H267 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/12/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8202230019 | |
| Download: ML20049H267 (65) | |
Text
. -
- - _ _ _ -. _ - _ _ = _ _ _ _ _ _ _
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 i
i t
APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 1
1 4
'I 8202230019 820212 DR ADOCK 05000447 i
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I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f
4 j
l APPENDIX 3A O
CONTENTS Section Title Page 3A APPENDIX 3A - SEISMIC SOIL-STRUCTURE INTER-ACTION ANALYSIS OF-THE NUCLEAR ISLAND 3A.1-1 4
3A.1 INTRODUCTION 3A.1-1 3 A. l.1 General 3A.1-1 3A.1.2 Seismic Analysis 3A.1-1 i
J 3A.2 SELECTION OF SITE PARAMETERS AND THEIR I
RANGES 3A.2-1 3A.2.1 Introduction 3A.2-1 3A.2.2 Site Conditions 3A.2-1 3A.2.3 Cases Studied 3A.2-7 3A.2.3.1 Reactor Building 3A.2-7 3A.2.3.2 Other Nuclear Island Buildings 3A.2-8 3A.3 INPUT MOTION AND DAMPING VALUES 3A.3-1
[
3A.3.1 Input Motion 3A.3-1 I ()
3A.3.2 Damping 3A.3-2 3A.4 THE STRUCTURAL MODEL 3A.4-1 I
3A.4.1 Structural Modeling Theory 3A.4-1 3A.4.2 Finite-Element Model 3A.4-2 i
3A.4.2.1 Reactor Building Model 3A.4-2 3A.4.2.2 Other Nuclear Island Building SSI Model 3A.4-4 l
3A.5 SOIL-STRUCTURE INTERACTION ANALYSIS 3A.5-1 3A.5.1 Introduction 3A.5-1 3A.S.2 Analysis Procedure 3A.5-4 t
l 3A.6 REFERENCES 3A.6-1 l
r O
[
3A-i/3A-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1
APPENDIX 3A I
TABLES Table Title Page 3A-1 Cases Considered for Reactor Building 3A.7-1 3A-2 Strain-Compatible Soil Properties (VP P
'^'
}
^* ~
3, 5'
3A-3 Strain-Compatible Soil Properties (VP3, VP5'
'^'
^* ~
O 1
l.
i O
3A-iii/3A-iv
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O APPENDIX A p
i
)
ILLUSTRATIONS
%./
Figure Title Page 3A-1 Variation of Shear Wave Velocity with Depth at South Texas and San Joaquin Nuclear Power Plant Sites 3A.8-1 3A-2 Variation of Shear Wave Velocity with Depth at San Onofre and FFTF Nuclear Power Plant Site 3A.8-2 3A-3 Variation of Shear Wave Velocity with Depth at Three Selected Nuclear Power Plan Sites 3A.8-3 3A-4 Range of Shear Wave Velocities for Nuclear Power Plant Sites in High Seismic Areas 3A.8-4 3A-5 Shear Wave Velocity Profiles to be Used in SSI Analyses 3A.8-5 3A-6 variation of Shear Modulus and Damping Ratio with Shear Strain Used in Analyses 3A.8-6 3A-7 Building Arrangement Plan 3A.8-7 3A-8 Plan View of Nuclear Island and Turbine Building Arrangement 3A.8-8 3A-9 Control Motion - H 3A.8-9 y
,(,)
3A-10 Control Motion - H 3A.8-10 2
3A-ll Control Motion - Vertical (V) 3A.8-ll 3A-12 Response Spectra Control Motion - H 3A.8-12 y
3A-13 Response Spectra Control Motion - H 3A.8-13 2
3A-14 Response Spectra Control Motion - V 3A.8-14 3A-15 Equivalent Plane-Strain Model for Buildings in 0* Direction 3A.8-15 3A-16 Equivalent Plane Strain Model for Buildings in 90 Direction 3A.8-16 3A-17 SSI Model for Section A-A - Other Nuclear Island Buildings 3A.8-17 I
3A-18 Schematic Representation of Soil / Structure Interaction Analysis Using Finite-Element Method 3A.8-18 3A-19 Typical Soil / Structure Model for 150-ft, 0 in. Soil Layer (O Direction) 3A.8-19 3A-20 Typical Soil / Structure Model for 75-ft, 0 in. Soil Layer (0 Directon) 3A.8-20 3A-21 Typical Soil / Structure Model for r~T 90 Direction 3A.8-21 3A-v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ILLUSTRATIONS (Continued)
Figure Title Page 3A-22 Free-Field Response Spectra at Basemat Level 3A.8-22 3A-23 Finite-Element Horizontal Response Spectra at Top of Basemat 3A.8-23 3A-24 Finite Element Vertical Response Spectra at Top of Basemat 3A.8-24 9
O I
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/
APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 3A.1 INTRODUCTION 3A.l.1 General This appendix describes the generic site. conditions adopted for seismic design bases of the Nuclear Island.
The rationale for determining the selected generic site conditions is detailed.
The selected generic site data were used to evaluate the dynamic interaction effects between the final plant structures of the Nuclear Island and the underlying soil for a 0.15g operating basis earthquake (OBE) and a 0.3g safa shutdown earthquake (SSE) excita-tion.
Ti.c finite-element method was used to perform the soil-structure interaction analysis and is described in detail in this appendix.
The results of the soil-structure interaction analysis are given in the form of response spectra at the center of the top of the basemat.
These were the bases for the detailed structural response evaluation described in Section 3.7.
The generic site data and methods described in this appendix are consistent with those previously approved by US NRC but modi-fied to incorporate up-to-date data on nuclear plant sites and State-of-the art techniques to perform the soil-structure inter-action analysis.
Hence, this study provides a form of verification of the committed seismic design.3 3A.l.2 Seismic Analysis To ensure the seismic adequacy of all building structures and
\\
equipment included in the Nuclear Island, an extensive seismic i
3A.1-1
-,_.,.m
...-r--
.-----zv-------
.m
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.l.2 Seismic Analysis (Continued) analysis was performed.
This analysis for qualification of the Nuclear Island was performed considering a range of site param-eters with variations to form the generic siting conditions.
The generic site conditions were selected to provide an adequate seismic design margin for any Nuclear Island facility with site parameters within the range of this study and for tnese facilities.
Further site-unique seismic analysis or review by the regulatory agencies should not be required.
Detailed descriptions of the site parameter s celected and the procedures used to perform seismic soil-structure interaction analysis of the Nuclear Island are provided herein.
The generic 1
site parameters and their ranges have been slightly modified to incorporate up-to-date data on nuclear plant sites as described in Section 3A.2.
The analytical procedures are generally con-sistent with those described in Reference 2 and are given in more detail in Section 3.7 of Reference 3.
Seismic qualification of any Nuclear Island can be done using the results presented in this appendix provided that:
(1) the peak ground acceleration is less than or equal to 0.3g SSE, 0.15g OBE; (2) the site design response spectra are less than or equal to those given in Regulatory Guide 1.60 normalized to the peak ground accelerations in (1);
(3) there is no potential for liquefaction at the plant site due to OBE and SSE as reviewed and concurred with by the Site Analysis Branch of the NRC staff; O
3A.1-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
3A.l.2. Seismic Analysis (Continued) i (4) there is no potential.for fault displacement at the plant site as reviewed and concurred with by the NRC staff; (5) the embedment depth of the Reactor Building is between 34 and'40 feet, ( 0.5 feet excavation tolerance) for soil sites (for sites with shear wave velocities greater than f
3500 fps, there is no limitation on embedment depth);
(6) the average. shear. wave velocity for the top 30 feet of
~
soil is greater than 500 fps; 4
4 (7) for layered soil sites with parameters which have very i
abrupt variations with depth, analysis with site-unique properties should be performed to confirm the applicability of the generic analysis; and Os i
l (8) the soil-bearing capacity at the site is adequate to accommodate plant design loads.
The study described in Appendix 3A of Reference 3 showed that 1
variation of the water level, material density, material composi-tion, or soil profile did not affect overall results; therefore, no special restrictions were provided for these parameters in the present study.
To demonstrate the seismic adequacy of the Nuclear Island design for these stated site conditions, 12 soil-structure interaction cases for the Reactor Building.nd 12 cases (3 soil types and 4 cross sections) for other Nuclear Island buildings were analyzed using the finite-element method.
This appendix details the basis for selecting the cases and the method of the seismic soil-f structure interaction analyses.
A description of the input motion 3A.1-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.l.2 Seismic Analysis (Continued) and damping values, the structural model, and the soil model are included.
A description of the modal equivalence concept used to model some of the structures is given in Reference 3.
The modal equivalent model of the Reactor Building structure, the Fuel and Auxiliary Building models, and the soil model were combined to determine the dynamic response of the individual basemats of these buildings.
The translational and rotational time histories of acceleration response of the basemat were then used to determine the structural response of buildings, the reactor, and its inter-nals.
The method for determining the structural response is provided in detail in Section 3.7.
O til l
3A.1-4 L
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O A
3A.2 SELECTION OF SITE PARAMETERS AND THEIR RANGES 3A.2.1 Introduction This section describes the generic site conditions and final design parameters based on a range of soil properties used in the soil-structure interaction (SSI) analysis described in this appendix.
The basic items required for soil-structure interaction study are:
(1)
Structural Model; (2)
Embedment Depth; (3)
Input Motion; (4)
Soil Profile; and (5)
Soil Properties.
For the present studies, the structural models, the embedment r"N depth, and the input motion were the same for all cases.
The structural models were based on the final Nuclear Island design.
m-The embedment depth was 39 feet and the input motion was an arti-ficial accelerogram that was a reasonable fit to the spectrum in 4
Regulatory Guide 1.60 with 0.15g zero-period acceleration for OBE.
The input motion was applied in the free field at grade.
The soil profile and properties were varied to cover a range of expected site conditions in relatively high seismic areas where a nuclear power plant may be constructed.
3A.2.2 Site Conditions The range of site conditions are most conveniently described in terms of:
(1) total thickness of the soil profile, (2) general layering, and (3) dynamic material properties -(modulus, damping, and Poisson's ratio).
u 3A.2-1 L
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.2 Site Conditions (Continued)
The response of a soil deposit is essentially a function of the profile depth and the dynamic material properties of the fourda-tion materials.
Appropriate ranges of dynamic material properties and profile depths previously agreed upon have been used for the bSI analyses.
The ranges of properties and profile depths were defined utilizing information on site conditions at actual nuclear plants and the results of the previous studies.3 For cases involving horizontal componenr3 of excitation, the dynamic material properties having the greatest effect on response are shear modulus and damping ratio.
Both of these material properties are strain dependent.
For cases involving vertical components of excitation, the constrained modulus is the most significant parameter.
Other properties required for response computations are total unit weight and Poisson's ratio.
The maximum shear modulus of the foundation material denotes the
-4 shear modulus at very low shear strains (approximately 10 percent).
The maximum shear modulus may be defined by the shear wave velocity, v measured at low energy levels.
The variations s,
of shear wave velocity with depth for a few nuclear plant sites are presented in Figure 3A-1 through 3A-3.
It can be seen from these figures that a wide range of the site shear wave velocity profile is possible.
The shear wave velocity profiles presented in these figures may be grouped into three generalized velocity profile zones:
a soil zone (sands, silts, clays and gravelly soils), a soft rock and well cemented soil zone, and a transiPion zone.
These velocity profile zones are shown in Figure 3A-4.
For sites with insitu shear wave velocities above 3500 ft/s 1
(represented by the cross-hatched area in Figure 3A-4), SSI l
analyses are not presently required by NRC (fixed-base analysis p
mitted).
Thus, the shaded region in Figure 3A-4 represents O
3A.2-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.2 Site Conditions (Continued)
G the range of shear wave velocities for which soil-structure interaction analyses can be made.
This range of shear wave velocities shown in Figure 3A-4 was used to define six basic shear wave velocity profiles.
These velocity profiles are designated VP through VP as shown in Figure 3A-5.
Velocity profiles VP y
6 and VP bound the soil zone; VP and W bound the transition 3
3 4
zone, and VP and VP un r ckwell cemented so us zone.
4 6
The velocity profiles are smooth curves representative of the average variation of shear modulus with depth that can be expected within each of the site zones.
It is recognized that specific sites may have a velocity profile much more irregular than those shown in Figure 3A-5.
However, for purposes of the generic all-soil envelope analysis, more uniform velocity profiles best serve the purpose of evaluating the effects of parametric variations.
In addition to these six curves, three basic velocity profiles A
from Reference 3 corresponding to the lower bound (LB), average (AV), and upper bound (UB) soil properties were also included in this study (Figure 3A-5).
The shear wave velocity profiles for LB, AV, and UB soil profiles obtained from Reference 3 are based on the following Reference 3 expressions:
G (y) = 1000 K m (y)
(G v
=
g where maximum shear modulus in psf:
G
=
m(y) effective mean pressure in (psf) at depth Y:
c
=
shear wave velocity v
=
s mass density o
=
(
K m dulus parameter 2m 3A.2-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.2 Site Conditions (Continued)
The Gmax(y) expression results in a smooth, continuous curve with v
qu 1 to zero at zero depth ( m(y) = 0 at y = 0).
However, in s
the finite-element discretization, the G value (and hence v )
max s
is computed at the mid-height of each element and assumed to be the same for the total depth of the element.
Hence, at the sur-face, the value of v will correspond to the value at the mid-s height of the first finite element soil layer.
This value is equal to 265 fps for the LB case, 375 fps for the AV case, and 460 fps for the UB case with the thickness of the first element soil equal to 2.5 feet.
For the first 30-ft depth of the soil layer, the average value of v is equal to 490 fps for the LB case, 690 s
fps for the AV case and 850 fps for the UB case.
These values are consistent with Reference 3 requirements to provide an aver-age value of v about 500 fps in the top 30 feet of soil layer.
s The maximum shear modulus for the foundation materials can then be defined using the data presented in Figure 3A-5 and the following relationship:
G
=0 V
(3A-3) max s
where o is the mass density in slugs /ft v
is the shear wave velocity (Figure 3A-5) (fps) s G
i the maximum shear modulus in psf max The shear modulus of soil as well as rock is strain dependent.
Based on published average relationships for sands and for rock the relationship between shear modulus and shear strain for the soil profile curves shown in Figure 3A-5 was obtained as shown in Figure 3A-6.
O 3A.2-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.2 Site Conditions (Continued)
(3
%/
The predominant motion for cases involving vertical components of excitation consists of vertically propagating compression uaves.
The constrained modulus, M,
is defined using the expression 2
M= pv (3A-4)
P where 3
p is the mass density in slugs /ft v
is the compressional wave velocity in fps P
For the cases involving vertical excitation, the compressional wave velocity was determined utilizing the shear wave velocity v s
and the following expression O) 2 (1 - Ii)
(3A-5) t v
=v V
p s
1 - 2u where p is Poisson's ratio selected as follows.
The constrained modulus was assigned as a constant (strain-independent) modulus.
The assignment of constant moduli is appropriate due to the small strains developed during vertical excitation.
The damping ratio for soils and rocks is train dependent.
Based
,6 on published average relationships for sands and for rocks the damping relationship between damping and shear strain used in this study (Figure 3A-6) was obtained.
This curve was selected as a reasonably conservative assessment of the damping character-(~')
istics for the soil types that may be expected.
\\v..)
3A.2-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.2 Site Conditions (Continued)
O Reasonable variations of total unit weight and Poisson's ratio have relatively small effects on the computed response of the foundation profiles and are taken into account by variations in the shear moduli.
Therefore, for purposes of this generic study, a constant average value of total unit weight and Poisson's ratio were assigned to all soil profiles as follows.
Depth (ft)
Total Unit Weight (pcf)
Poisson's Ratio (p)
O to 30 120 0.38 30 to 100 125 0.35 100 to 200 130 0.32 Two profile depths were selected for the SSI analyses, representa-tive of shallow (75 feet) and moderately deep (150 feet) soil profiles.
The previous studies showed that, in general, the shallower soil profiles produced higher structural response.
Furthermore, SSI effects have been found to be diminishing at depths greater than approximately 200 to 250 feet.
Therefore, cases with a profile depth greater than 150 feet were not con-sidered in this study.
Consideration of a total thickness signi-ficantly less than 75 feet would result in the reactor foundation i
being slightly above or supported directly on rock.
The develop-j ment of design parameters in this case would be most conveniently done by analyzing the structure on a rigid base.
The water level for each soil profile was assumed at a depth of i
40 feet.
The effect of variations in the water level on response (assuming no soil failure) was accounted for by a variation in the modulus values.
1 9
3A.2-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.3 Cases Studied s
3A.2.3.1 Reactor Building A total number of 12 cases were studied; the cases are summarized in Table 3A-1 and are also shown in Figure 3A-5.
Cases 1 through 6 were the shallow (75 feet) soil cases with horizontal earthquake (H2) ppli d to the zero degree building cross-section (Figure 3A-7).
Case 7 was the fixed base case corresponding to a hard rock site with H applied in the zero degree direction 2
(Figure 3A-7).
Case 8 was the shallow soil case (75 feet) with horizontal earthquake (H ) applied to the 90-degree building y
cross-section (Figure 3A-7).
Case 9 was the moderately deep (150 feet) soil case with H applied to the zero-degree building 2
cross section.
Cases 8 and 9 were the same as thoue selected for 3
the 90-degree direction and 150 feet soil case.
Case 10 was the fixed base flooded containment case with H applied in the zero-2 degree direction.
Case 11 was the case with the vertical com-
-s A,/
ponent of earthquake (V) applied to the shallow soil site Case m
12 was the fixed-base case corresponding to a hard-rock site with (V) applied.
Seven of these twelve cases were the same as in Reference 3 which contained twenty cases.
This selection was based on the review of results in Reference 3 which showed that (1) shallow soil layer cases (75 feet) in general controlled the maximum responses, (2) upper bound soil profile cases generally controlled the maxi-mum responses and (3) the lower and average bound profiles con-trolled a few of the responses.
Further, additional cases ith soil profiles VP3 and VP5 (not considered in Reference 3) were included in this evaluation so that SSI cases in the gap between UB and HR profiles (see Figure 3A-5) can be considered and thus cover a wide spectrum of soil properties for the SSI analysis, m
3A.2-7
GESSAR II 22R7007 238 NUCLEAR ISLAND Rev. 0 3A.2.3.1 Reactor Building (Continued)
These generic cases coupled with the SSI analyses cases (Reference
- 3) cover a broad spectrum of sites and, therefore, are representa-tive for evaluating the seismic adequacy of the Nuclear Island.
3A.2.3.2 Other Nuclear Island Buildings The other Nuclear Island buildings include the Fuel Building, Auxiliary Building, Control Building, Diesel Generator Buildings, and Radwaste Building.
The plan view of the typical building arrangement is shown in Figure 3A-8.
To establish the Nuclear Island all-soil envelope design, a suffi-cient number of analysis cases were studied to examine primary fac-tors which have significant effects on dynamic response of the buildings.
These factors include soil shear, wave velocity pro-file, depth of the soil, and structure-to-structure interaction.
Soil cases were selected based on these factors in addition to an evaluation of the results of the Reactor / Fuel / Auxiliary Buildings analyses and the Nuclear Island Buildings dynamic characteristics.
Considerations of these factors in the selection of analysis cases are described as follows.
One of the important factors affecting seismic response of a soil-structure system is the soil shear wave velocity profile.
Ranges of the site conditions are presented in Figure 3A-5.
The results of Reactor / Fuel / Auxiliary Buildings analyses (0 direction) indi-cated that stiffer soil-profile cases resule in higher responses for the Auxiliary and Fuel Buildings in the frequency range of importance.
In addition, it was observed that the natural fre-quencies of all these buildings did not deviate far apart.
Thus, the basic analysis cases with soil profiles VP3' 5
(Figure 3A-5), and fixed base were selected for all the other Nuclear Island buildings.
3A.2-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.2.3.2 Other Nuclear Island Buildings (Continued)
\\
1 V
For the selection of the soil depth, the results of Reactor / Fuel /
Auxiliary Buildings analyses showed that shallower soil-profile cases result in higher responses in the frequency range of importance.
Thus, the basic soil cases for all Nuclear Island buildings were based on 75-feet profiles.
In order to properly account for structure-to-structure inter-action, besides the 0*-directional cross-section of Reactor / Fuel /
Auxiliary Buildings, a total of six additional soil-structure cross sections were identified (Sections A-A through F-F, Figure 3A-8).
Of these, sections 0*,
A-A, C-C, and E-E were selected as basic sections for BOP building analyses.
Furthermore, based on the result from Reactor Building vertical carthquake analysis, the fixed-base case which dominated the responses was selected as the basic analysis case for Nuclear
~~
(,)
Island buildings subject to the vertical component of earthquakes.
(O 3A.2-9/3A.2-10
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 r'] 3A.3 INPUT MOTION AND DAMPING VALUES
(
N_/
3A.3.1 Input Motion The time-history method was used in performing the seismic analysis of the foundation / structures / reactor complex by using the finite-element method.
Earthquake motion in the form of acceleration time histories for all three components was used.
Since the input carthquake motion specified in Regulatory Guide 1.60 is in the form of response spectra, artificial earthquake acceleration time histories must be created based on the given spectra.
Numerous schemes have been proposed for synthesizing an earthquake 7-12 time history from a given spectrum However, the method pro-posed by Vanmarcke and Cornell, was adopted here due to its intrinsic capability of imposing statistical independence among
/"'S the synthesized time-history acceleration components.
This is i
e
\\/ consistent with that given in Reference 3.
Three earthquake acceleration time-history components were created by using this method and identified as H H
and V.
These are y,
2, the same time histories used in Reference 3.
H and H were the 1
2 two horizontal components mutually perpendicular to each other and were based on the US NRC horizontal ground spectra in Figure 3.7-1.
H and H re shown in Figures 3A-9 and 3A-10, respectively.
V y
2 was the vertical earthquake time-history component and was based on the vertical ground spectra in Figure 3.7-2.
It is shown in Figure 3A-ll.
Figures 3.7-1 and 3.7-2 are for maximum ground accelerations of 0.3g in the horizontal direction for the SSE condition.
Two conditions are to be met for the synthesized acceleration time histories.
First, the response spectra generated from the time fg history must be a reasonable fit to the appropriate response t
i qj 3A.3-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.3.1 Input Motion (Continued) spectra.
Secondly, the three components shall be sufficiently statistically independent.
In order to verify the first condition, response spectra with damping values of 1%, 2%, 3%, 4%, 7%, and 10% for all three com-ponents of the earthquake time history were generated.
These response spectra, shown in Figures 3.7-4 through 3.7-21, are compared with the required spectra in Regulatory Guide 1.60.
The peak acceleration of the time histories of these comparison curves was normalized to a horizontal acceleration of 0.15g OBE.
The closeness of the two spectra in all cases indicates that the synthesized earthquake time histories are acceptable.
To check the statistical independence of the three earthquake 13 components, the coherence function was used.
The coherence functions for the three earthquake acceleration and V were generated.
The coher-time-history components Hy,H2, l
I ence function for H and H2, Hy 2
and V, and H and V are given in y
Figures 3.1, 3.2, and 3.3, respectively, in Appendix 3A of Refer-ence 3.
All values within the frequency range of interest between 0 and 50 Hz were calculated at a frequency increment of 0.1 Hz.
The small values of these coherence functions indi: ate that the three components are sufficiently statistically independent.
Figures 3A-12 through 3A-14 show the response spectra for the three earthquake time-history components at 2% damping for 0.15 OBE and the associated NRC Regulatory Guide 1.60 spectra.
3A.3.2 Damping The damping values for each component in the Reactor Vessel Building structure system used in this appendix are in accordance with those specified in Regulatory Guide 1.61.
These values for both the OBE and the SSE conditions are summarized in Table 3.7-1.
3A.3-2
- ~ ^ ~ ^ ^ ~ ^ ' ~ ^ ~
^ ~ ~ ~ ~ ~
" ~
^
i i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 s
i i
3A.3.2 Damping O
Damping in soil has already been described in Subsection 3A.2.2.
f In the soil-structure interaction analysis, an iterative scheme,
)
was used to assure that the final soil damping values and soil i
1 strain were compatible.
Details of this procedure are discussed l
in Section 3A.S.
I l
i i
l 1
f I
i S
I i e 1
{
i 1
[
i t
I l
i P
P 4
E 3A.3-3/3A.3-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O 3A.4 TIIE STRUCTURAL MODEL AND MODAL EQUIVALENCE p
3A.4.1 Structural Modelling Theory The Nuclear Island and the Turbine Building structures must be included in the analysis of the Nuclear Island.
Each structure must be adequately represented so that its response and the effect
~
of its presence on the response of other structures are realis-tically predicted.
Structural responses in turn provide the bases for determining the seismic loads of all the Seismic Category I components of the Nuclear Island.
To find the response of each component of any given structure, however, requires a detailed, and, therefore, extensive, seismic model of the structure.
The combination of the detailed soil and structural models would create a total model which size would be very large from both the analysis cost and computer g=pability points of view.
- However, any simplification of the overall model must contain an adequate structural representation of each important structure to provide
, -~g
(_,)
accurate dynamic behavior due to interaction effects.
The Reactor Building, Fuel Building, Turbine Building, and Auxiliary Building are on separate and rigid basemats.
In addi-tion, there are no structural ties between buildings.
Thus, each basemat became the only interface between a structure and the rest of the Nuclear Island.
There is no need to place a complex model on the basemat if a simplified model will make the basemat respond in the same manner.
The motions of the basemat may then be used as the forcing functions for the more detailed model in a separate analysis.
The Fuel and Auxiliary Building models were It simple enough so that a simplified model was not necessary.
was only necessary to reduce the three-dimensional model to a two-dimensional model.
The Reactor Building, however, was much more complex, particularly in the modelling of the reactor and inter-nals with interconnecting systems and hydrodynamic mass considera-tions.
IIere the structure was replaced by a modal-equivalent g-s b
3A.4-1
GESSAR II 22A7007 238 flUCLEAR ISLAND Rev. 0 3A.4.1 Structural Modelling Theory (Continued) model which is a series of single-degree-of-freedom systems which represent the principal modes of the structure.
Similarly, the Turbine Building was also replaced by a modal-equivalent model.
Derivation and verification of the modal-equivalent model are presented in Reference 3.
The dynamic structural models were then represented as a series of finite elements to be compatible with the finite-element soil model (Section 3A.5).
The resultant equivalent finite-element structural model was again verified to ensure that it has essen-tially the same fixed-base characteristics (frequency, d amping,
modal participation factors) as the original complex structural model with a fixed base.
3A.4.2 Finite-Element Model 3A.4.2.1 Reactor Building flodel Two-dimensional plane-strain analyses were conducted in this study.
All cases analyzed except Case 8 Table 3A-1) are in the 0 direc-tion (Figure 3A-7).
The mass and stiffness properties of the Reactor, Fuel, and Auxiliary Buildings were averaged over the 170-feet width of the buildings and that of Turbine Building over 158 feet.
Plane-strain finite-elemrnt models were obtained by using a unit thickness of the building models.
For the case in the 90 direction, only the shaded area of the Fuel and Auxiliary Buildings (Figure 3A-7) was considered effective in the structure-to-structure interaction.
The mass of the shaded area was averaged over a length of 138 feet for the Reactor Building atd a length of 69 feet for the Auxiliary and Fuel Buildings.
The st iffness of the Fuel and Auxiliary Buildings were determined so tnat their fundamental natural frequency was maintained.
For the Reactor Building, all modes up to 35 Hz were included.
The 3D effects 3A.4-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.4.2.1 Reactor Building Model (Continued)
J of the soil was simulated by adding a viscous boundary to the out-of-plane direction.
In addition, transmitting boundaries were provided on the two lateral sides of the soil layers to absorb any wave effects emanating from the structure.
The soil backfill is modeled using finite elements with common boundaries for the backfill elements and structural elements.
This is consistent with the degree of sophistication available in FLUSH because it is not feasible currently to model potential separation between the backfill soil and the structure.
The final equivalent plane-j strain models for the Reactor, Auxiliary, and Fuel Buildings in the two directions are shown in Figures 3A-15 and 3A-16.
The dynamic 1
response of the soil-structure system was evaluated from the analyses of these plane-strain models.
The Reactor Building was represented by its first 20 fixed-base nornal modes and the Turbine Building by its first 5 fixed-base s_)
normal modes.
For each normal mode, an equivalent planc-strain finite element was compute 3 so that the finite-element model has the same inertial properties, fundamental natural frequency, and modal height as the modal model.
The matching of the natural frequencies can be easily achieved by scaling up or down the clastic modulus of the finite element since the natural frequen-cies of a finite element are directly proportional to the square root of its clastic modulus.
4 Due to the relative simplicity of the Fuel and Auxiliary Buildings, the buildings were directly reduced to planc-strain finite-element models with the same overall dimensions and mass distribution.
The clastic modulus of the finite element was scaled so that its fundamental natural frequency matches that of the original building model.
O a
3A.4-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O l
3A.4.2.2 Other Nuclear Island Building SSI Models i
In order to properly account for the structure-to-structure interaction effect, varous cross sections were considered for the modeling of the Nuclear Island Building SSI finite-element analysis (Figure 3A-8).
Figure 3A-17 presents a typical model for section A-A which includes Auxiliary, Control, and Diesel Genera-tor Buildings.
All the modeling techniques used were the same as described for the Reactor Building analysis except that stick models were used in place of modal equivalent models for the buildings.
Stiff beams were used between the building stick and its foundation in order to properly transmit the responses from the basemat to the superstructures.
O O
3A.4-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/
3A.5 SOIL-STRUCTURE INTERACTION ANALYSIS V) 3A.5.1 Introduction Nuclear Island structures are massive structures typically embedded at a considerable depth in a soil deposit.
An important aspect in the seismic design of these structures is the evaluation of the dynamic interaction between the structure and the soil.
Such interaction effects may significantly affect the response of the structures as well as the equipment systems.
The problem of soil-structure interaction has been the subject of considerable research in recent years and a number of interaction models have been developed and are in use in the industry.
Basi-cally, however, there are two general approaches available at present to estimate dynamic soil-structure interaction effects.14
(~'%
(1)
A direct approach (tocal solution) where the system
'\\ ')
(soil and one or more structures) is modeled and analyzed either in the time domain or in the frequency domain with a mathematical model of the soil and the structures most commonly based on finite elements.
Finite difference schemes may also be used to model the soil, although this procedure is less frequently used in practice.
(2)
A three-step solution is used where:
(a) the motion that would occur at the level of the foundation is first determined; (b) the frequency-dependent compliances or stiffnesses of the foundation (or foundations if there are several adjacent buildings) are computed; and bo 3A.5-1 t
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.5.1 Introduction (Continued)
(c) a dynamic analysis of the structure is performed (either in the time domain or in the frequency domain) with the stiffness functions computed in (b) reproducing the soil and the motion derived in (a) as input.
Each method has its own advantages and limitations from a practical standpoint, and significantly different results can be obtained from the two modeling techniques depending on the properties assigned to the elements of the model, the assumptions, and the solution algorithms used.
A thorough comparative study of the l5 merits of the two modeling techniques was made by Seed with regard to the evaluation of the soil-structure interaction effects for embedded structures.
In this study, the authors concluded that:
O "The finite element method, properly performed with due regard to the extent and fineness of the mesh and varia-tions of damping characteristics, can provide a good evaluation of the response of embedded structures and is the best tool currently available for this purpose."
In this appendix, evaluation of the dynamic interaction between the plant structures and the surrounding and unJ'rlying soil was made by modeling the soil-structure system by finite elements.
This is consistent with the method used in Reference 3.
The general method of app;oach is shown in Figure 3A-18 and involves the following main steps:
(1) because the control motion is typically specified at some point in the free field; as a first step, a deconvolution analysis is necessary to determine the compatible 3A.5-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.S.1 Introduction (Continued)
(N;
/
base-rock motions (motions which would have to develop in an underlying rock-like formation to produce the specified motions at the control point); and (2) the same base-rock motion is then used as excitation for a two-dimensional analysis (plane strain) of the soil-structure system leading to an evaluation of the motions at any selected points such as the base of the struc-ture, the operating floor of the structure, etc (accuracy of the results can be checked by inspecting the degree of agreement between the computer motions at the control point with the specified values).
In the studies reported herein, computations were made using the recently developed program which permit the use of variable modu-(^N lus and variable damping in the soil.
The deconvolution analyses I
\\
required in step (1) and the two-dimensional analyses in step (2) were carried out using the finite-element computer program FLUSH.
The nonlinear soil properties were accounted for in this program by a combination of the equivalent linear method described in Reference 17 and the method of complex response with complex moduli.18 The latter method makes it possible to use different damping properties for each element of the finite element model.
In addition, FLUSH has the capability to simulate 3D effects by adding a viscous boundary to the out-of-plane direction.
It can also provide a transmitting boundary near the structural interface to absorb any wave effects emanating from the structure and thus 4
simulate the effects of an extensive soil deposit.
The high frequency ranges, which must be considered in the study of soil-structure interaction for nuclear power plants, were also adequately accounted for in this method.
O
\\
i x_-
3A.5-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.5.2 Analysis Procedure The dynamic interaction between the plant structures and the surrounding and underlying soil was evaluated by modeling the soil-structure system by finite elements.
The steps involved using the finite-element method are as follows.
(1)
Perform a deconvolution analysis to determine compatible base rock motions.
This is the motion which would have to be developed in an underlying bedrock to produce the specified control motion at the finished grade in the free field.
This is essentially a one-dimensional analysis using the principle of propagation of shear waves in the vertical direction.
The vertical height of soil elements was chosen so that it is small compared to the wave length of shear waves propagating through the soil.
The vertical element size was chosen as Vs hs3-A
= jk (3A-6) 5 s
5f max where A
is the wave length of the shortest wave s
V is the shear wave velocity in the element s
f is the highest frequency of interest.16 The computations required in this step were performed using the computer program FLUSH.
O 3A.5-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
()
3A.S.2 Analysis Procedure (Continued)
'N j For vertical components of earthquakes, it was assumed that the propagating seismic waves were P-waves.
However, if the shear modulus C is replaced by an equivalent modulus G* as G* = 2 (1 - p)
G 1 - 2p (3A-7) where p is the Poisson's ratio, the solution for the shear wave propagation will be identical to that for the P-wave propagation.
In reality, the vertical and horizontal components of earthquakes must be considered to act simultaneously.
Hence, neglecting, the effect of the P-wave on the shear wave propac, tion, G* was calculated using the final G from the solution for
,]
horizontal earthquake component.
The vertical com-
'\\ j ponents of base rock motion was then generated through the deconvolution analysis using G* without iterations.
(2)
Develop a full finite element mesh to represent the structures and the soil surrounding and underlying the structures and evaluate the response of this soil-structure system to the bedrock motion developed in step (1).
This analysis was performed using FLUSH.
In this analysis, three-dimensional effects were approximated by using an energy-absorbing viscous boundary, and lateral extent of the two-dimensional model was reduced by providing transmitting boundaries to absorb any wave effects emanating from the structure.
These options are built into FLUSH.
/
\\
Y i
3A.5-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.S.2 Analysis Procedure (Continued)
The structures included in the analysis are the Reactor Building, Fuel Building, Auxiliary Building, and the Turbine Building.
These buildings are supported on independent basemats with no connection between buildings and hence each basemat is the only interface between each structure and the rest of the system.
The struc-tures were idealized as two-dimensional plane-strain structures using finite elements.
Each of the structures was idealized using the modal equivalent approach to determine their modal masses, damping, and equivalent shear modulus values.
In addition, equivalent modal heights were also calculated for the Reactor Buildings.
Modes with frequencies up to 35 Hz for the horizontal and vertical analyses for the Reactor Building and 45 Hz for the horizontal model of the Turbine Building were included to determine the modal models.
The modal G values for the Fuel and Auxiliary Buildings were based on their first mode and their moaal masses and heights were maintained to equal the actual building models.
For the 90-degree case, only the shaded area of the Fuel and Auxiliary Buildings (Figure 3A-7) was consid-cred effective in the soil-structure interaction.
Typical finite-element models used in this study are shown in Figures 3A-19, 3A-20, and 3A-21.
The embedment depth (bottom of foundation mat) was maintained at 39 feet for the Reactor Building, 29 feet for the Fuel Building, 35 feet for the Auxiliary Building, and 23 feet for the Turbine Building.
O 3A.5-6
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
/}
3A.S.2 Analysis Procedure (Continued)
N,._.J The soil profile and properties for the soil layers for all the cases were the same as in Section 3A.2.
The other properties used in the analysis are:
unit weight of concrete (150 pcf), Poisson's Ratio for concrete (0.20), and damping ratio for concrete (0.04 for OBE and 0.07 for SSE).
Minor variations in these properties will not affect the analysis results.
An important decision was the choice of the maximum frequency to be included in the analyses.
The execution time (using FLUSH) is proportional to the highest fre-quency considered during the iteration; furthermore, smaller elements are required for high frequency analy-ses.
Therefore, it is very important not to consider frequencies higher than absolutely necessary.
Typically
/N maximum frequency values of 20 to 25 Hz are used in the soil-structure analyses involving nuclear power plants.
It is, however, generally believed that most interaction effects between the structure and the underlying soil would involve frequencies well below 20 Hz.
In select-ing the value of maximum frequency, structural design considerations and the maximum frequency that can be transmitted through a soil layer for a given accelera-tion level must also be taken into account.
Based on these considerations, for the horizontal cases, responses up to a frequency of 20 Hz for softer soil properties and 33 Hz for the stiffer soil properties and for the vertical case were included in the analysis.
In this study, the control motion was applied in the free field at grade and deconvoluted to the bottom of the soil layer.
This results in a gradual and reasonable
(~N gradient of maximum accelerations along the depth of the 3A.5-7
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.5.2 Analysis Procedure (Continued) g soil profile with maximum accelerations generally decreasing with depth in all the cases analyzed.
This is consistent with conclusions of similar analy-19,20,21 ses.
In addition, the vibratory motions calcu-lated in the free field at the elevation of the bottom of the Reactor Building basemat gave response spectra at all frequencies (0.2 to 50 Hz) not less than 60 per-cent of the design spectra (Figure 3A-22).
It can be noted from this figure that the envelope of the spec-trum for all the cases considered exceeds 60 percent of the NRC spectrum at all the frequencies.
The control motions used have a duration of 22 seconds with a time step of 0.005 second.
The Fast Fourier Transform (FFT) algorithm used in FLUSH requires that the number of terms in the series be some power of 2.
Hence, sufficient trailing zeroes were added to the transient time histories so that the periodic nature of FFT did not induce errcrs in the response.
In the analyses, 8192 terms were used for the control motions with values between 22.0 seconds and 40.96 seconds set equal to zero.
Another important item to consider in saving execution time was interpolation on the transfer function in the frequency domain.
Results were used to establish the interpolation control number used in this study.
The strain-dependent soil material properties used in this appendix are shown in Figure 3A-6 and given in Tables 3A-2 and 3A-3.
The nonlinear behavior of soil was accounted for in the analysis by an equivalent 5
linear method using strain-dependent soil properties.
g 3A.5-8
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
(; 3A.5.2 Analysis Procedure (Continued)
\\,i This approach is iterative and requires about three to five iterations to obtain strain-compatible solutions.
Within three to four iterations, the soil properties were found to converge to a strain-compatible solution.
Analysis to obtain vertical components of deconvoluted bedrock motion was carried out without iterations using the final soil properties from the analysis of the corresponding horizontal component case (e.g.,
GE-75-A-H2 for GE-75-A-V) for both free-field and finite-element analyses.
Response spectra plots and acceleration time-history responses at various node points on the basemat of the structures were generated from the finite-element
/N analysis.
In addition, response spectra were also
\\
'l generated in the free field at top and bottom of soil layers and at elevation of the Reactor Building basemat.
Response spectra for all the cases, generated at the center of top of basemat, are shown in Figures 3A-23 and 3A-24.
These response spectra plots were used to The compare the response with various other cases.
spectra in Figures 3A-23 and 3A-24 are well within the basemat envelope spectra.3 Both translational and rotational time histories of acceleration response generated at the top of basemat of structures were used to analyze a more detailed model of the structures to evaluate their response and generate response spectra at critical elevations for analysis of equipment.
~>
3A.5-9/3A-5.10
4 GESSAR II 22A7007
- 238 NUCLEAR ISLAND Rev. 0 I
3A.6 REFERENCES s
1.
Minutes of Meeting on GESSAR (BWR/6-Mark III) Seismic Design Envelope Generation Program, USAEC Structural Engineering Branch, March 4, 1974.
2.
Minutes of Meeting on Seismic Design Requirements of GESSAR Application, USAEC Structural Engineering Branch, December 18, 1973.
3.
General Electric Company BWR/6-238 Standard Safety Analysis Report (GESSAR), Docket No. STN 50-447, November 7, 1975.
4.
U.S. Atomic Energy Commission Regulatory Guide 1.60, December 1973.
5.
H.
B.
Seed and I. H.
Idriss, Soil Moduli and Damping-Factors 4
for Dynamic Response Analyses, Earthquake Engineering Research Center Report No. EERC 70-10, University of California, i
Berkeley, December 1970.
6.
P.
B.
Schnabel, Effects of Local Geology and Distances from Sonyca on Earthquake Ground Motions, Ph.D. thesis, University of California, Berkeley, 1973.
f
/'*\\
( )
7.
E.
H. Vanmarcke and C. A. Cornell, Seismic Risk and Design j
Response Spectra, ASCE Specialty Conference on Safety and Reliability of Metal Structures, Pittsburgh, Pennsylvania, November 1972.
f i
1 8.
N.
C. Tsai, Spectrum-Compatible Motions for Design Purposes, j
Journal of the Engineering Mechanics Division, ASCE, 7
i Vol. 98, No. EM.
2, Proc. Paper 8807, April 1972, pp. 345-356.
I 9.
P.
C.
- Rizzo, D. E.
Shaw, and S.
J.
Jarecki, " Development of i
Real/ Synthetic Time Histories to Mat-h Smooth Design Spectra, Proc., Second International Conference on Structural Mechanics in Reactor Technology, American Nuclear Society, Berlin, Germany, September 1973.
10.
M. Shinozuka, Digital Simulation of Ground Accelerations, I
Paper 360, Fifth World Conference on Earthquake Engineering, Rome, Italy, June 1973.
t 11.
S.
N.
Hou, Earthquake Simulation Models and Their Applica-i tions, Massachusetts Institute of Technology Civil Engineering i
Report R68-17, Cambridge, Mass., May 1968.
i m
3A.6-1 l
\\
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 3A.6 REFERENCES (Continued) 12.
R.
H.
Scanlan and K.
Sacks, Earthquake Time Histories and Response Spectra, Journal of the Engineering Mechanics Division, ASCE, Vol. 100, No. EM 4, Proc. Paper 10703, August 1974.
13.
J.
S.
Bendat and A.
G. Piersol, Random Data:
Analysis and Measurement Procedures, Wiley - Interscience, 1971.
14.
E.
Kausel and J.
M.
Roesset, Soil Structure Interaction Problems for Nuclear Containment Structures, paper presented at the ASCE Power Division Specialty Conference, Boulder, Colorado, August 1974.
15.
H.
B.
- Seed, J.
Lysmer, and R.
Hwang, Soil-Structure Inter-action Analyses for Evaluating Seismic Response, Report No. EERC 74-6, Earthquake Engineering Research Center, University of Calibornia, Berkeley, April 1974.
16.
J.
- Lysmer, T.
- Udaka, C.
F.
Tsai, and H.
B.
Seed, FLUSH -
A Computer Program for Simplified 3-D Analysis of Soil-Structure Interaction Problems, Report No. EERC 75-30, University of California, Berkeley, November 1975.
17.
I.
M.
- Idriss, H.
Dezfulian, and H.
B.
Seed, Computer Programs for Evaluating the Seismic Response of Soil Deposits with Non-Linear Characteristics Using Equivalent Linear Procedures, Research Report, Geotechnical Engineering, University of California, Berkeley, 1969.
18.
J.
- Lysmer, T.
- Udaka, H.
B.
Seed, and R.
Hwang, LUSH - A Computer Program for Complex Response Analysis of Soil-Structure Systems, Report No. EERC 74-4, Earthquake Engineer-ing Research Center, University of California, Berkeley, April 1974.
l 19.
I.
M.
Idriss and R. Akky, Generation of Earthquake Motions in a Soil Deposit by a Deconvolution Process, paper presented at the ASCE Power Division Specialty Conference, Boulder, Colorado, August 1974.
20.
Analysis for Soil-Structure Interaction Effects for Nuclear Power Plants, report by the ASCE, November 1978.
21.
Seismic Input and Soil-Structure Interaction, prepared for U.
S. NRC by D'Appolonia Consulting Engineers, February 1979 (NUREG/CR-06 93 ).
O 3A.6-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O f-Table 3A-1 CASES CONSIJERED FOR REACTOR BUILDING Case No.
Case Direction Description 1
GE-75-L-H 75 ft soil with lower bound 2
properties
- in 0* direction 2
GE-75-A-II 75 ft soil with average properties
- 2 in 0* direction 3
GE-75-U-H 75 ft soil with upper bound 2
properties
- in 0 direction 4
GE-75-VP -U 75 ft soil with VP P# fil I"
3 2
3 0* direction 5
GE-75-VP -U 75 ft soil with VP5 profile in 5
2 0
direction 6
GE-7 5 -ilR-li 75 ft soil with uniform rock 2
3500 fps in profile with V
=
3 0* direction 7
GE-FB-H Fixed base case in Ol* direction
(-]
2 V
8 GE-75-U-H 75 ft soil with upper bound y
properties
- in 90 direction 9
GE-150-U-il 150 ft soil with upper bound 2
properties
- in 0 direction 10 GE-FFB-II Flooded fixed base in 0 direction 3
11 GE-75-A-V 75 ft soil with average properties
- in vertical direction 12 GE-FB-V Fixed base case in vertical direction
/
g
- See Reference 3 3A.7-1
i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O l
Table 3A-2 STRAIN-COMPATIBLE SOIL PROPERTIES (VP VP
"' "' ^'
0' 3,
5' i
Effective Shear Strain, Praction of Critical (1) 109 (Y Y ggg eff)
Damping (%)
~4*
il x 10
-4.0 0.60
-4 3.16 x 10
-3.5 0.75
-3 1.00 x 10
-3.0 1.30 1
_2 3.16 x 10
-2.5 2.59
-2 1.00 x 10
-2.0 4.70
-2 3.16 x 10
-1.5 8.40
-1
)
1.00 x 10
-1.0 13.50 j
0.316
-0.5 18.64
)
1.00 0.0 22.60 3.16 0.5 23.80 i
- 10.00 1.0 25.00 O
- This factor has to be applied to the shear modulus at low shear
-4 i
strain amplitudes (here defined as 1 x 10
%) obtain the modulus at higher strain levels.
i
{
i i
3A.7-2 l
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Table 3A-3 9
STRAIN-COMPATIBLE SOIL PROPERTIES (VP3' 5, VPliR, LB, AV, UB)
Effective Shear Strain, Shear Modulus Reduction Factor
'Icff I9 IY I
(VP3, VP5, VPIIR)
I'I eff (LD,_AV, U B_1
~4*
<1 x 10
-4.0 1.000 1.000
~4 3 16 x 10
-3.5 0.990 0.990
~
1.00 x 10
-3.0 0.940 0.960
-3 3.16 x 10
-2.5 0.860 0.905
-2 1.00 x 10
-2.0 0.740 0.813
-2 3.16 x 10
-1.5 0.520 0.660
~1 1.00 x 10
-1.0 0.300 0.462 0.316
-0.5 0.135 0.303 1.00 0.0 0.060 0.195 3.16 0.5 0.060 0.127
>10.00 1.0 0.060 0.080 0
- This factor has to be applied to the shear modulus at low shear strain amplitudes (here defined as 1 x 10~ %) to obtain the modulus at higher strain levels.
O 3A.7-3/3A.7-4
l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
STP h l
l SOUTH TE XAS PROJECT "9
SAN JOAOUIN PROJECT s0 r
I i
h SJP l
iOO l
]
E iso l
0 l
s 200 I
-1 I
250 l
l l
l 1
0 1000 2000 3000 4000 5000 SHE AR WAVE VELOCITY (fps) l l
{
l Figure 3A-1.
Variation of Shear Wave Velocity with Depth at South Texas and San Joaquin Nuclear Power Plant Sites 3A.8-1 j
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
O N
I SO h k _,
$ 50 50 1
I I
100 I
_l
=t I 150 D
\\
O FFTF E_
200 5
l 250 FFTF PROJECT f
== = = SAN ONOF RE - FIE LD D AT A
-.== SAN ONOFRE - LABOR ATORY DATA I
I I
300 o
1000 2000 3000 4000 5000 SHE AR WAVE VELOCITY (fps)
(
Figure 3A-2.
Variation of Shear Wave Velocity with Depth at l
San Onofre and FFTF Nuclear Power Plant Site 3A.8-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev, 0 O
O W/f/MM/A 50 H
I 100 p.-q g
o
$ 150 O
a 8
W-4 N
200 k--4 W--H H
250 -
SATSOP (WPPSS) NUCLEAR PROJECT H
1-=H ME NDOCINO PROJECT h
DIA BLO C AN YON l
l I
I 300 O
1000 2000 3000 4000 5000 SHE AR WAVE VELOCITY (fps)
Fiqure 3A-3.
Variation of Shear Wave Velocity with Depth at O
Three Selected Nuclear Power Plant Sites 3A.8-3
4000
.. 4
... e,
......... -.. -.,.... +
,... +... +.........,...........
...,4-.
.. +........... -................. -.. -.
_..,.4 4.....
COMPETE NT ROCK (ASSUMED RIGID FOR AN ALYSES)..... -.... _ -... _..... _... ~. ~..... ~ *..... - - - -
- '*****^^^"****
4.....
..m,..._..._.,............_.....-..-....-...-....... -..............
......-......u.....4..
.......- -...+,........ i
............... -.. -... -...,............ +..... _.
i.....;..
4..
....,4..
...+..-.....y
....a..-....-..............-.
i SOFT ROCK AND CEMENTED SOILS !
3000 A:::
2000 b
j M
SOIL SITES (SANDS. CLAYS. SILTS AND GR AVELLY SOILS) to m yy y
e xx O
W HH h
MH U
1000 LOOSE TO MEDIUM DENSE SAND; SOFT AND MEDIUM STIF F CLAY o
I I
l l
l 0
50 100 150 200 250 300 N
ww DEPTH (ft) ep
<: 4 Figure 3A-4.
Range of Shear Wave Velocities for Nuclear Power Plant Sites in a
o4 liigh Seismic Areas O
O O
~ - -.
j GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O
.i 1
1 0
1.
)
i i
< i7,10,12 (FIXED BASE) 50 5
< '6 1
2,11 3.8 4
100 i
E-l UB
b O
W j
2 3
4 6
300 o
1000 2000 3000
> 3500 SHEAR WAVE VE LOCITY (fps)
Figure 3A-5.
Shear Wave Velocity Profiles to be Used in SSI Analyses 3A.8-5
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 1.0 0
0.8 m
? b ROCK (PROFILE HR) y g PROFILE VP6 g o y
0.6 PROFILES g j VP. VP, VP5 3
4 g 3 a
2 O
3 2
2 0.4 E
E y
2 m
1 2
l l
l g
10'#
10 10 1
to 3
w2 3
SHE AR STR AIN,7 (%)
30 0
25 20
_d 9
7m 15 PROFILES VPg j
THROUGH VP,
2<
10 PROFILE VP6 ROCK (PROFILE HR) 0 l
I i
10'd 10'3 10 0'I 1
10
-2 SHE AH STRAIN,y (%)
Figure 3A-6.
Variation of Shear Modulus and Damping Ratio with Shear Strain Used in Analyses 1
3A.8-6
l Ommm>y gg wM> WOO 4
=
aeM>y gm $8
%o<*
0 O
t f
85 1
']
G N
n I
a D
t L
f I
l 0
U P
4 B
3 E
t N
n I
B e
R m
U e
T 0
0 gn arrA g
O l
n i
d
}
iuB Y
7 RG AN A
ILD 3
I I L t
X I e
f i
d1I UU
\\
4 AB r
2 u
1 R
g i
O h
F TC A
E R
G t
N f
I 8
D 3
L 1
I U
G B
t f
\\
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
%j-SYMMETRY AXIS I
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(]
SUMMARY
TABLE OF CONTENTS G'
Chapter /
Section Title Volume 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
1 1.1.1 Type of License Required 1.1.2 Identification of Applicant 1.1.3 Number of Plant Units 1.1.4 Description of Location 1.1.5 Type of Nuclear Steam Supply System 1.1.6 Type of Containment 1.1.7 Core Thermal Power Levels 1.1.8 Scheduled Completion and Operation Dates 1.2 GENERAL PLANT DESCRIPTION 1
1.2.1 Principal Design Criteria 1.2.2 Plant Description Os 1.3 COMPARISON TABLES 1
1.3.1 Comparisons with Similar Facility Designs 1.3.2 Comparisons of Final and Preliminary Information 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1
1.4.1 Applicant 1.4.2 Architect Engineer - Nuclear Island Design 1.4.3 Nuclear Steam Supply System Supplier 1.4.4 Turbine-Generator Vendor 1.4.5 Consultants 1.5 REQUIREMENTS FOR FURTHER TECHNICAL 7"5' (( " ION 1
1.5.1 Current Development Prog y 2 1.5.2 PSAR Commitment Items 1.6 MATERIAL INCORPORATED BY REFERENCE 1
iii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1
1.7.1 Electrical, In'3 trumentation, and Control Drawirgs 1.7.2 Piping and Ir.strumentation Diagrams 1.7.3 Abbreviations and Symbols 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1
1.8.1 Compliance Assessment Method 1.9 STANDARD DESIGNS 1
1.9.1 Interfaces 1.9.2 Exceptions I
i O
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O R
SUMMARY
TABLE OF CONTENTS-(Continued)
Chapter /
Section Title Volume 2
SITE CHARACTERISTICS 2.0
SUMMARY
l 2.1 GEOGRAPHY AND DEMOGRAPHY l
2.1.1 Site Location and Description 2.1.2 Exclusion Area Authority and Control Population Distribution 2.1.3 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 1
2.2.1 Location and-Routes 2.2.2 Descriptions 2.2.3 Evaluation of Potential Accidents
-2. 3 METEOROLOGY l
2.3.1 Regional Climatology 2.3.2 Local Meteorology
()
2.3.3 Onsite Meteorological Measurements Program 2.3.4 Short-Term Atmospheric Diffusion Estimates 2.3.5 Long-Term Atmospheric Diffusion Estimates 2.4-HYDROLOGIC ENGINEERING 1
2.4.1 Hydrologic Description 2.4.2 Floods 2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers-2.4.4 Potential Dam Failures, Seismically Induced 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4.6 Probable Maximum Tsunami Flooding l
2.4.7 Ice Effects 2.4.8 Cooling Water Canals and Reservoirs 2.4.9 Channel Diversions
()
2.4.10 Flooding Protection Requirements v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 2.4.11 Low Water Considerations 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters 2.4.13 Ground Water 2.4.14 Technical Specifications and Emergency Operation Requirements 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING 1
2.5.1 Basic Geologic and Seismic Information 2.5.2 Vibratory Ground Motion 2.5.3 Surface Faulting 2.5.4 Stability of Subsurface Materials and Foundations 2.5.5 Stability of Slopes 2.5.6 Embankments and Dams I
l 1
)
i vi 2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
}
SUMMARY
TABLE OF CONTENTS (Continued)
%/
Chapter /
Section Title Volume 3
DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 2
3.1.1 Summary Description 3.1.2 Criterion Conformance 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 2
3.2.1 Seismic Classification 3.2.2 System Quality Group Classifications 3.2.3 System Safety Classifications 3.2.4 Quality Assurance 3.2.5 Correlation of Safety Classes with Industry Codes 3.3 WIND AND TORNADO LOADINGS 2
()
3.3.1 Wind Loadings 3.3.2 Tornado Loadings 3.3.3 BOP Interface 3.3.4 References 3.4 WATER LEVEL (FLOOD) DESIGN 2
3.4.1 Flood Protection 3.4.2 Analytical and Test Procedures 3.4.3 BOP Interface 3.4.4 References 3.5 MISSILE PROTECTION 2
3.5.1 Missile Selection and Description 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles 3.5.3 Barrier Design Procedures 3.5.4 BOP Interface 3.5.5 References vii i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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Chapter /
Section Title Volume 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 2
3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.3 References 3.7 SEISMIC DESIGN 3
3.7.1 Seismic Input 3.7.2 Seismic System Analysis 3.7.3 Seismic Subsystem Analysis 3.7.4 Seismic Instrumentation 3.7.5 BOP Interface 3.7.6 References 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 3
3.8.1 Concrete Containment 3.8.2 Steel Containment 3.8.3 Concrete and Steel Internal Structures of Steel Containment 3.8.4 Other Seismic Category I Structures 3.8.5 Foundations 3.8.6 BOP Interface 3.9 MECHANICAL SYSTEMS AND COMPONENTS 4
3.9.1 Special Topics for Mechanical Components 3.9.2 Dynamic Testing and Analysis 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures 3.9.4 Control Rod Drive System 3.9.5 neactor Pressure vessels Internals 3.9.6 Inservice Testing of Pumps and Valves 3.9.7 BOP Interface 3.9.8 References viii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
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\\
SUMMARY
TABLE OF CONTENTS (Continued)
N~
Chapter /
Section Title Volume 3.10 SEISMIC QUALIFICATIONS OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING HYDRODYNAMIC EFFORTS) 5 3.10.1 Seismic Qualification Criteria (Including Hydrodynamic Loads) 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10.3 Methods and Procedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumentation (Including Hydrodynamic Load 3) 3.10.4 Seismic Qualification Tests and Analyses (Including Hydrodynamic Loads) 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED
(
)
MECHANICAL AND ELECTRICAL EQUIPMENT 5
'w.J 3.11.1 Equipment Identification and Environmental Conditions 3.11.2 Qualification Tests and Analyses 3.11.3 Qualification Results 3.11.4 Loss of Ventilation 3.11.5 Estimated Chemical and Radiation Environment APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 5
APPENDIX 3B CONTAldMENT LOADS 6,7 APPENDIX 3C COMPUTER PROGRAMS USED IN THE DESIGN OF SEISMIC CATEGORY I STRUCTURES 8
APPENDIX 3D ANALYSIS OF RECIRCULATION MOTOR AND PUMP UNDER ACCIDENT CONDITIONS 8
APPENDIX 3E DESCRIPTION OF SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 8
APPENDIX 3F DYNAMIC BUCKLING CRITERIA FOR CONTAINMENT VESSEL 8
v) ix
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume APPENDIX 3G PIPE FAILURE ANALYSIS 8
APPENDIX 311 EFFECT OF CONCRETE ANNULUS BELOW ELEVATION (-) 5 FT.,
3 IN. ON SEISMIC DESIGN LOADS AND BUILDING RESPONSES 8
O 4
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 (3
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TABLE OF CONTENTS (Continued)
\\
)
us Chapter /
Section Title Volume 4
REACTOR 4.1
SUMMARY
DESCRIPTION 9
4.1.1 Reactor Vessel 4.1.2 Reactor Internal Components 4.1.3 Reactivity Control Systems 4.1.4 Analysis Techniques 4.1.5 References 4.2 FUEL SYSTEM DESIGN 9
4.2.1 General and Detailed Design Base 4.2.2 General Design Description 4.2.3 Design Evaluations 4.2.4 Testing and Inspection 4.2.5 Operating and Developmental Experience
/
4.2.6 References 4.3 NUCLEAR DESIGN 9
4.3.1 Design Bases 4.3.2 Description 4.3.3 Analytical Methods 4.3.4 Changes 4.3.5 References 4.4 THERMAL - HYDRAULIC DESIGN 9
4.4.1 Design Basis 4.4.2 Description of Thermal Hydraulic Design of the Reactor Core 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor Coolant System 4.4.4 Evaluation 4.4.5 Testing and Verification 4.4.6 Instrumentation Requirements 4.4.7 References
,s (v) xi
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Chapter /
Section Title Volume 4.5 REACTOR MATERIALS 9
4.5.1 Control Rod System Structural Materials 4.5.2 Reactor Internal Materials 4.5.3 Control Rod Drive liousing Supports 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 9
4.6.1 Information for Control Rod Drive System (CRDs) 4.6.2 Evaluations of the CRDs 4.6.3 Testing and Verification of the CRDs 4.6.4 Information for Combined Performance of Reactivity Systems 4.6.5 Evaluation of Combined Performance 4.6.6 References APPENDIX 4A CONTROL ROD PATTERNS AND ASSOCIATED POWER DISTRIBUTION FOR TYPICAL BWR 9
4A.1 Introduction 4A.2 Power Distribution Strategy 4A.3 Results of Core Simulation Studies 4A.4 Glossary of Terms 4A.5 References O
xii
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SUMMARY
TABLE OF CONTENTS (Continued)
]
Chapter /
Section Title Volume 5
REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1
SUMMARY
DESCRIPTION 10 5.1.1 Schematic Flow Diagram 5.1.2 ~
Piping and Instrumentation Diagram 5.1.3 Elevation Drawing 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 10 5.2.1 Compliance with Codes and Code Cases 5.2.2 Overpressure Protection 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2.5 Reactor Coolant Pressure Boundary and ECCS System Leakage Detection System
'% )
5.2.6 References 5.3 REACTOR VESSEL 10 5.3.1 Reactor Vessel Materials 5.3.2 Pressure / Temperature Limits 5.3.3 Reactor Vessel Integrity 5.3.4 References 5.4 COMPONENT AND SUBSYSTEM DESIGN 10 5.4.1 Reactor Recirculation Pumps 5.4.2 Steam Generators (PWR) 5.4.3 Reactor Coolant Piping 5.4.4 Main Steam Line Flow Restrictors 5.4.5 Main Steam Line Isolation System 5.4.6 Reactor Core Isolation Cooling System (RCIC) 5.4.7 Residual Heat Removal System 5.4.8 Rt. actor Water Cleanup System 5.4.9 Main Steam Lines and Feedwater Piping 5.4.10 Pressurizer xiii
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Chapter /
Section Title Volume 5.4.11 Pressurizer Relief Valve Discharge System 5.4.12 Valves 5.4.13 Safety and Relief Valves 5.4.14 Component Supports 5.4.15 References O
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Section Title Volume 6
ENGINEERED SAFETY FEATURES 6.0 GENERAL ll 6.1 ENGINEERED SAFETY FEATURE MATERIALS 11 6.1.1 Metallic Materials 6.1.2 Organic Materials 6.2 CONTAINMENT SYSTEMS 11 6.2.1 Containment Functional Design 6.2.2 Containment Heat Removal System 6.2.3 Secondary Containment Functional Design 6.2.4 Containment Isolation System 6.2.5 Combustible Gas Control in Containment 6.2.6 Containment Leakage Testing (A) 6.2.7 Suppression Pool Makeup System 6.2.8 i
References 6.3 EMERGENCY CORE COOLING SYSTEMS 11 6.3.1 Design Bases and Summary Description 6.3.2
System Design
6.3.3 ECCS Performance Evaluation 6.3.4 Tests and Inspections 6.3.5 Instrumentation Requirements 6.3.6 References 6.4 HABITABILITY SYSTEMS 11 6.4.1 Design Basis 6.4.2
System Design
6.4.3 Systems Operational Procedures 6.4.4 Design Evaluations 6.4.5 Testing and Inspection 6.4.6 Instrumentation Requirements 6.4.7 Nuclear Island / BOP Interface O
XV
e GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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O Chapter /
Section Title Volume 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 11 6.5.1 Standby Gas Treatment System (SGTS) 6.5.2 Containment Spray Systems 6.5.3 Fission Product Control Systems 6.5.4 Ice Condensers as a Fission Product Control System 6.5.5 References 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS ll 6.6.1 Components Subject to Examination 6.6.2 Accessibility 6.6.3 Examination Techniques and Procedures 6.6.4 Inspection Intervals 6.6.5 Examination Categories and Requirements 6.6.6 Evaluation of Examination Results 6.6.7 System Pressure Tests 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures 6.7 MAIN STEAM POSITIVE LEAKAGE CONTROL SYSTEM (MSPLCS) 11 6.7.1 Design Bases 6.7.2
System Description
6.7.3 System Evaluation 6.7.4 Inspection and Testing 6.7.5 Instrumentation Requirements 6.8 PNEUMATIC SUPPLY SYSTEM ll 6.8.1 Design Bases 6.8.2
System Description
6.8.3 System Evaluation 6.8.4 Inspection and Testing Requirements 6.8.5 Instrumentation Requirements APPENDIX 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS 11 xvi
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SUMMARY
TABLE OF CONTENTS (Continued)
~\\Y Chapter /
Section Title Volume 7
INSTRUMENTATION AND CONTROL SYSTEMS
7.1 INTRODUCTION
(All Systems) 12 7.1.1 Identification of Safety-Related Systems 7.1.2 Identification of Safety and Power Generation Criteria 7.2 REACTOR PROTECTION (TRIP) SYSTEM (RPS) 12 7.2.1 Description 7.2.2 Conformance Analysis 7.3 ENGINEERED SAFETY FEATURES SYSTEM, INSTRUMENTATION AND CONTROL 13 7.3.1 Description 7.3.2 Analysis
-HPCS
-Shield Building A"""
"U
- "9
-')
-ADS
\\m /
- econdary Cgntain-
-LPCS ment Isolation
-RHR/LPCI
-Primary Containment
-CRVICS Isolation LCS
-MSPLCS
-Standby Power
-RHR/ Containment
-D-G Support Systems Spray
-Essential Service
-RHR/ Suppression Pool Water Cooling
-ESF Area Cooling
-Suppression Pool
-Pneumatic Supply Makeup
-CB Atmospheric
-Combustible Gas Control f
Control
-CB Chilled Water l
-SGTS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 14 l
7.4.1 Description 7.4.2 Analysis I
-RCIC
-RHR/ Shutdown Cooling
-SLC
-Remote Shutdown
(< s s
~/
i xvii
r GESSAR II 22A7007 238 NUCLEAR ISLRND Rev. O
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O Chapter /
Section Title Volume 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 14 7.5.1 Description 7.5.2 Analysis
-Nuclenet Control
-BOP Benchboard nsol
-Supervisory Moni-
-Standby Information toring Console Panel
-Display Control
-Rx Core Cooling BB System 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR 14 SAFETY 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses 7.6.4 References
-Neutron Monitoring
-FPCCS l
-Process Radiation
-DW/ Containment Monitoring Vacuum Relief
-Refueling Interlocks
-Vent & Pressure Control
-Leak Detection
-^
-Rod Pattern Control
-Suppression Pool
-HP/LP System perature Interlock Monitoring
-Recirculation Pump Trip 9
xviii
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N]
Chapter /
Section Title Volume 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 14 7.7.1 Description 7.7.2 Analysis 7.7.3 References
-RPV Instrumentation
-Leak Detection
-Rod Control &
-Rod Block Trip Information
-Fire Protection
-Recirculation Flow
-Drywell Chiller &
Control Cooling
-Feedwater Control
-Plant Instrument Air
-Performance Moni-
-Neutron Monitoring toring System
-Radwaste 7.8 NI/ BOP INTERFACES 14 (k -)
7.8.1 Essential Service Water (Supply)
System Instrumentation and Controls 7.P.2 Diesel Generator Fuel Oil Transfer System APPENDIX 7A I&C ELEMENTARY DIAGRAMS 15 O(.)
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Chapter /
Section Title Volume 8
ELECTRIC POWER
8.1 INTRODUCTION
16 8.1.1 Utility Grid Description 8.1.2 Onsite Electric Power System 8.1.3 Design Bases 8.2 OFFSITE POWER SYSTEM 16 8.2.1 Description 8.2.2 Analysis 8.2.3 Nuclear Island - BOP Interface 8.3 ONSITE POWER SYSTEMS 16 8.3.1 AC Power Systems 8.3.2 DC Power Systems 8.3.3 Fire Protection of Cable Systems e
O XX
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V' Chapter /
Section Title Volume 9
AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 17 9.1.1 New Fuel Storage (High Density) 9.1.2 Spent Fuel Storage (High Density) 9.1.3 Fuel Pool Cooling and Cleanup System 9.1.4 Fuel-Handling System 9.2 WATER SYSTEMS 17 9.2.1 Essential Service Water (ESW) System 9.2.2 Closed Cooling Water System 9.2.3 Demineralized Water Makeup System 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 9.2.6 Condensate Storage Facilities and Distribution System
's )
9.2.7 Plant Chilled Water Systems m
9.2.8 Heated Water Systems 9.2.9 Nuclear Island / BOP Interfaces 9.3 PROCESS AUXILIARIES 17 9.3.1 Compressed Air Systems 9.3.2 Process Sampling System 9.3.3 Floor and Equipment Drainage Systems 9.3.4 Chemical and Volume Control System (PWR) 9.3.5 Standby Liquid Control System 9.4 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEM 17 9.4.1 Control Room HVAC System 9.4.2 Fuel Building HVAC System 9.4.3 Auxiliary Building HVAC Systems 9.4.4 Turbine Building Area Ventilation System 9.4.5 Reactor Buildirig HVAC System f-)
I
(
/
~s xxi l
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Section Title Volume 9.4.6 Radwaste Building HVAC System 9.4.7 Diesel-Generator Buildings HVAC Systems 9.5 OTHER AUXILIARY SYSTEMS 17 9.5.1 Fire Protection System 9.5.2 Communications Systems 9.5.3 Lighting Systems 9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System 9.5.5 Diesel-Generator Cooling Water System 9.5.6 Diesel-Generator Starting Air System 9.5.7 Diesel Engine Lubrication System 9.5.8 Diesel Generator Combustion Air Intake and Exhaust System 9.5.9 Suppression Pool Cleanup System 9.5.10 Nuclear Island - BOP Interface APPENDIX 9A FIRE HAZARD ANALYSIS 18 O
xxii
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Chapter /
Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM 10.1
SUMMARY
DESCRIPTION 19 10.2 TURBINE GENERATOR 19 10.2.1 Design Bases Functional Limitations by Design or Operational Charac-teristics of the Reactor Coolant System 10.2.2
System Description
10.2.3 Turbine Disk Integrity 10.2.4 Evaluation 10.3 MAIN STEAM SUPPL 2 19 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 19 10.4.1 Main Condensers
/-
10.4.2 Condenser Air Removal System
((-
10.4.3 Main Condenser Evacuation System 10.4.4 Turbine Bypass System 10.4.5 Circulating Water System 10.4.6 Condensate Cleanup System 10.4.7 Condensate and Feedwater System 10.4.8 Steam Generator Blowdown System (PWR) 10.4.9 Auxiliary Feedwater System (PWR) xxiii
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Chapter /
Section Title Volume 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS 19 11.1.1 Fission Products 11.1.2 Activation Products 11.1.3 Tritium 11.1.4 Fuel Fission Production Inventory and Fuel Experience 11.1.5 Process Leakage Sources 11.1.6 Radwaste System 11.1.7 Radioactive Sources in the Gas Treatment System 11.1.8 Source Terms for Component Failures 11.1.9 Other Releases 11.1.10 References 11.2 LIQUID WASTE MANAGEMENT SYSTEM 19 11.2.1 Design Basis 11.2.2 System Descriptions 11.2.3 Estimated Releases 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 19 11.3.1 Design Bases 11.3.2 Main Condenser Steam Jet Air Ejector Low-Temp RECIIAR System Description 11.3.3 RECilAR System Operating Procedure 11.3.4 Radioactive Releases 11.3.5 References 11.4 SOLID RADWASTE SYSTEM 19 11.4.1 Design Bases 11.4.2
System Description
O xxiv
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Chapter /
Section Title Volume 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 19 11.5.1 Design Bases 11.5.2
System Description
11.5.3 Effluent Monitoring and Sampling 11.5.4 Process Monitoring and Sampling 11.5.5 Calibration and Maintenance 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 19
^
(v) l I
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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Chapter /
Section Title Volume 12 RADIATION PROTECTION 12.1 ENSURING TIIAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACllIEVABLE (ALARA) 19 12.1.1 Policy Considerations 12.1.2 Design Considerations 12.1.3 Operational Considerations 12.2 P^.DIATION SOURCES 19 12.2.1 Contained Sources 12.2.2 Airborne Radioactive Material Sources 12.2.3 References 12.3 RADIATION PROTECTION DESIGN FEATURES 19 12.3.1 Facility Design Features 12.3.2 Shielding 12.3.3 Ventilation 12.3.4 Area Radiation and Airborne Radioactivity Monitors 12.3.5 References 12.4 DOSE ASSESSMENT 19 12.5 IIEALTII PIlYSICS PROGRAM 19 O
xxvi
i GESSAR II 22A7007
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O Chapter /
I Section Title Volume I
13 CONDUCT OF OPERATIONS 19 4
i 1
}
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O XXVii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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O Chapter /
Section Title Volume 14 INITIAL TEST PROGRAM 14.1 TEST PROGRAM 20 14.1.1 Administrative Procedures (Testing) 14.1.2 Administrative Procedures (Modifications) 14.1.3 Test Objectives and Procedures 14.1.4 Fuel Loading and Initial Operation 14.1.5 Administrative Procedures (System Operation) 14.2 SPECIFIC INFORMATION TO BE INCLUDED IN SAFETY ANALYSIS REPORTS 20 14.2.1 Summary of Test Program and Objectives 14.2.2 Organization and Staffing 14.2.3 Test Procedures 14.2.4 Conduct of Test Program 14.2.5 Review, Evaluation and Approval of Test Results 14.2.6 Test Records 14.2.7 Conformance of Test Programs with Regulatory Guides 14.2.8 Utilization of Reactor Operating and Testing Experiences in the Develop-ment of Test Program 14.2.9 Trial Use of Plant Operating and Emergency Procedures 14.2.10 Initial Fuel Loading and Initial Criticality 14.2.11 Test Program Schedule 14.2.12 Individual Test Descriptions O
xxviii
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I Chapter /
Section Title Volume 15 ACCIDENT ANALYSES i
15.0
' GENERAL 21' 15.0.1 Analytical Objective 15.0.2 Analytical Categories 15.0.3 Event Evaluation 15.0.4 Nuclear Safety Operational Analysis (NSOA) Relationship 15.0.5 References 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 21 15.1.1 Loss of Feedwater Heating 15.1.2 Feedwater Controller Failure -
Maximum Demand I
15.1.3 Pressure Regulator Failure - Open 15.1.4 Inadvertent Safety / Relief Valve Opening O.
15.1.5 Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1.7 References 15.2 INCREASE IN REACTOR PRESSURE 21 f
15.2.1 Pressure Regulator Failure - Closed 15.2.2 Generator Load Rejection 15.2.3 Turbine Trip 15.2.4 MSLIV Closures 15.2.5 Loss of Condenser Vacuum 15.2.6 Loss of Offsite AC Power 15.2.7 Loss of Feedwater Flow 15.2.8 Feedwater Line Break 15.2.9 Failure of RHR Shutdown Cooling q
xxix
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TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 21 15.3.1 Recirculation Pump Trip 15.3.2 Recirculation Flow Control Failure -
Decreasing Flow 15.3.3 Recirculation Pump Seizure 15.3.4 Recirculation Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 21 15.4.1 Rod Withdrawal Error - Low Power 15.4.2 Rod Withdrawal Error at Power 15.4.3 Control Rod Maloperation (System Malfunction or Operator Error) 15.4.4 Abnormal Startup of Idle Recirculation Pump 15.4.5 Recirculation Flow Control with Increasing Flow 15.4.6 Chemical and Volume Control System Malfunctions 15.4.7 Misplaced Bundles Accident 15.4.8 Spectrum of Rod Ejection Assemblies 15.4.9 Control Rod Drop Accident (CRDA) 15.5 INCREASE IN REACTOR COOLANT INVENTORY 21 15.5.1 Inadvertent HPCS Startup 15.5.2 Chemical Volume Control System Malfunction (or Operator Error) 15.5.3 BWR Transients Which Increase Reactor Coolant Inventory O
XXX
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 15.6 DECREASE IN REACTOR COOLANT INVENTORY 21 15.6.1 Inadvertent Safety / Relief Valve Opening 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6.3 Steam Generator Tube Failure 15.6.4 Steam System Piping Break Outside Containment 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment 15.6.6 Feedwater Line Break - Outside Containment 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS
~
21 (O) 15.7.1 Radioactive Waste System Leak or Failure 15.7.2 Liquid Radioactive System Failure 15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure 15.7.4 Fuel-Handling Accident 15.7.5 Spent Fuel Cask Drop Accidents APPENDIX 15A PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS 21 APPENDIX 15B BWR/6 GENERIC ROD WITHDRAWAL ERROR ANALYSIS 21 XXXi i
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Chapter /
Section Title Volume 16 STANDARD TECHNICAL SPECIFICATIONS FOR GENERAL ELECTRIC BOILING WATER REACTORS 16.1 DEFINITIONS 22 16.1.1 Action 16.1.2 Average Planar Exposure 16.1.3 Average Planar Linear Heat Generation Rate 16.1.4 Channel Calibration 16.1.5 Channel Check 16.1.6 Channel Functional Test 16.1.7 Core Alteration 16.1.8 Critical Power Ratio 16.1.9 Dose Equivalent I-131 16.1.10 E-Average Disintegration Energy 16.1.11 Emergency Core Cooling System (ECCS)
Response Time 16.1.12 Frequency Notation 16.1.13 Identified Leakage 16.1.14 Isolation System Response Time 16.1.15 Limiting Control Rod Pattern 16.1.16 Linear Heat Generation Rate 16.1.17 Logic System Functional Test 16.1.18 Maximum Total Peaking Factor 16.1.19 Minimum Critical Power Ratio 16.1.20 Operable - Operability 16.1.21 Operational Condition (Condition) 16.1.22 Physics Test 16.1.23 Pressure Boundary Leakage l
16.1.24 Primary Containment Integrity 16.1.25 Rated Thermal Power 16.1.26 Reactor Protection System Response Time 16.1.27 Recirculation Pump Trip System Response Time xxxii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued) a Chapter /
Section Title Volume 16.1.28 Reportable Occurrence 16.1.29 Rod Density 16.1.30 Secondary Containment Integrity 16.1.31 Shutdown Margin 16.1.32 Staggered Test Basis 16.1.33 Thermal Power 16.1.34 Total Peaking Factor 16.1.35 Unidentified Leakage 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 22 16.2.1 Safety Limits 16.2.2 Limiting Safety System Settings 16.B2 SAFETY LIMITS 22 16.B2.1 Bases
( \\
16.B2.2 Limiting Safety System Settings 16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 22 16.3/4.0 Applicability 16.3/4.1 Surveillance Requirements 16.3/4.2 Power Distribution Limits 16.3/4.3 Instrumentation 16.3/4.4 Reactor Coolant System 16.3/4.5 Emergency Core Cooling Systems 16.3/4.6 Containment Systems 16.3/4.7 Plant Systems 16.3/4.8 Electrical Power Systems 16.3/4.9 Refueling Operations 16.3/4.10 Special Test Exceptions O.
XXXiii
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Chapter /
Section Title Volume 16.B3/4.0 Applicability 16.B3/4.1 Reactivity Control Systems 16.B3/4.2 Power Distribution Limits 16.B3/4.3 Instrumentation 16.B3/4.4 Reactor Coolant System 16.83/4.5 Emergency Core Cooling System 16.B3/4.6 Containment Systems 16.B3/4.7 Plant Systems 16.B3/4.8 Electrical Power Systems 16.83/4.9 Refueling Operations 16.B3/4.10 Special Test Exceptions 16.5 DESIGN FEATURES 22 16.5.1 Site 16.5.2 Containment 16.5.3 Reactor Core 16.5.4 Reactor Coolant System 16.5.5 Meteorological Tower Location 16.5.6 Fuel Storage 16.5.7 Component Cyclic or Transient Limit O
1 XXXiV L
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}
I i
Chapter /
Section Title Volume 17 QUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 22 l
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17.2 QUALITY ASSURANCE DURING THE OPERATING PHASE _
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