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ABB-C-E Sys 80+ Structural Analysis of Nuclear Island & Nuclear Annex Structures
	ML20044C006
Person / Time
Site:	05200002
Issue date:	03/17/1993
From:	; ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	;
Shared Package
ML20044C005	List: LD-93-042, Forwards ABB-C-E Sys 80+ Progress Rept on Structural Analysis of Nuclear Island & Nuclear Annex Structures, to Support 930317 & 18 Meeting in Charlotte,Nc Consisting of Std Structural Design Criteria Draft Spec; ML20044C006;
References
	NUDOCS 9303160103
	Download: ML20044C006 (72)
	v • d • e

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'ABB-CE SYSTEM 80+

PROGRESS REPORT ON l

STRUCTURAL ANALYSIS ~ OF-NUCLEAR ISLAND 1

AND NUCLEAR ANNEX STRUCTURES

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PRESENTED TO:

U.S. NUCLEAR REGULATORY COMMISSION j

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PRESENTED : BY:

ABB COMBUSTION ENGINEERING j

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SYSTEM 80+

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9303160103 930310-l PDR.ADOCK 05200002 A

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AGENDA

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.I Presentation of Structural Finite-Element Model of Nuclear Island' 1.

and Nuclear Annex Structures

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Benchmark Tests for Computation of Design Forces-t

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Seismic Loading i

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Study of modeling Soll; Media with Soil Springs i

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SYSTEM 80+ -

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OBJECTIVE Compute the distribution of applied loads on Nuclear Island and Annex Structures.

Develop Design Loads for individual shear walls, floor slabs, and the basemat to be used in detailed design.

Assess areas of high stress.

METHODOLOGY Develop a detailed 3-D Finite Element Model of the Nuclear Island and the Annex Structures Use shell elements to model the shear walls and the floor diaphragms Use solid elements to model thick walls and the basemat Use beam elements to model the columns The Finite Element Model will be used for load distribution i

of all structures except for the SCV.

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SYSTEM 80+

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Benchmark tests with simplified finite element models were performed to verify proper computation of desian forces from FE analysis results.

The behavior of the FE Model is also cross-checked versus that of the Stick Model used in the Dynamic SSI analysis.

The distribution of internal forces, such as floor shears, is also cross-checked versus hand calculations.

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Seismic Loading ARR rs **..se

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The applied loading at each elevation (for all structures) will be the envelope of the floor accelerations from all soil cases from the SSI analyses, multiplied by the appropriate mass at each elevation of each structure Three Separate static analyses will be conducted for the applied earthquake loading in the:

N-S Direction (X)

E-W Direction Y)

Vertical Directi(on (Z)

All loads in the same direction will be applied in phase Slabs have minimum 3 ft. thickness. The maximum dimensions of a supported slab are approximately 25'x25',

with natural frequency in the rigid range (>33 Hz).

Therefore, vertically, floor flexibility effects do not exist.

The model will distribute the loading in accordance with various wall and floor stiffness characteristics Earthquake loading will be combined using the 100% - 40% - 40% rule Resultant earthquake loading will_ be combined with other loadings as per CESSAR Table 3.8-5 ABB

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Study of Modeling Soil Media with Soil Springs ARR rsna na.
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I SYSTEM 80+ STANDARD PLANT STRUCTURAL DESZGN CRITERIA SPECIFICATION FILE No. 4248-02-1622.00-0001 h
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March 8, 1993 SYSTEM B0+ STANDARD PIANT
.i STRUCTURAL DESIGN CRITERIA SPECIFICATION Table of Contents t
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1.0 INTRODUCTION
/ OBJECTIVE 2.0 DESIGN CRITERIA SCOPE 3.0 DEFINITIONS AND ABBREVIATIONS 3.1 Definitions t
3.2 Abbreviations 4.0 SYSTEM B0+ STANDARD PLANT DESCRIPTION 4.1 Nuclear Island 4.2 Other Category.I Structures (later)
-l 4.3 Non-Nuclear Safety Related Structures (later) 5.0 GOVERNING DESIGN CRITERIA DOCUMENTS AND REFERENCES 5.1 Regulatory Documents 5.1.1 Title 10 Code of Federal Regulations l
5.1.2 Nuclear Regulatory Guides 5.1.3 Standard Review Plan 5.2 CESSAR-DC Design Criteria 5.3 codes and Standards 5.4 Licensing submittals 5.5 Specifications 5.6 Other References and Supporting Documents 5.6.1 Design calculations 5.6.2 Correspondence 5 G.3 Reference Drawings-5.6.4 Other Reference Documents 6.0 Quality Assurance Requirements 7.0 Structural Design Loads and Load Combinations 7.1 Design Loads 7.1.1 Normal Loads e
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i 7.1.2 Severe Environmental Loads t
f 7.1.3 Extrame Environmental Loads 7.1.4 Abnormal Loads
'7.1.5 other Loads
'l 7.2 Design Load Combinations I
7.2.1 Loading Combinations for Category 1 Concrete Structures 7.2.2 Loading Combinations for Category 1 Steel Structures 7.2.3 Loading Combinations for Stability Checks (later) 7.2.4 Applicability of Loads 7.2.5 Notes and Definitions (later)
B.0 Structural Analysis and Design E.1 Analysis B.1.1 General 8.1.2 Seismic Analysis i
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8.1.2 Structural Analysis t
B.2 Structural Design t
8.2.1 General Requirements i
B.2.2 Design Procedures 8.2.3 Special Design Criteria 9.0 Construction; Forming, Fabrication, and Erection 9.1 Concrete 9.2 Steel 9.3 Structural Attachments I
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10.0-
-Structural Acceptance criteria 11.0 Materials.
l Appendix A-Supplemental Criteria for Outlying Category I Structures e
Appendix B Supplemental Criteria for Non-Safety Related Structures j
Appendix C.
Tables Table C-1 Structural Live Loads (ref. 6.5.3) i Table C-2 Summary of Applicable Structural Loads l
3 (ref. CESSAR-DC Table 3.2-4)
I Appendix D Figures i
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SYSTEM 80+ STANDARD PIANT STRUCTURAL DESIGN CRITERIA i
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1.0 INTRODUCTION
/ OBJECTIVE
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This document provides the design criteria for the analysis and design of structures that comprise the System 80+ Standard Plant.
It is intended that this document will serve as a single source reference, see Section 5.0, to l
design requirements specified in other licensing and implementation documents.
i Licensing documents would include; 10CFR 50, Standard Review Plan, CESSAR-DC, and Safety Evaluation Report (NUREG-1462).
Implementation documents would include; Design Calculations, Design and Constructions Specifications, and
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Engineering Drawings.
In addition, this document will provide added design information and criteria not adequately presented elsewhere.
Variance from these criteria are not permitted for safety related components i
without adequate justification and documentation. For non-safety related components, variances in keeping with good engineering and construction practices are permitted but must be reflected in final design and construction documents.
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-i 2.0 DESIGN. CRITERIA SCOPE' t
6 The information presented in thir document is to be used in the analysis sud design of all structural components for the System 80+ Standard Plant structures as identified in Reference 5.2.0, Table 3.2-1 (Sh 23 of 27).
Structural components required as part of the primary containment pressure boundary or for its support are considered Safety Class 2.
Those components required to perform some other safety related function are considered Safety I
Class 3.
Safety Class 1 applies to the NSSS components and does not apply within this criteria document.
All structures required to shut down and maintain the reactor in a safe and j
orderly condition or prevent the uncontrolled release of excessive amounts of i
radioactivity have a seismic classification of category I.
These structures are designed to withstand, without loss of function, the most severe natural phenomena on record for the site, with appropriate margins included in the design for uncertainties in historical data.
The non-Category I structures determined to have sufficient mass to possibly i
impair the integrity of Seismic Category I structures or components upon collapse are classified as Category II.. Category II structures are analyzed i
to prevent their failure in the direction of a Seismic Category I structure under SSE conditions. The margin of safety of these structures is equivalent i
to that of Seismic Category I structures.
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'i Refer to CESSAR-DC Table 3.2-1 for a listing of all System 80+ Standard Plant I
Structures. In addition to listing each structure, this table identifies the applicable Seismic Category and Safety Class for each.
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t While explicitly written for the category I, Safety classes 2 and 3, structures attached to the Nuclear Island, this criteria will also apply to -
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other outlying Category I structures as well as non-safety related structures.
These other structures are identified in Section 3.2 of this criteria. Refer to Appendices A and B (later)for specific criteria for these outlying Category I and non-safety related structures. Applicable Nuclear Island structures identified in Table 3.2-1 of Ref 6.1.0 are the containment Shield Building, containment Internal Structures, Shield Building sub-structures, and the connecting Nuclear Annex.
Station
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Outlying Category I structures will include the Radwaste Facility, Service Water Pump Structure, Component Cooling Water Heat Exchanger Structure, Diesel Fuel Storage, and Dike (CVCS outdoor tanks).
The Non-Nuclear Safety Related Structures would include the Turbine Building, Service Building, Administration Building, Warehouse, and Fire Pump House.
Primary components shall consist of floor and roof slabs (including the t
foundation basemat), walls, and steel beams and' columns. Steel beams and columns will be considered within the scope of this criteria if their primary function is to provide support to walls, floors, or roof slabs. Steel members whose primary function is equipment support will be qualified under other design criteria / specification.
i Information presented in this document will be sufficiently comprehensive in i
i nature tot provide the criteria necessary to perform an analysis and translate that a.
analysis into a workable design... - -
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provide correlation of analysis, design, and construction requirements j
I with those in Sections 3.8.3, 3.8.4, and 3.8.5 of the CESSAR-DC.
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Miscellaneous components listed below, while not primary structural supports, must be considered in the design of primary components as to their loads and Miscellaneous components will be further addressed in method of attachment.
specific design / installation specifications.
The following is a list of F
related miscellaneous components; I
Personnel Airlocks Equipment Hatch 5
NSSS Component Supports piping Supports t
HVAC Equipment Supports Cable Tray Supports Cranes and Hoists Equipment Mountings e
Refueling Equipment Miscellaneous platforms ICI Tube Supports Sump Screens Steel Tanks Wall, Floor penetration seals Pressure Doors Fire Barriers Typical Support / Anchorage Embedments I
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March 8, 1993 3.0 DEFINITIONS and SYMBOLS 3.1 Definitions (later) 3.2 Abbreviations ABB-CE
' Asea Brown Boveri-Combustion Engineering ACI American Concrete Institute ADB Administration Building AISC American Institute of Steel Construction ALWR Advanced light water reactor ANSI American National Standards Institute ANSYS Ceneral purpose finite element computer program ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials AWS American Welding Society BGS Bulk Gas Storage Area CCWX Component Cooling Water Ex Structure CCX Control Complex CESSAR-DC Combustion Engineering Standard Safety Analysis Report-Design Certification CFR Code of Federal Regulations CLT Cooling Tower CTF Combustion Turbine Facility CTFS Cumbustion Turbine Fuel Storage CWPS Circulation Water Pump Structure DBA Design Basis Accident DBE Design Basis Earthquake DEES Duke Engineering & Service DFS Diesel Fuel Storage DGB Diesel Generator Building (Area)
DS Discharge Structure EFST or Emergency Feedwater Storage Tank,
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'ETWST EPRI Electric Phwer Research Institute EVSE Ex-vessel' steam explosion FHB Fuel Handling Building (Ares)
FPH Fire Pump House FSAR Final Safety Analysis Report i
FSER Final Safety Evaluation Report GDC general design criteria / criterion HVAC Heating Ventilation and Air Conditioning INT Holdup Volume Tank I&C(s)
Instrumentation & Control (s)
ICI In-core Instrumentation Institute of Nuclear Power Operastions INPO IS Intake Structure i
LBB leak-before-break f
MSVH Main Sterm Valve House NA Nuclear Annex National Fire Protection Association NFPA NI
. Nuclear Island NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NUREG NRC technical report designation OBE Operating Basis Earthquake PAP Personnel Access Portal PMF Probable Maximum Flood i
PMP Probable Maximum Precipitation PRT Pressurizer Relief Tank 1
PRZ Pressurizer QA Quality Assurance RCP Reactor Coolant Pump 1
f RDT Reactor Drain Tank I
RFAI Request for additional information
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RH Relay House RWF Radwaste Facility RXB(valid?) Reactor Building r
(Should this be the Shield Building?)
SB Service Building SAR Safety Analysis Report SCV Steel Containment Vessel SER Safety Evaluation Report SFSA Spent Fuel Storage Area F
SG Steam Generator SRP Standard Review Plan SSC structure, system, and component SSCV Structural Steel Containment Vessel r
SSDS Station Service Water Discharge Structure SSE Safe Shutdown Earthquake l
SSIS Station Service Water Intake Structure SSPS Station Service Water Pump Structure SSWP Station Service Water Pond SIB ??
Service Building b
TB Turbine Building TBD to be determined WH Warehouse 4.0 SYSTEM 80+ STANDARD PLANT DESCRIPTION Refer to CESSAR-DC Section 1.2 with accompanying figures.
t 4.1 Nuclear Island The structures described below, including the common underlying
'basemat, comprise the Nuclear Island. Refer to System 80+ General Arrangement drawings ALWR-001 through ALWR-011 for details of these,
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March 8, 1993 structures. The Nucioar Island, except for electrical and mechanical system tie-ins, is structurally isolated from adjacent structures.
b 4.1.1 The containment Internal structures are housed inside the i
spherical steel containment vessel. The purpose of these internal structures is to provide structural support and shielding for the i
containment interior and its vital safety-related equipment.
These with
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structures will be primarily constructed of reinforced concrete, F
some structural steel.
These structures are described in CESSAR-DC 5
Section 3.8-3.
4.1.2 The Containment Shield Building surrounds the containment vessel and sub-structures as well as their equipment and protects them from It consists of postulated missiles and other environmental effects.
i a cylindrical reinforced concrete shell wall which is topped by a l
hemispherical reinforced concrete dome roof.
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i March 8, 1993 4.1.3 The Shield Building Sub-structures (or Lower Volume) consist of a system of reinforced concrete walls and slabs and a containment support pedestal beneath the containment vessel and inside the shield.
building. The purpose of these sub-structures is to support the containment vessel and house auxiliary safety related equipment.
4.1.4 The Nuclear System Annex is a multi-level reinforced concrete structure surrounding the Shield Building. The Nuclear Annex is i
integrally attached to the Shield Building and adds lateral bracing to it while providing partial tornado wind and missile protection.
The Nuclear Annex provides additional protected areas'(Control Complex, Diesel Generator Bldg., Fuel Handling Bldg., Valve House, etc) for safety related equipment that cannot be housed inside containment.
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4.1.5 The Nuclear Island foundation is a reinforced concrete basemat j
which underlies all of the safety-related components identified l
above, providing them with a common foundation.
4.2 Other Category I Structures (later) 4.2.1 Radwaste Facility 4.2.2 Station Service Water Pump Structure 4.2.3 Component Cooling Water Heat Exchanger Structure 4.2.4 Diesel Fuel Storage 4.2.5 Dike (CVCS outdoor Tanks) 4.3 Non-Nuclear Safety Related Structures (later) 4.3.1 Turbine Building b
4.3.2 Station Services Building 4.3.3 Administration Building 4.3.4 Warehouse March 8, 1993 4
4.3.5 Fire Pump House 4.3.6 Combustion Turbine Building i
1 5.0 COVERNING DESIGN CRITERIA DOCUMENTS and REFERENCES TBD Oater) DSER item 3-1 and SRP section 5.2.11 ask that all specific editions be identified.
CESSAR-DC Table 1.8-4, page 4 of 10, The following references identify regulatory and licensing requirements applicable to the analysis and design of the System 80+ Standard Plant structures.
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i 5.1 Regulatory Documents 5.1.1 Title 10, Code of Federal Regulations o
The CODE OF FEDERAL REGULATIONS, Title 10, Chapter I, Part 50 provides the legislative requirements for the licensing of production and utilization facilities. Appendix A to 10CFR Part 50 provides the Design Basis general design criteria for nuclear power plants. General Design Criteria are also found in Section 3.1 of the CESSAR-DC.
Of the criteria identified in Appendix A, the following criteria are directly related to the design and construction of the System 80+ Standard Plant structures.
Criterion 1 - Quality Standards and Records i
Structures, systems, and components important to safety shall be designed, fabricated, erected,and tested to quality standards commensurate with the importance of the safety functions performed.
Criterion 2 - Design Bases for Protection Against Natural phenomena Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches e
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without loss of capability to perform their safety functions.
criterion 3 - Fire Protection Structures, systems, and components shall be designed and located a
to minimize, consistent with other safety requirements, the probability and ef fect of fires and explosions.
p criterion 4 - Environmental and Dynamic Effects Structures, systams, and components important to safety shall be designed to accommodate the effects of and to be compatible.with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-cooling accidents.-
Criterion 5 - Sharing of structures, systems, and components Structures, systems, and components important to safety shall not t
be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in'the event of an' accident in one unit, an orderly shutdown and cooldown of the remaining units.
Criterion 16 - Containment Design Reactor Containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Criterion 19 - Control Room A control room shall be provided from which actions can be taken l
to operate the nuclear power unit safely under normal conditions I
and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents t
Criterion 50 - Containment Design Basis The reactor containment structure, including access openings,-
i March 8, 1993 i
penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and te=perature conditions resulting from any loss-of-coolant
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accident.
Criterion 52 - Capability for Containment Leakage Rate Testing j
The reactor containment and other equipment which may be subjected to containment test conditions shall-be designed so that periodic integrated leakage rate testing can be conducted at l
t containment design pressure.
Criterion 53 - Provisions for Containment Testing and Inspection l
The reactor containment shall be designed to permit (1) such as
-j appropriate periodic inspection of all important areast, penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak l
tightness of penetrations which have resilient seals and expansion bellows.
4 5.1.2 Nuclear Regulatory Guides 5.1.2.1 USNRC Regulatory Guide 1.29, Sept. 1978, " Seismic Design Classification" f
5.1.2.2 USNRC Regulatory Guide 1.31, Apr. 1978, " Control of Stainless Steel Welding" 5.1.2.3 USNRC Regulatory Guide 1.59, Aug. 1977, " Design Basis Floods for Nuclear Power Plante" 5.1.2.4 USNRC Regulatory Guide 1.60, Dec. 1973, " Design Response Spectra for Seismic Design of Nuclear Power Plants" 5.1.2.5 USNRC Regulatory Guide 1.61, Oct. 1973, " Damping values for Seismic Design of Nuclear Power Plants" J
5.1.2.6 USNRC Regulatory Guide 1.76, Apr. 1974, " Design Basis Tornado for Nuclear Power Plants" i 1 i
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5.1.2.7 USNRC Regulatory Guide 1.94, Apr. 1976, " Quality Assurance t
Requirements for the Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction
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Phase of Nuclear Power Plants" 5.1.2.8 USNRC Regulatory Guide 1.102, Sept. 1976, " Flood Protection for Nucienr Power Plants" 5.1.2.9 USNRC Regulatory Guide 1.115, July 1977, " Protection Against Low-Trajectory Turbine Missiles" 5.1.2.10 USNRC Regulatory Guide 1.117, Apr. 1978, " Tornado Design Classification" 5.1.2.11 USNRC Regulatory Guide 1.142, Oct. 1981, " Safety Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments)"
i 5.1.3 Standard Review Plan i
Refer to CESSAR-DC Tables 1.8-4 & 1.0-5 for deviation and compliance comments regarding site specific criteria.
5.1.3.1 USNRC Standard Review Plan 2.4.10, July 1981, " Flooding Protection Requirements" 5.1.3.2 USNRC Standard Review Plan 3.2.1, July 1981, " Seismic
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Classification" 5.1.3.3 USNRC Standard Review Plan 3.3.1 July 1981, " Wind Loadings" 5.1.3.4 USNRC Standard Review Plan 3.3.2, July 1981, " Tornado Loadings" 5.1.3.5 USNRC Standard Review Plan 3.4.1, July 1981, " Flood.
Protection" 5.1.3.6 USNRC Standard Review Plan 3.4.2, July 1981, " Analysis Procedures" 5.1.3.7 USNRC Standard Review Plan 3.5.1.1, July 1981, " Internally Generated Missiles (Outside Containment)"
5.1.3.8 USNRC Standard Review Plan 3.5.1.2, July 1981, " Internally Generated Missiles (Inside Containment)",
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l March 8, 1993 5.1.3.9 USNRC Standard Review Plan 3.5.1.3, July 1981, " Turbine Missiles" 5.1.3.10 USNRC Standard Review Plan 3.5.1.4, July 1981, " Missiles 7
Generated by Natural Phenomena" i
5.1.3.11 USNRC Standard Review Plan 3.5.1.5, July 1981, " Site l
Proximity Missiles (Except Aircraft)"
5.1.3.12 USNRC Standard Review Plan 3.5.1.6, July 1981, " Aircraft i
i Hazards" 5.1.3.13 USNRC Standard Review Plan 3.5.2, July 1981, " Structures, t
t Systems, and Components to be Protected from Externally Generated Missiles" l
5.1.3.14 USNRC Standard Review Plan 3.5.3, July 1981, " Barrier Design Procedures" 5.1.3.15 USNRC Standard Review Plan 3.6.1, July 1981, " Plant Design for Protection Against Postulated Piping Failures in Fluid Systems outside containment" 5.1.3.16 USNRC Standaid Review Plan 3.6.2, July 1981, " Determination of Rupture Locations and Dynamic Effects Associated with the' r
Postulated Rupture of Piping" 5.1.3.17 USNRC Standard Review Plan 3.7.2, July 1981, " Seismic System-
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Analysis" 5.1.3.18 USNRC Standard Review Plan 3.8.1, July 1981, " Concrete Containment" 5.1.3.19 USNRC StLndard Review Plan 3.8.3, July 1981, " Concrete and Steel Internal Structures of Steel or concrete containments" 5.1.3.20 USNRC Standard Review Plan 3.8.4, July 1981, "Other Seismic Category I structures" 5.1.3.21 USNRC Standard Review Plan 3.8.5, July 1981, " Foundations" 5.2.0 CESSAR-DC Design Criteria i
5.2.1 Section 3.3; Wind and Tornado Loadings 5.2.2 Section 3.4; Water Level (Flood) Design [
March 8, 1993 j
5.2.3 Section 3.5; Missile Protection l
i 5.2.4 Section 3.8.3; Concrete and Structural Steel Interior Structure 5.2.5 Section 3.8.4; other Category 1 Structures 5.2.6 Section 3.8.5; Foundation t
5.3.0 codes'and Standards 1.
5.3.1 ACI 318-897, " Building Code Requirements for Reinforced concrete" 5.3.2 ACI 349-85, " Code Requirements for Nuclear Safety-Related concrete Structures", thru 1990 Supplement?
5.3.3 AISC Manual of Steel Construction, Allowable Stress Design w/
" Specification for Structural Steel Buildings", June 1, 1989 5.3.4 ANSI N45.2.5-(later) "Supplamentary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structrual Steel During the Construction Phase of Nuclear Plants" 5.3.5 ANSI /AISC N690-1984, " Nuclear Facilities - Steel Safety-Related Structures for Design, Fabrication,-and Erection."
5.3.6 ANSI /ASCE 7-88, " Minimum Design Loads for Buildings and Other, Structures", Nov 27, 1990 (previously ANSI A58.1-1982) 5.3.7 ASCE Standard 4-86, " Seismic Analysis.of Safety-Related Nuclear Structures" 5.3.8 Annual Book of ASTM Standards, current issue, as applicable for naterial properties.
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5.3.9 AWS " Structural Welding code", ANSI /AWS D1.1 (latest?[
l 5.4.0 Licensing Submittals 5.4.1 Structures ITAAC (later)
T 5.5.0 Specifications 5.5.1 " Specification for Structural Joints Using ASTM A325 or A490 Bolts", Nov 13, 1985, Research Council on Structural Connections
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5.5.2 Specification for the Design, Fabrication, and Erection of Structural Steel, Miscellaneous Steel and Other Steel Construction for Nuclear Safety Related Structural Steel (later) l 5.5.3 Specification for attachments to QA condition 1 structures (later) 5.5.4 Specification for the installation and use of standardized t
embedded plates (later) 5.5.5 Specification for the design and installation of expansion anchors (later) 5.5.6 Specification for the Design and Placement of Concrete for Nuclear Safety Related Structures (later)
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5.6.0 Other References and Supporting Documents 5.6.1.0 Calculations 5.6.1.1 DE&S Calculation No. 4248-04-1614.03-0001, aseismic i
Structural Model Development Calculation" (later) 5.6.1.2 DE&S Calculation No. 4248-04-1622.00-0001, " Interior
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1 and Shield Building Analysis criteria Development" Structure (later) 5.6.1.3 DE&S Calculation No. 4248-04-1622.00-0002,," Interior Structure and Shield Building Preliminary Analysis calculation (Sliding and Overturning)", Rev. 0 1
5.6.1.4 (later), Roger Wagstaff's calc on steam explosion due to
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failure of the Reactor Vessel 5.6.2.0 Correspondence 5.6.2.1 Letter from NRC Director Lester S. Rubenstein to ALWR Utility Steering Committee Chairman Edwin E. Kinter, Project
669, dated March 25, 1988 (includes Interim Reg. Guide 1.76) 5.6.2.2 Letter No. ALWR-88-293 from Combustion Engineering to Duke i
Power Company, dated June 20, 1988.
5.6.2.3 Letter No. ALWR-544 dated January 29,1993 from Duke Engineering and Services to ABB-Impell, Subj.-Global Loads and
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6 Combinations 5.6.3.0 Reference Drawings (later) 5.6.3.1 CE System 80+ ALWR Reactor Building General Arrangement Drawings, ALWR-0001 through ALWR-011,'Rev.
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-5.6.4.0 Other Reference Documents 5.6.4.1 Advanced Light Water Reactor Utility Requirements Document, issued by the Electric Power Research Institute (EPRI), Chapter-1 through 6 Rev.0 5.6.4.2 ASCE Paper No. 3269, " Wind Forces on Structures,"
t Transactions of the American Society of Civil Engineers, Vol.
126, Part II, 1961.
5.6.4.3 ASCE Paper No. 4933, " Wind Loads on Dome-cylinder and Dome-Cone Shapes," Journal of the Structural Division - Proceedings l
of the American Society of Civil Engineers, Oct.L1966.
5.6.4.4 Introductory soil Mechanics and Foundations,- by George B.
f Sowers and George F. Sowers, 3rd Edition, 1970, MacMillan Publishing Co.
5.6.4.5 (TBD later)_NRC Paper SECY-89-013 " Design Requirements Related to the Evolutionary Advanced Light Water Reactors (ALWRs)", Jan 19, 1989 (verify for structural applicability).
6.0 QUALITY ASSURANCE REQUIREMENTS The requirements for a QA progrms were established in 10CFR50, Appendix B.
These requirements are further spelled out in Reg Guide 1.94 by reference to ANSI Standard N45.2.5.
Duke Engineering & Services has instituted a quality.
j assurance program based upon these regulatory requirements and ASME Code NQA-1-19D9, through the NQA-1b-1991 Addenda. This QA program is documented in the i
March B, 1993 l
DE&S Quality Assurance Program Manual and is administered by the DEkB QA Manager. The Program is based upon the 4 QA Classifications, QAl - QA4, listed below and the 2 seismic classifications, category I and II, identified i
in section 2.0 of this criteria.
I The following QA Condition designations are applicable to the DEES Graded QA program and'are identified in Document QAPD of the DE&S QA Manual TBD later;_
check this with Reg Guide 1.26 b
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March 8, 1993' QA Condition 1 - applicable to all Category I Nuclear Safety Related structures and components QA Condition 2 - applicable to structures and components involved in the control or retention of. radioactive waste l
QA condition 3 - applicable to fire protection relat'ed components QA Condition 4 - applicable to all Category II structures and components This criteria is intended for the design of structures and components conforming to any of the above designations plus those structures and components that are non-safety related. Refer to CESSAR-DC Table 3.2-1 for applicable classifications. All QA conditions shall therefore apply to this criteria i
6.1 Calculations Address quality and accuracy?
6.2 Materials (later)
Ref. CESSAR-DC 3.8.4.6.2 for mill test report required by ASTM sopecifications plus other codes and specs.
6.3 Construction i
Ref CESSAR-DC 3.8.4.7 for quality control testing reference.
r Allowable tolerances 7.0 STRUCTURAL DESIGN LOADS and LOAD COMBINATIONS Design loads on Category 1 structural components for this Nuclear Generation Facility are identified in CESSAR-DC Table 3.8-5.
These loads are separated l
I into four (4) categories: normal loads, severe environmental loads, extreme environmental loads, and abnormal loads.
Each Nuclear Generation Facility designed under this standardized plant I
program must verify adherence to the standard envelop of design loads presented in this document for all Category I structures and components.
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Specific loa,1s as defined below, such as E, F ', W, We, or F', must be shown to lie within the standard envelop or additional analyses nest be performed to l
t verify structural adequacy.
4 Not all listed loads are necessarily applicable to all structures. The loads for which each structure, or part thereof, should be designed will. depend on the conditions to which that particular structure could be subjected.
Refer to Table C-2 (later), (formerly CESSAR-DC Table 3.2-4) for a summary of' applicable loads for a summary of applicable loads.
is assumed that the plant turbines will be oriented to preclude turbine It missile strikes on the Reactor Building. Protection must be provided against missile strikes on safety related structures from missiles internally generated by plant equipment or components due to operating malfunctions, earthquake s, etc.
It is assumed that the plant will not be sited in areas subject tot potential.
i aircraft crashes, explosive hazards from nearby industrial sites or transportation routes, projectiles or missiles generated from activities of nearby military installations, landslides, tsunami or volcanic activity.
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7.1 Design Loads 7.1.1 Normal Loads i
Normal loads are those loads encountered during normal plant j
operation and shutdown. They includes Dead loads (D), Ground water i
liquid pressure loads (F), Live loads (L), Soil pressure loads (H),
i t
Thermal loads (To), and Pipe reactions (Ro).
Dead loads or their related internal forces and moments due to D-the mass of the structure plus any permanently mounted equipment loads and hydrostatic loads. Normal ground water hydrostatic loads are broken out separately and defined below as "F"
loads). i i
1
.q
f.
6 March 8, 1993 L-Live loads or their related internal forces and moments, 1
including any movable equipment loads and other loads which vary with intensity and occurrence. Soil pressure loads, "H"
loads, are considered one part of the live loade in CESSAR-nc Table 3.8-5.
For equipment supports, it also includes loads due to vibration and any support movement effects. See Table C-2 (later) for design live loads.
Live loads shall be applied, removed, or shifted in location as i
necessary to obtain the worst case loading conditions for maximizing internal moments and forces.
Impact forces due to 1
moving loads shall be applied as appropriate. Designated laydown/ maintenance areas shall be designed for the live loada due I
to the materials and equipment intended to be placed there.
Temporary laydown areas will apply to staging areas for equipment replacement such as various heat exchangers, reactor coolant pumps, f
stern generators, etc.
For designated laydown areas see (later) t Uniform design loads for walls and floors due to equipment and other live loads were provided in reference 6.5.3, see Appendix A, l
page A-1.
The tabulated floor which ones? live loads _do not include (wording? What do the tabulated loads represent? the
_y i
equipment live loads.
Precipitation, snow and/or ice, will produce live loads of 50 psf (CESSAR-DC Section 3.8.4.3 F). What about rain?
i i
H - Lateral soil pressure load with normal ground water level at f
El. B88-9",
2'-0" below plant finished yard grade elevation (El.
I 90'-9").
The effects of vehicles, cranes, material stockpiles, etc. exerting loads on the soil adjacent to building walls shall be considered. In CESSAR-DC Table 3.0-5, soil pressures are includsa with live loads.
("H" is used in CESSAR-DC Table 3.8-5 Section IV)
I
! f
1 l
March 8, 1993 t
H'-Soil pressure-lateral load with flood level l'-0" below plant finished yard grade.
H' should be used in lieu of H whenever flood loads are being considered.
F-(this is from ACI 349, CESSAR-DC lists only F' for Max Flood under Normal Load) Liquid pressure load due to norDal ground water level.
The liquid pressure load has a lateral component (Fh) and a vertical, or buoyancy component (PV).
Maximum ground water level is established at 2'-0" below plant finished grade elevation.
Lateral force Fh may be considered to be included with the lateral i
soil pressure. Loads due to F and F' are included in CESSAR-DC
,.i j
Table 3.8-5 as dead loads.
i See F'- Loads generated by the design flood specified for the plant.
1 I
CESSAR-DC section 2.4.
A maximwm flood level of l'-0" below plant finished yard grade, CESSAR-DC Section 3.4.1, shall be used in design. Site specific floods greater than this must be addressed as needed for the applicable plants. The hydrostatic effects of
?
the flood waters, plus the effects'of saturated soil up to grade elevation, plus the dynamic effects of wind generated waves shall be considered.
For ease in analyses, flood load F' is sub-categorized as follows:
i I
i F,'
= Total liquid pressure load = F'r + F,,', where l
Fy' = Vertical component or buoyancy force of flood liquid I
pressure load F,g ' = Lateral component of flood liquid pressure load.
F,'
= Flood generated missile loads Reg Guide 1.29' requires flood protection for the Diesel Fuel 011 Storage Tank. Won't it be above the projected flood level along with I i
+
March 8, 1993 t
}
1 everything else?
t To - Thermal effects and loads during normal operating or shutdown conditions, based on the most critical transient or steady state j
I condition.
-[
The following ambient temperature values during normal conditions shall be used as a basis for design, based on CESSAR-DC
-I Table 2.0-1:
e 1% Exceedance Values -
Definition of "exceedance", (later)
Maximum:
1000 F dry bulb /778 F coincident wet bulb Minimums
-108 F 0% Exceedance Values (Historical limit)
Maximum:
115' F dry bulb /800 F coincident wet bulb r
Minimum:
-40' F Ro - Pipe reactions during normal operating or shutdown conditions
.[
based on the most critical transient or steady state condition.
Appropriate dynamic load factors shall be used when applying-transient loads (such as water hammers)
+
4 7 1.2 Severe Environmental hoads Severe environmental loads are those loads that could infrequently be encountered during the life of the plant.
Included in this category.
will be Wind loads (W).
Operating Basis Earthquake -OBE-loads (E) are not being considered for the System 80+ Standard plant.
Loads generated by the design wind identified in CESSAR-DC W-section 3.3.1.
The wind is specified as being 110 mph at 30ft t 4 7
1
i March 8, 1993 above nominal ground elevation. Loads are calculated using ANSI A58.1-1982 (or ANSI /ASCE 7-88, dated 11-27-90) based upon a l
50 year recurrence period for non-safety related structures.
l Safety related Category 2 and 3 structures will use the same wind
]
For speed but with an assumed recurrence period of 100 years.
of 1.11 witb safety related structures use an importance factor (I) l an exposure level of "C" as defined in ANSI ASB.1(TBD7) i In accordance with ANSI A58.1, design wind pressure, p,
shall be calculated by the formula p = q G, C,, where q = velocity pressure
= 0.00256K, ( TV ) ', for V = 130 mph design wind speed l
5 G = gust response factor 5
i C, = external pressure coef ficient, dependent on shape of the b
structure
- See Attachment i for wind pressure coefficient distribution s
=
curves for the Shield Building dome and cylinder.
(Refer to Calc. file #4248-04-1622.00-0001 (later) for more details.)
Provide wind loads if possible based upon a table of applicable gust values from Table 8 of the above reference.
E - Loads generated by the Operating Basis Earthquake (OBE) with a peak ground acceleration of 0.1g.
t 7.1.3 Extreme Environmental Loads Extreme environmental loads are those which are credible but are highly improbable. They include:
E' - Loads generated by the Safe Shutdown Earthquake (SSE).
SSE =
0.30g shall be used as the peak ground acceleration. SSE loads
- 2B -
6
+
March 8, 1993 are obtained by multiplying the dead load by the appropriate structural acceluration obtained from the ABB-IMPELL seismic analysis. Amplification of these accelerations due to flexibility of structural members should be considered.
i SSE damping values used in design shall be as follows:
% of critical Damping Structure Type i
Welded Steel 4
Bolted Steel 7
Reinforced Concrete 7
containment vessel (later)
Masonry Block Walls (later)
Tank sloshing loads in the IRWST, refueling pool, and all other L
tanks shall be considered during an SSE.
W, - Loads generated by the design tornado specified for the plant.
Tornado loads include loads due to the tornado wind pressure (W,,), the tornado-created dif ferential pressure. (W,), and tornado-generated missiles (W,). See CESSAR Table 3.8.5 to t
-i combine these components.
L i
The following parzmeters from CESSAR-DC Table 2.0-1 shall-be used for the design basis tornados s
Maximum wind speed = 330 mph Maximum rotational speed = 260 mph i
Maximmn translational speed = 70 mph Radius of maxLsma rotational speed = 150 ft Maximum pressure drop = 2.4 psi Rate of pressure drop = 1.7 psi /sec,
i March 8, 1993 I
i i
Tornado wind loads shall be converted to wind pressure loads in i
accordance with ANSI A58.1, except velocity pressure shall be assumed constant with height (there fore K, ' = 1. 0 ), the importance factor shall be assumed as unity (I = 1.0), and the gust factor 4
shall be taken as unity (G = 1.0).
Therefore, tornado wind i
3 2
pressure loads will be computed by the formula p = 0.00256Vc,
p where the external pressure shape coefficient, C,,
is taken from
+
the distribution curves shown in Attachment 1.
i
'f Tornado missiles shall be considered in accordance with CESSAR-DC Table 3.5-2.
Design for missile impacts shall be in accordance-with CESSAR-DC Section 3.5.3 and ACI-349, Appendix C.
Minimum concrete wall and roof thicknesses shall be in ace cdance with
.f i
Standard Review Plan 3.5.3 Table 1.
Non-Category 1 structures shall not be assumed to shield the Nuclear Island from tornado s
wind, differential pressure, or missile loads.
i i
7.1.4 Abnormal Loads i
l Abnormal loads are those loads generated by a postulated high-energy pipe break accident. can there be other abnormal loads due to other system failures such as steam explosion due to RV failure. Included in this category are: Pressure loads ( P,), Thermal loads (T ),
Pipe reactions ( R,), Load on the structure generated by the reaction.on I
the pipe (y,), Jet impingement loads (y3), and Missile impact loads j
t (Y.).
l i
P, - Pressure equivalent static load within or across a compartment I
generated by the postulated break, and including an appropriate i
dynamic load factor to account for the dynamic nature of the load.
CESSAR-DC section 6.2.1.2 says that due to LBB there will be no I
+ ;
i
March 8, 1993 t
dynzmic effects for containment subcompartments?
i IK;fST pressure loads as a result of hydrogen burning, pool swell, vant clearing, vent impingement, or vent and goal chugging shall be treated as P. loads also, with the appropriate dynamic load factors applied.
I
?
T. - Thermal loads under thermal conditions generated by the
?
postulated break and including T,.
i R, - Pipe reactions under thermal conditions generated by the postulated break and including R.
i Y, - Equivalent static load on the structure generated by reaction on.
the broken high-energy pipe during the postulated break, and including an appropriate dynamic load factor to account for the dynzmic nature of the load.
In determining an appropriate equivalent static load for Y,,
l elasto-plastic behavior may be assumed with appropriate ductility ratios, provided excessive deflections will not result in loss of I
function of any safety related system.
t Y, - Jet impingement equivslent static load on a structure generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.
l i
In determining an appropriate equivalent static load for Y,
j 3
.i elast'o-plastic behavior may be assumed with appropriate ductility l
1 ratios, provided excessive deflections will not result in loss of I
function of any safety related system.
l q,
- March 8, 1993 Y, - Missile impact static load on the structure generated by or during the postulated breat, as from pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load.
Potential missiles are identified in CESSAR-DC Section 3.5.1 and P
Table 3.5-1.
In determining an appropriate equivalent static. load for Y.,
elasto-plastic behavior may be assumed with appropriate ductility ratios, provided excessive deflections will not result in loss of function of any safety related system.
7.1.5 OTHER LOADS
.I l
Construction Loads (later)
Potential loads'due to collapse of Category II Structures.
(TBD),
Will the Turbine Bldg have to be labeled as Cat. II to protect the
'I Nuclear Island?
Environmental effects (later)
(elevated temperatures, radiation exposure, chemical corrosion, etc.
t 72. DESIGN LOAD COMBINATIONS s
The following loading combinations, given in CESSAR-DC Table 3.8-5 and i
reference 6.5.3, shall be used for analysis and design of' Category I structures and their components for the SYS80+ Nuclear Generation Facility.
7.2.1 Loading Combinations for Category 1 Concrete Structures The following set of load combinations and allowable design limits.is.
I,
r p
~
i i
i
' March 8, 1993 i
1 used for all Category I concrete structures:
I Service Load Combinations represent Normal, Severe Environmental, and Normal / Severe Etvironmental loads.
For all combinations, Live Load must be considered both at its full value and absent.
U is the section strength required to resist the design loads and is I
determined using ACI 349-85.
I t
U = 1.4D + 1.7L
.I U = 1.2D + 1.7W U= 1. 4 D + 1. 4 F + 1. 7 L + 1. 7H + 1. 7W + 1. 7 R.
Note that "F"
"H" are combined in the CESSAR-DC "D" "L"
respectively i
Live Loads, including wind and other variable operating loads, must
.j be considered as being at either full value or absent for all load t
combinations.
the following If R, and-the thermal stresses due to T, are present, P
combinations representing a 1/3 increase in allowables shall apply
(.7 5 ) (1. 4D + 1. 4F + 1.7L + 1.7H + 1. 7T, + 1.7R )
U=
U=
(. 7 5 ) ( 1. 4D + 1. 4F + 1.7L + 1.7H + 1. 7W +1. 7T, + 1. 7 R.
P t
Factored Lead combinations represent; Extreme Environmental, l
Abnormal, and Abnormal / Extreme Environmental loads.
U = D + F + L + H + T. + R, + E '
[t U = D + F + L + H + T + R,
+W.
where W. = W, W,, W, (W, + 0. 5W,), (W + - W,), or
)
(W,, + 0. 5 W, + W,)
=
1 U = D + F'
' L + H'
+ T. + R,
!q where F' = F, ',
F,',
or (F,' + F,' )
1 U = D + F. + L + H + T. + R, + 1. 5 P.
+ Y,) ' + E '
U = D + F + L + H + T, + P + 1. 0 (Y, + Y 3
For the last 2 load combinations immediately above, the maximum t
? !
h t
March 8, 1993
)
values of P.,
T., Y, Y,, and Y, including an appropriate dynamic.
3 load factor, are used unless a time-history analysis is' performed to i
justify otherwise.
t The second factored load combination shall first be satisfied without the tornado missile load.
The last factored load combination shall first be satisfied without the Y loads. When including these loads however,-local section strength capacities may be exceeded under the ef fect of these concentrated loads, provided there will be no loss of function of any safety related system.
creep, or Where the structural effects of differential settlement, shrinkage may be significant, they shculd be included with the dead
load, D,
as applicable.
Where any load reduces the effects of other loads, the corresponding coefficient for that load should be taken as 0.9 if it can be i
demonstrated that the load is always present or occurs simultaneously with other loads. Otherwise the coefficient for that load should be taken as zero.
-)
I
-[
7.2.2 Loading combinations for Category 1 Steel structures The following set of load combinations and allowable design limits is j
used for all Category I steel structures, including individual i
k support members:
Service Lead Conditions -
If elastic allowable strength design methods are used:
S is the required section strength based on the elastic design i
methods and allowable stresses defined in Part 1 of ANSI /AISC N690-i 1992
),
P f
March 8, 1993
?
S=D+L i
S=D+L+W If thermal stresses due to T, and R, are present, the following combinations are also satisfied:
1. 5 S = D + L + R, + T.
1. 5 S = D + L + W + R, + T, t
i If plastic design methods are used:
Y is the section strength, based on the plastic design methods, r
required to resist design loads using allowable stresses defined in Part 2 of ANSI /AISC N690-1992 Y = 1.7 (D + L) j Y = 1.7 (D + L + W)
Y = 1.3 ( D + L + T, + R )
Y = 1.3 ( D + L + W + T. + R,)
l 1
Factored Load Conditions If elastic allowable strength design methods are used:
l
1. 6 S = D + L + - R, + T + E '
1. 6 S = D + L + R, + T + W,
-(See concrete note Re W)
'i
1. 6 S = D + L + R + T + F '
l (See concrete note Re F')
i (flooding, F', is not in CESSAR for steel structures)_
1. 6 S = D + L + R, + T. + P '
t
1. 7 S = D + L + R, + T + Y, + Y. + Y.. + E ' + P.
f; s
'l (The plastic section modulus for steel shapes may be used for f
this load combination.)
i
. r.
3 If plastic design methods are used:
Y' = 1. 0 ( D + L + R, + T, + E ' )
Y' = 1. 0 ( D + L.+ R, + T. + W,)
i l
r f
i
' I March 8, 1993 (See concrete note Re W,)
Y' = 1. 0 ( D + L + R, + T + F ' )
(See concrete note Ret F')
Y* = 1. 0 ( D + L + R, + T + 1. 2 5 P. )
+ Y, + ' E ' + P. )
Y* = 1. 0 ( D + L + R, + T. + Y, + Y3
- per CESSAR Table 3.8.5, multiply "Y" by 0.90 for. Internal Structures
.s 7.2.3 Loading combinations for Stability checks 7.2.4-Applicability of Loads
(((Add a table for structures and applicable loads with a reference included here)))
Design loads and loading combinations for the containment Internal is provided in reference 6.0.11 and CESSAR Section 3.8.3.3 Structures and CESSAR Table 3.8-5.
For the Shield Building and the Nuclear Annex Structures see reference 6.0.11 and CESSAR Section 3.8.4.3.
Lateral loads due soil bearing pressure shall apply to all exterior walls up to El. 908-9" (pending). Refer to drawing (later) for the Nuclear Island perimeter subject to soil loads.
{
Tornado loads are. applicable to the Shield Building above El. 156.. and to the roofs and exterior walls of the Nuclear Annex.
Some Interior walls of the Nuclear Annex have been designated as pressure i
boundaries when required boundaries cannot reasonably be obtained at the exterior walls. ' Refer to Drawing (s) (later) through (later) for the location of these missile and pressure boundaries.
I Hydrostatic forces are applicable to all Exterior walls of the I
Nuclear Island up to elevation 89'-9".
Hydrostatic uplift forces are i
i L
March 8, 1993 t
-applicable to the entire Nuclear Island basemat.
1 7.2.5 Notes and Definitions (later) r 8.0 Structural Analysis and Design
'i ABB-CE has committed to conform with the criteria included in-the EPRI ALWR Program for the design of the System 80+ Standard Plant (DSER Sec 1.1) 8.1 Analysis Identify the reinforcing used in the model and be prepoared to address
.t differences in the as modeled reinforcing and the as designed.
0.1.1 General All Category I Structures forming the Nuclear Island,. identified in supported on a common reinforced concrate.
Sections 2.0 and 3.0, are
+
basemat foundation. In addition these structures share interconnecting reinforced concrete structural components such as j
floors and walls. The integral construction of the Nuclear Island requires a single combined analysis of the Nuclear Island components.
1 The analysis will be done in two phases.
Phase'1 is the seismic analysis.
Phase 2 will be the structural analysis based upon the seismic accelerations generated in Phase 1.
t The seismic and structural analyses for outlying Category I structures will be addressed in Appendix A.
Non-nuclear safety.
l related structures will require only a static structural analysis'and-will be addressed in Appendix B.
8.1.2 S.emic Analysis The seismic analysis will determine the response of Category I
'i '
March 8, 1993 structures to the horizontal and vertical movements produced by the Safe Shutdown Earthquake (SSE).
This analysis.will generate transfer functions or response spectra for multiple nodes at selected elevations or floor levels on.the Nuclear Island. TheseL transfer functions / response spectra are then used to predict-5 seismic loads for major equipment, i.e. the NSSS components and associated piping, as well as other components mounted on the i
Nuclear Island. The seismic analysis results will consider ~
multiple potential soil profiles under the Nuclear Island basemat.
8.1.3 Structural Analysis The structural analysis will provide the final'informatio'n required for the design of the structural components on the Nuclear
'I Island. The Nuclear Is3and will be analyzed using PC computer based SAP 90 software. SAP 90 is sLmilar to main frame based ANSYS.
f in data input and analysis output.
The model used for the structural analysis will consist of 3 dimensional finite element.
I These elements will allow modeling of walls,' floors and' roofs as 2-way slabs. The basemat finite elements, modeled on soll spring supports, will allow conside ration of the foundation mat l
1 flexibility.
l
')
The structural analysis computer model will consist of a common reinforced concrete mat foundation supporting interconnecting walls
[
and floor slabs. The computer model, created to generate General r
Arrangement drawings, will be used by the' SAP 90 program to generate properties for elements / members representing reinforced concrete structural components. Elements will consider concrete to i
have 4000 psi strength in the basemat and 5000 psi elsewhere on the' Nuclear Island. A limited number of steel members will also be included.
j i
r
March 8, 1993 4
The structural analysis will consider those static and dynamic loads and load combinations identified in Sections 8.0 and 9.0.
l All three components of earthquake loads (two horizontal and one i
vertical) shall be considered simultaneously. Uniform loads representing expected fixed equipment and equivalent uniform live loads were identified in reference 6.5.3. Specific loads due to P
NSSS piping and equipment were also identified in reference 6.5.3.
The SAP 90 program will generate dead loads internally due to the mass of identified structural members. These loads are then combined by the structural analyst with loads described above according to the required load ecmbinations.
i Analysis results consisting of forces and moments will be used for preliminary design of walls, slabs, beams, columns, and basemat.
f Miscellaneous loads due to the attachment of equipment or portions of electrical or mechanical distribution systems will be addressed for localized effects on a case by case basis.
i 0.2 Structural Design l
8.2.1 General Requirements Analysis and design of the Nuclear Island Category 1 concrete Shield Building Internal Structures, containment Shield Building, Containment sub-sphere structures, and foundation basemat shall be by the Strength Design Method. These are spelled out in codes ACI 318-89 and ACI 349-85.
Although ACI 349-85 is identified in CESSAR-DC as the design basis, the actual structural design will use the more stringent of the requirements of ACI 318 and ACI 349. I
March 8, 1993 ACI 318 Chapter 21 shalil be used in the design for seismic loads because the design requirements are currently spelled out there in more detail. Design for impulsive and impactive effects shall be in accordance with ACI 349-85, Appendix C (later).
'i l'
ACI 318-89 may be used for non-safety related applications on the Nuclear Island with appropriate QA documentation. The ASME Code will
+
not apply to these concrete structures since none of them are intended to serve as pressure retention boundaries, tornado air i
pressure drops being excluded, nor do any of the structures constitute a concrete containment or reactor vessel.
Overall physical dimensions for Nuclear Generation Facility structures, as well as specific dimensions for most concrete components, have been established on General Arrangement drawings.
Designs shall proceed with that info and analyze loads to determine suitable reinforcing or proper structural steel cross section.
The design of Reactor Building Category 1 steel structures and/or components, excluding the Containment Vessel, shall use Allowable i
Strength design methods in accordance with ANSI /AISC N690-(later).
The design of the Nuclear Island will be based upon forces and l
t moments generated by the integrated analysis. This analysis was done using the loads and loading combinations previously identified in i
Sections 8.0,' 9.0 and 11.1.3 of this criteria. How will we address live loads on specific floors or individual equipment load when we I
design for loads produced by the IMPELL analysis.
[
t No increase in allowable stresses for concrete or steel will be permitted due to operating basis seismic or normal wind loads.
Need clarification of CESSAR-DC Section 3.8.4.5.
Restriction is March 8, 1993 placed on seismic and wind when it should not apply to SSE and tornado winds. Note 1 in SRP following Section 3.8.4 II.5 is marked for concrete WSD only even though wording is for concrete and steel.
t With suitable qualification and no applicable material restrictions, substitute materials may be used.
t Exterior walls and roof slabs designated on drawing (later) shall i
be designed as missile barriers for tornado and any other designated missiles. Designs shall assure that the structure will:
not collapse under the missile load nor will there be penetration through the wall or roof slab.
8.2.1.1 Fire Protection Fire Barriers (later) 8.2.1.2 Flooding Nuclear Annex shall hav-s no unsealed interior wall or floor penetrations up to plant flood level, El.
89'-9".
8.2.1.3 OSHA (later) 8.2.2 Design Procedures 8.2.2.1 Basemat The basemat will be designed for the maximum forces and moments provided by the SAP 90 analysis. The concrete mat will be provided with flexural reinforcing based upon these results but not less l
t than the minimimum reinforcing required by Section 10.5 of ACI 349.
l The basemat will be designed for a construction sequence (later) j i
which will expedite the initiation of the erection of the.
)
J 1 l
L i
March 8, 1993 1
containment vessel. Design of the basemat should consider stresses due to pouring sequence of the mat as well as the erection sequence for components located above the nat.
Pours will be laid out based upon a target pour size of 2000 cu-yds.
Pour layouts should minimize skewed intersection of construction joints with walls due to conflicts in placement of wall dowels. Construction joints shall be located on appropriate design drawings. (later) i Shrinkage cracks in the exposed faces of massive concrete pours, t
~
such as those for basemat shall be controlled by minimum reinforcing as specifed in ACI 349 Section 7.12.
This reinforcing required by ACI 349 section 7.12 shall apply (TBD)_to temporary e
exposed faces such as interior construction joints.
.[
i Sumps (later) 8.2.2.2 containment Shield Building 8.2.2.3 Shield Building Sub-Structures (later) address shear connections between the containment and the Iower concrete support pedestel.
9 8.2.2.4 containment Internal Structures
[
The Internal Structures are described in CESSAR-DC Section 3.8.3.1.
{
The applicaable load categories are given in Section 3.8.3.3.
i 8.2.2.4.1 Reactor Vessel Primary Fhield Wall t
The Primary Shield Wall is the primary internal support for the j
i Reactor Vessel. With a minimmm wall thickness of 6 feet the Primary Shield Wall will be thick enough to protect the Reactor vessel from all potential missiles and also sufficiently thick to protect workers from excessive radiation.
The inner face of the lower Primary Shield Wall will be provided with projecting reinforced concrete corbels to be used as the,
i 4
March 8, 1993 support bases for the Reactor vessel steel support Columns.
The Primary Shield Wall will be designed.for normal dead loads as well as equipment live loads and related seismic forces.
In addition the PSW will resist the dynamic loads due to the NSSS components. The dynamic effects due to a low probability-accident involving a steam explosion produced by the failure of the Reactor Vessel and the collapse of the Reactor Core on the Containment Sump will also be considered.
A worse case accident involving the failure of the Reactor Vessel and collapse of the reactor core onto the Containment ~ sump floor will create potential additional dynamic loadsreated by Refer to i
reference calculation (later) for special design qualifications for the Prbmary Shield Wall as well as the support corbels.
Will there be any ex-core radiation detectors in the primary i
Shield wall?
l Will there he any loads due to potential CRD missiles hitting the r
missile shield 8.2.2.4.2 Stems Generator Enclosures (later) provide load info for steam generators 8.2.2.4.3 Crane Wall (Secondary Shield Wall)
I (later) crane design requirements dead loads design lifting capacity i
crane rail info with hold down requirements can we reference a design spec for the crane 8.2.2.4.4 Refueling Canal The Refueling Canal is the reinforced concrete enclosure that i
surrounds a pool of borated water above the reactor vessel. The water provides shielding from irradiated reactor vessel components as well as nuclear fuel assemblies during refueling fuel transfers. l i
i
...-e.
March 8, 1993 The reinforced concrete enclosure wall shall be 6 feet thick.
The design of the Refueling Canal will incorporate details for the Fuel Transfer Tube, Core Support Barrel Laydown Area, Upper Guide Structure Laydown Area, and a waterproof seal around the Reactor Vessel flange where it drops through the Refueling Canal i
floor.
The pool sides of these walls will be covered with stainless steel plate (type ????).
Welded joints in cover plates will be leak chased to allow detection of damaged or defective welds.
f-8.2.2.4.5 Opetrating Floor B.2.2.4.4 In-containment Refueling Water Storage Tank 8.2.2.4.7 Lower Concreto Dish (later)Will there be any connections between containment and the dish concrete?
8.2.2.4.8 Fuel Transfer Tube 8.2.2.5 Nuclear Annex i
l 8.2.2.5.1 Diesel Generator Rooms 8.2.2.5.2 Control Room 8.2.2.5.3 Exterior Valve House 8.2.2.5.4 Spent Fuel Pool 8.2.2.5.5 8.2.2.5.6 8.2.2.5.7 8.2.2.6 8.2.3 special Design Criteria f
j 8.2.3.1 seals Penetrations on the Nuclear Island will be provided with suitable seals to meet flood, interior or exterior sources, protection requirements, or with pressure seals adequate for support of the -
1 March 8, 1993 leakrate requirements (later). Floods may result from int erior or exterior water sources.
8.2.3.2 Grout (later) here or in section 9.1 8.2.3.3 Bearing pads special bearing pads will be required on both faces of the Containment vessel, for its full perimeter, at El.
91'-9".
These pads will be designed to optimize localized stresses between the Containment Vessel plate and the adjacent concrete slabs.
Requirements for these pads will be identified in reference calculation (later) 8.2.3.4 Block Walls n
(later) Will we be using them?
P 8.2.3.5 others (later) 9.0 Construction; Forming, Fabrication, and Erection
{
+
9.1 Concrete 9.1.1 Procurement and Placement
'i (later) height of concrete pours (later) Do we want to identify rate concrete is to be placed?
i (later) 9.1.2 Reinforcing (later) do we want to specify or recommend extended height of l
?
reinforcing?
(later) Ehear friction?
Placing of reinforcing shall conform to those tolerances found in
[
ACI 349 Section 7.5 Reinforcing will be detailed for full lap splices for most installation. Construction will be provided with the option to use i
March 8, 1993 the lap splice or substitute CADWELDS or other approved mechanical spicing devices. Lap splices will be prohibited for bar sizes greater than #11 except as provide by ACI 349 Section 12.14.2.1.
9.1.3 Construction Joints'(later) e (forming, stripping, or cleaning ?)
Shear keys ?
9.1.4 Tolerances 3
(later)
CONSTRUCTION TOLERANCES, REBAR PLACEMENT PER ACI 301 9.1.5 Curing methods for concrete pours (later)_
9.1.6 Controlling Internal Tamperatures due to mass of concrete pour or due to hot weather (later)
~
9.2 Steel 9
9.2.1 Structural Steel; fabrication and erection Reference (later) 5.2.2 Metal Decking Reference (later) 9.3 Structural Attachments 9.3.1 Fabricated Embedments a Typical Embedded plates Reference (later) b Special Equipment Supports Reference (later) equip mntg c Placement setting of embedments prior to concrete pours and monitoring their position during the pour
.r 9.3.2 Expansion Anchors e
I ACI349-85 Appendix B required safety factor; allowable loads by Brand; acceptable brand,
^
only undercut anchors will be acceptable (TBD). no wedges, sleeves or self drills?
f,
l
a L
i March 8, 1993 t
h Reference (later) 9.4 Installation of Equipment and Misc. Components (later) 9.5 Attachments to Containment vessel Requirements (later) 10.0 STRUCTURAL ACCEPTANCE CRITERIA i
10.1 General 10.1.1 Structural Acceptance criteria spelled out in CESSAR-DC Section 3.8.4.5 provides that the nuclear safety-related components of this Nuclear Generation Facility shall be deemed to be acceptable if they.
meet the design requirements set forth in this criteria and its references and in addition comply with quality assurance requirements applicable to construction.
10.1.2 Construction Quality control files shall document all deviations from approved drawings for safety related Category I or II structures and components as well as the engineered resolution. To e
be acceptable, final construction details must comply with the applicable engineering drawings, except where approved by appropriate Quality control documents, and the System 80+ Standard Plant Structures ITAAC.
10.1.3 Separation of category I and non-Category I structures shall be verified to assure no adverse seismic interaction of the structures.
10.1.4 The site specific SSE and soil profile must be enveloped by the conditions and design values used in the qualifying analysis.
10.2 Concrete 10.2.1 Concrete design shall be deemed acceptable as long as.
5 components of each structural element satisfy the requirements of ACI 318-IX, ACI-349-XX, or ASME Code Section III Division 2 Sub-Section CC as appropriate.
10.2.2 Design cross sections with their respective reinforcing will be compared with member sizes and reinforcing anticipated in the
-w
March 8, 1993' analysis mode.. Where member sizes or the amount of reinforcing has been changed significantly (later), a seismic and/or structural reanalysis will be performed to consider the effects of the revised structural stiffness, i
e i
Steel Structural steel members (beams and columns) and elements (plates and fabricated components) shall be deemed acceptable as long as each structural member or element satisfies the applicable requirements of the current version of ANSI /AISC N690, 1984 or 1992(draft). AISC
" SPECIFICATION for the DESIGN FABRICATION AND ERECTION of STRUCTURAL STEEL FOR BUILDINGS." (EFFECTIVE June 1, 1989) will apply to all situations not addressed by ANSI N690.
The forces and moments on the lower concrete dish supporting the containment and the internal concrete dish slab resting within the containment will be checked to verify stability of the containment Vessel. Overall stability will be checked against overturning, sliding, and flotation. A preliminary evaluation was performed.and documented in DE&S calculation 4248-04-1622.00-0002.
An ad hoc evaluation must be performed for each site specific SAR.- Thel required safety factors are given in Section IV of CESSAR-DC Table 3.8-5.
The following areas are identified in the Section 3.8.5 of the Standard Review Plan and Section 3.8.4.5 of CESSAR to be checked as a mintswas-
1. Overturning of the Nuclear Island about the edge of the basemat when supported on soil.
f
2. Sliding of the Nuclear Island basemat on soil.
i
3. Floating of the Nuclear Island j
Slipping of the Containment Vessel within the lower concrete support 4.
dish.
5. Overturning of the Containment vessel about the edge of the lower,
v
' March 8, 1993 concrete support dish.
6. Slipping of the Interior Structure concrete inside the Containment Vessel.
Acceptability of natural material under the Nuclear Island Basamat and the outlying Category I structures is site specific information and will-be address in the site specific SAR.
l 11.0 Materials The following materials will be asssumed for preliminary analysis and design of Category I structures:
Cament - material shall conform to ASTM C 150 per ACI 349 para 3.2 Aggregates - material shall conform to ASTM C 33 per ACI 349 para 3.3 Concrete - f'c = 5000 psi (4000 psi for the basemat) the design mix shall be as documented in specification (later)
Reinforcing Steel - Ab?M A615 Grade 60, Fy = 60,000 psi
[
or
- ASTM A706 Fy = 60,000 psi l
ASTM A615 Supplemental requirements S1 will apply (TBD). Compare this with Chapter 21 of ACI 318-89 where the ratio of ultimate tensile to actual yield must be less than 1.25 and actual yield must be no more than 1.3*specified yield (based upon Fy=60ksi.
ASTM A706 shall be required for reinforcing that must be welded.
(excluding CADWELDS)
Reinforcing with a minimmn yield strength greater than 60ksi is not 1
permitted per ACI 349 Para. 9.4
'l 4
Structural Steel (excluding round & tubular shapes) - Fy = 36,000 psi Structural Steel Tube Shapes - A500 Grade B, Fy = 42,000 pai Rebar Splice Systems (later) l.]
i i
~
. I
March 8, 1993 Mechanical reinforcing coupler devices, similar to those produced by DYWIDAG Systems International, will be acceptable upon receipt of approved QA certification. CADRELD coupler devices will be acceptable f
with proper QA certification but their use should be kept to a minimum i
due requirements for QA qualification of each splice.
Metal Decking J1ater)
Corrugated metal decking'used for supporting concrete slabs will be
[
k acceptable when the composite slab b qualified for all applicable load combinations and potential missiles.
Selection of decking will be part of the final design criteria Expanded Metal Mesh (later).
5 It is postulated that this material will be used for all exposed edges Galvanized materials and aluminum (later)
Waterstops around sumps (later)
Waterstops through regular construction joints (later)
Compressible material between the Steel containment vessel and both the i
Shield Bldg concrete slab and the Internal Structures base slab at E1. 91'-9" J1ater) CESSAR-DC section 3.8.7.1.2 requires compressible materials around all sub-sphere penetrations. This is based upon assumption that the plate could move between 2 layers of concrete.
Movement would be a shear induced displacement rather than axial.
(TBDr Compressible material is not' labeled or material specified on Figure 3.8.2 of CESSAR-DC additional materials may be used subject to qualification by the appropriate ASTM standards and NRC approval.
3 6
s P
l i
.?
March 8, 1993 Appendix A f
Appendix B Appendix C Tablo C-1 Summary of Applicable structural Loads (later)
Table C-2 Applicable Live Loads by Area (later) t Appendix D Figure D-1 Shear reinforcing-will the computer analysis give us results that we can readily interpret as shear forces on the joint. If edges of pours don't line
~
up with the analysis, how are we going to develope the shear loads to qualify the construction joints???
Minimum cover water stops rebar spices; lap.vs. mechanical couplers or CADWELDS expanded metal forms for edge of pours use of shear keys, ARE THEY PERMITTED Straight embedments for dowels or use hooks Use reduced concrete strength of 4000 psi for basemat only i
Walls walls & Floors shall be designed to prevent-missile penetration structural co11 apes or potential scabbing of the concrete surface.
-t target maximum height of pours Reactor Vessel Primary Shield wall how will this and the other interior structures be anchored through the containment vessel. Containment support Pedestal Method of design for curbs & flood doors associated with ITTAC pipe break-.
flooding.
-)
seals for subsphere quadrant penetrations.
ACI 349 requires positive or welded splices in elements with l
membrane tension stresses.
This could be the shield blog and r,
t e
. 1
. March 8, 1993
.possibly the crane wall
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l DUKE ENGINEERING
& SERVICES, INC DMDE Bus PD4) 373-2473 C/$:820i.iu
" F d) 73'"885 i
ALWR-544 January 29,1993 Mr. Sohrab Esfandiari ABB-Impell Corporation 5000 Executive Parlavay PO Box 5013 San Ramon, CA 94583
Subject:
System 80+ Design Certification Civil Structural Design Details Stmetural Analysis - Nuclear Island-Global Loads and Combinations File: 4248-00-1622.00
Dear Mr. Esfandiari:
i In support of ABB-Impell's structural analysis of the Nuclear Island, DE&S is to provide loads and load combinations by January 30,1993. The purpose of this letter is to transmit i
this information.
l The loading information is discussed and presented as follows:
i Loads and Load combinations:
Loads and combinations are presented in CESSAR-DC Chapters 2 and 3. Attachment 1 provides tables and sections from the CESSAR-DC, which summarize this information. Note that the copy of Table 3.3-5, included in Attachment I, has been marked to delete-l consideration of OBE.
It should also be noted that the uniform live and equipment loads provided below represent global loading information for use only in the overall structural analysis. These uniform loads are " smeared" loads, applicable to the areas shown. 'Ile unifonn design load for an individual member (local area) may be more or less than provided herein.
i Dead Load:
Stmeture dead loads are not provided as they will be internally generated by the structural -
l analysis program. 'Ihe water levels in pools and tanks (ie. SF Pool, ERVT) are noted on
i
' Mr. Sohrab Esfandiari a
January 29,1993 Page 2 Live Load:
j Uniform live loads are provided in Attachment 2. Significant temporary loads to accommodate 1 piece steam generator removal are shown in Attachment 3.
]
Eauipment Loads Uniform equipment loads are provided in Attachment 2. Weights of equipment, which should'
~
be considered in addition to the uniform loads are provided in Attachment 4.
I F
Maior Pipine Loads:
f
' Anchor loads on the SB at the MSL and EFW penetations and nipture restraint loads due to -
MSL and EFW pipe breaks in the MSVH are provided in Attachment 5.-
l RCS loads are provided on Attachment 6.
l t
r Wind \\Tomado Loads:
Resulting pressures on the structures due to the extreme and tornado wind speed is provided in Attachment 7. The tornado pressum differentialloading is specified in Attachment 1.
j j
Soil:
Maximum groundwater and flood levels are provided in Attachment-1. Static lateral earth pressures are local loads, thus, not provided. The dynamic loads on the external walls, including effects of adjacent structures, due to the SSE will be generated by ABB-Impell.
'l
'I Seismic:
Seismic loading information is not provided; ABB-Impell has completed this analysis.
r
~ Mr. Sohrab Esfandiari January 29,1993 Page 3.
If you have any questions, please call Todd Oswald at (704) 382-2831 or Bill Fox at (704) 373-5020. Todd and other DE&S support will be in your offices on Febmary 2,1993 to discuss loads and review the model.
Sincerely, 1
l R.W Bonsall, Vice President Advanced Nuclear Programs 4
JTO/ksl/TB.001 Attachments C
xc:
L.D. Gerdes (ABB-CE) j R. S. Turk (ABB-CE) w/o att.
J.T. Oswald W.A. Fox
[
T.L. Bradley H.E. DeMart R.C. Abernethy D.C. Mottern i
G.L Green J.A. Johnson D.R. Ingle '
i P.B. Pettie D.B. Fisher Project File Central Records i
"g
4 1
2 2:
18Btf 3.2-4
$UMMARY OF CRITERIA - STRUCTURES l
Loeding Normal Dead and Stearks Structures Contelnment Selsmic fornado
. Including Any
. Wind Eculpment (lvt eccident Pressure 08[ 11( Wind Missile Environmental Reautremente Contelnment X
X X
X X
Containment Interior Concrete thermal Stresses; Equipment Hiselle Protected X
X X
X Dif ferentlet Accident Pressure; Pipe Rugture Loads; therest Stresses; Shield Bullding including foundation X
X X
Equipment Mleslie Protected
^
X X'
X X
fornedo Pressure Olfferentist; Containment Penetrations X
X X
X.
X Soll and Water Pressure Centelnment Structuret Steel Pipe Whipping X
X Underground Cable systems X
X Ihermet Stresses X
X X
X X Station Service Water Pipe X
X X
X X
X X
Soll and Water Pressures or Burled Portion Station Service Water Structures including Hydraulle Pressures; Moving Equipment Loade X'
X X
Pump Structure X
X X
X Solt and Water' Pressures;
' Intake & Discharge Structure-Selsmic Analysts Dem Spillway Control Structure Aunillary guilding Including X
X X
Control guilding X-
-X-
-X X.
_ Salt and Water Pressures on Substructure; Diesel Building tornado Pressure Orop; ruel Pool thermal Stresses and Cesk Drop;L Main Steam and feedwater Enclosures Pipe and Plpe Rupture Londe; fuel Storage Rocks Station test X
X Thermal Stresses X
X X
X X
X X
na s
Amendment I
.... 2..
,u, _
.a
CESSAR Pair,cy,2 m1-mr
.,x TABLE 2.0-1 (Sheet I of 3)
ENVELOPE OF PURT SITE DESIGN PARAMETERS 1
Ground Water Maximum Level:
2 feet below grade II)
Flood (or Tsunami) L'evel Maximum Level:
I foot below grade Precipitation (for Roof Design)
Maximum rainfall rate:
19.4 in/hr. and 6.2 in/5 min.I2)
Maximum snow load:
50 lb/sq. ft.
Design Temperatures t
Ambient 1% Exceedance Values J
Maximum:
100*F dry. bulb /77*F coincident i
wet -bulb l
80*F wet bulb (non-coincident)
Minimum:
-10*F 0% Exceedance Values (Historical Limit excluding' peaks-<' 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).
1 Maximum:
115'F dry bulb /80*F coincident wet bulb 81*F wet bulb (non-coincident)
Minimum:
-40*F I3)
Station Service Water Inlet:
95*F Condenser Cooling Water Inlet:
sIO0*F O
l
Att, 11--3/17
CESSAR Heinc,now i
TABLE 2.0-1 (Cont'd)
(Sheet 2 of 3)-
ENVELOPE OF PLANT SITE DESIGN PARAMETERS Extreme Wind 130 Basic Wind Speed:
110 mnh (Ref. Sec. 3.3.1.1) 1.0W/1.11(5)
Importance Factors:
Tornado (6)
Maximum tornado wind speed:
330 mph Rotational Speed:
260 mph Translational velocity:
70 mph 150 ft Radius:
Maximum pressure differential:
2.4 psi Rate of pressure drop:
1.7 psi /sec Missile spectra:
per SRP 3.5.1.4 Spectrum 11 Soil Properties Minimum Bearing Capacity (demand):
15ksf(styc)
Minimum Shear Wave Velocity:
500 ft/sec Liquefaction Potential:
None (at site-specific SSE level);
3 Seismology OBE Peak Ground Acceleration (PGA):
0.10 g SSE PGA:
0.30 g SSE Response Spectra:
Section 3.7.1 SSE Time History:
Section 3.7.1 ROIES:
1.
Probable maximum flood level (PMF),
as defined in ANSI /ANS-2.8,
" Determining Design Basis Flooding at Power Reactor. Sites."
Maximum value for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 sq. mile PMP with ratio of 5 minutes to I hour 2.
PMP of.32 as found in National Weather Service Publication HMR No. 52.
3.
Maximum normal power and normal shutdown temperature of the Station Service Water System Intake based on one percent exceedance meteorologic conditions.
See item C of Section 9.2.5.1.3 for Ultimate Heat Sink.
temperature interface requirement for a design basis accident concurrent with a-loss-of-offsite power.
Amendment J nnvil
,n.
too>
)
1 Att.
1.- 4/17 C{w&MK1 CERTIFICATI2N j
C C A E3 E;ESIGN
1 l
l TABLE 2.0-1 (Cont'd)
(Sheet 3 of 3) i ENVELOPE OF PUWT SITE DESIGN PARAMETERS NOTES: (Cont'd) 4.
50-year recurrence interval; value to be utilized for design of-non-safety-related structures only.
5.
100-year recurrence interval; value to be utilized for design ofL safety-related structures only.
J' 6.
10,000,000-year tornado recurrence interval, with associated parameters based on the NRC's interim position on Regulatory Guide 1.76.
Pressure l
effects associated with potential offsite explosions are assumed 'to be-1 non-controlling for the design.
l 7.
Site profiles are given in Section 2.5.
8.
The control motions are defined in Section 2.5.
t 4
.i 1
i i
s f
i b
c.
Fh*~
PE"C e A D Cu!GN Att. 1 -:5/17 w E*w G M EW CD TIF1::ATioN l.
/
TABLE 3.8-51 i
s L(Sheet 1 of'10)'
1 LOAD COMBINATIONS FOR CATEGORY I STRUCTURES -
,i INDEX I
I.
Load Definitions.
1.
Normal Loads 2.
Severe Environmental Loads 3.
Extreme Environmental Loads j,
4.
Abnormal Loads -
j 5.
Other' Definitions j
II.
Load Combinations and Acceptance Criteria for CategoryL I Concrete:
Structures l
1.
Service Load Conditions l
2.
Factored Load Conditions III. Load Combinations"and Acceptance Criteria for. Category'I Steel. Structures?
1.
Service Load Conditions
-i a.
Elastic Design b.
Plastic Design s
2.
Factored Load Conditions i
a.
Elastic Design b.
Plastic Design i
IV.
Load Combinations and Acceptance Criteria for Category I Foundations-t i
[l
~
l l
l
'i
i
^"'
~""
CESSAR 88Hncamu
('
TABLE 3.8-5 (Cont'd)
(Sheet 2 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES I.
Load Definitions All the major loads to be encountered and/or to be postulated in a Category I structure are grouped into four categories described below.
All the loads listed, however, are not necessarily applicable to all the structures and their elements in the plant.
Loads and the applicable load combinations for which each structure is designed will depend on the conditions te which that particular structure could be subjected.
1.
Normal Loads Normal loads are those loads to be encountered during normal plant I
operation and shutdown.
They include the following:
D
--- Dead loads or their related internal moments and forces, including any permanent equipment loads and hydrostatic loads L
Live loads or their related internal moments and forces, including any movable equipment loads and other loads which vary with intensity and' occurrence, such as' soil i
pressure F'
Buoyant force of probable maximum flood Thermal effects and loads during normal operating or T
o shutdown conditions, based on the most critical transient or steady state condition Pipe reactions during normal operating or shutdown R
conditions, based on the most critical transient or steady o
state condition 2.
Severe Environmental Loads Severe environmental loads are those loads that could infrequently be encountered during the plant life.
Included in this category are:
E LtadS1 6HErat6d by ths Opt: a i. s a 4g 3ddis Ldr i.liqudkt:
W Loads generated by the design wind specified for the plant
Att.'l - 7/17, l
CESSARianncame l
l i
TABLE 3.8-5 (Cont'd).
i (Sheet 3 of 10).
)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES r
3.
Extreme Environmental Loads l
Extreme environmental loads are those which are credible ~ butt are highly improbable. :They-include:
i.
Loads generated by the Safe Shutdown Earthquake-E' l
Loads generated by.the Design Basis Tornado specified for W
t the plant.
They include loads due' to the tornado wind.
pressure (W
loads due to the tornado-created di fferential,) pressures (W ),
and loads due-to. the-tornado-generated missiles
,)
j I
i
-}
The combined effect-of W,,
W,
'and W
is determined-in a
conservative manner for each harticular, structure or portion l
thereof, as applicable, by using one or more of the following
-l combinations as appropriate:
[
(i)
Wt" w
(ii)
Wt"Wp (iii)
Wg = W, (iv)
Wt"w +0.5W w
(v)
W
+
t" w
m t"
w + 0.5 Wp + W, (vi)
W 4.
Abnormal Loads Abnormal loads are those loads generated by a postulated high energy-pipe break accident within a building and/or compartment -thereof.
i Included in tnis category are the following
}
Pressure equivalent static load within or across a
P, compartment and/or building, generated by.the postulated break, and including an appropriate dynamic load' factor to account for the dynamic nature of the load a
I
iCESSAR naincme,,
Au. r-anr a
TABLE 3.8-5 (Cont'd):
l (Sheet 4_ of 10) l LOAD COMBINATIONS FOR CATEGORY-I STRUCTURES-Thermal loads -under thermal conditions generated by the l
T, postulated break and including T o
pipe reactions under thermal conditions generated by the:
R 3
postulated break and including R, Y
Equivalent static load on' the structure generate'd by the reaction on the broken high-energy pipe. - during -l the -
r postulated break, and including an. appropriate dynamic load factor to account for the dynamic. nature _of-the load
?
Jet impingement _ equivalent. static load _ on.' a structure j
Y.
generated by the -postulated -break, and including 'an-3
. appropriate dynamic load factor to account for the dynamic-- 1:.
nature of the load-Missile impact equivalent static-lodd on a. structure' i
Y*
generated by or. during the postulated break,7 such as-pipe 1
whipping, and ! including an appropriate dynamic load factor to account for the dynamic nature of the load In determining an appropriate equivalent static-load' for Yr, Yj,_ and '
Ym, elastic-plastic behavior may be-assumed with appropriate'
)
ductility ratios as 1ong as. excessive deflections will-not result in 1
loss of function of any safety related system.
5.
Other Definitions For structural steel, 'S is the required 'section' strength S
based on the elastic design methods. and' the allowable:
stresses defined in ANSI /AISC _N690-1984 For concrete structures, U
is the section strength U
required to resist design loads ~ based on; the ultimate strength design method described in' ACI 349-85 Y
For structural steel,: Y is the 'section: strength required to resist design loads based on ' plastic: design ' methods '
described in ' ANSI /AISC N690-1984~
Amendment I~
n---_u--
,enn
^tt 9/17 CESSAR Eanticuios 1
<m i
TABLE 3.8-5 (Cont'd)
(Sheet 5 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES II.
Load Combinations and Acceptance Criteria' for Category I Concrete i
Structures The following set of load combinations and allowable design limits is used for all Category I concrete structures-1.
Service load Conditions Service Load Conditions, represent Normal, Severe Environmental and Normal / Severe Environmental loads.
The Ultimate Strength Design method is used with the following load combinations:
1)
U = 1. 4 D ' + 1.7 L 2)
U -
1.'
O 1.7 L 1.9 E-D 2 3)
U - 1,4 D + 1. 7 L + 1. 7 W If thermal stresses due to To and Ro are present, the following combinations are also satisfied:
4)
U = (0.75) (1.4 0 + 1.7 L + 1.7 T + 1.7 R )
g g
179) 5)
11 = (n 75) (1_a n 4 1_ 7 i_ 4 1 or+1 7T 4
5 6)
U = (0.75) (1.4 D + 1.7 L + 1.7 W + 1.7 T + 1.7 R )
o g
Both cases of L having its full value or being completely absent are checked.
In addition the following combinations are considered:
7}
11 - 1_2 n 1 1,g r 7 8)
U = 1.2 D + 1.7 W i
Where soil and/or hydrostatic pressures are present, in addition to all the above combinations where they have been included in L and D, i
respectively, the requirements of ACI 349-85 are also satisfied.
Amendment I December 21, 1990
Att. 1 - 10/17 CESSAR 8Hnneuion 9
TABLE 3.8-5 (Cnnt'd)
(Sheet 6 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES 2.
Factored Load Conditions Facter Load Conditions represent Extreme Environmental, Abnormal, Abnormal / Severe Environmental and Abnormal / Extreme Environmental loads.
'a Ultimate Strength Design method is used with the following load combinations:
1)
U-D+L+T
+R
+ E' o
g 2)
U=D+L+To+R
+W g
t I
3)
U = D + L + T, + R, + 1. 5 Pa 1.25 D
+1 n (Y
+Y.+Y) + 1.25 E 4}
U D: L'T 9
1 a
a a
r a
w 4 E)
U=0+L+T
+R
+ 1.0 P + 1.0 (Yr+Y3 + Y,) + 1.0 E' a
a a
4 In factored load combinations (3), (4),- and (5), the maximum values
, including an appropriate dynamic load of P, T ' N, Y, Y, and Y" time-hi. story analysis is performei to used unless a r
a a
a factor, are justify otherwise.
Factored load combinations (2), (A), and (S) are satisfied first without the tornado missile load in (2), and without Y,
Y;, and Y in (4) and (5).
When considering these loads, howevet, local,section strength capacities may be exceeded under the r
effect of these concentrated loads, provided there will be no loss of function of any safety related system.
Both cases of L having its full value or being completely absent are checked.
Where any load reduces the effects of other loads, the corresponding coefficient for that load should be taken as 0.9 if it can be demonstrated that the load is always present or.
occurs simultaneously with other loads.
Otherwise the coefficient for the load should be taken as zero.
Where the structural effects of differential settlement, creep, or shrinkage may be significant, they should be included with the dead load, D, as applicable.
Amendment I i
. n r. n
Att. 1 - 11/17 CESSAR !anncanou D
TABLE 3.8-5 (Cont'd)
(Sheet 7 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES III. Load Combinations and Acceptance Criteria for Catecory I Steel Structures The following set of load combinations and allowable design limits is used for all Category I steel structures:
1.
Service Load Conditions Either the elastic working stress design methods or the plastic design methods of ANSI /AISC N690-1984 may be used.
If the elastic working stress design methods are used:
a.
1)
S=D+L 2)
S-0: L: E I
2 3')
S=D+L+W If thermal stresses due to To and Ro are present, the following combinations are also satisfied:
4) 1.5S=D+L+T
+R g
g E-R
-- 5 )
1. 5 S - 0 i L -T 1
1 o
o 5 6')
1.5 S - D + L + T
+R
+W g
g Both cases of L having its full value or being completely absent are checked.
b.
If plastic design methods are used:
1)
Y = 1.7 0 + 1.7 L I
2)
Y = 1.7 0 : 1.7 L : 1.7 E 2 3')
Y = 1.7 0 + 1.7 L + 1.7 W D
Amendment I
,,, n n
^ " ' ' ~ ' ' ' "
CESSAR Ealincamu I
TABLE 3.8-5 (Cont'd)
(Sheet 8 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES If thermal stresses due to To and Ro are present, the following combinations are also satisfied:
3/F)
Y = 1.3(D + L + T +R) g g
5)
Y = 1.3(D : L: E'To'R) 0 4 6')
Y = 1.3(D + L + W + T +R) g g
Both cases of L having its full value or being completely absent are checked.
2.
Factored Load Conditions The following load combinations are satisfied:
I If elastic working stress design methods are used:
I a.
1) 1.6 S = D + L + T
+R
+ E' g
g 2) 1.6 S = D + L + T
+R
+W g
g t
3) 1.6 S = D + L + T
+R
+P a
a a
Y)A 1.0 F
^ Y 4
1 n (V 1) 1.5 S* - D ' L T, 1 R a
a j
r
+ Y ) + 1.0 E' 1.0 (Y) + Y 4 5')
1.7 S* = D + L + T + Ra+P
+
r m
a a
this For the:c two combinatiord, (4) and (R), in computing the required section strength, S, the plastic section modulus of steel shapes may be used.
b.
If plastic design methods are used:
1)
Y* = D + L + T
+R
+ E' g
g 2)
Y* = D + L + T
+R
+W g
g t
3)
Y* = D + L + Ta+Ra + 1.S Pa 1.0 (Y. + Y + Y 1 + 1.25_E J) v*-nt I t T 4 p
+1 MP A
d d
a J
e Y,) + 1.0 E' 1.0 (Y) + Y
+
45)
Y* = D + L + T + R, + 1.0 P
+
r a
a i
%,a n m % n M T
^
CESSAR EEnCAnoN TABLE 3.8-5 (Cont'd)
(Sheet 9 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES i
In the above factored load combinations, thermal loads can be neglected when it can be shown that they are secondary and self-limiting in nature and where the material is ductile.
Y(for the factored load combinations) should be multiplied by 0.90 for the Internal Structures and 1.0 for other Category I structures.
In factored load combinations (3), -(4), and -(h), the minimum including an appropriate dynamic load fadtor, are! use, and Y values of P, T, T, Y, Y"d unless*,a time-history analysis is i
otherwise.
Factored load combinations
)[y performed to ju are first satisfied - without the tornado (2), (4), and Y
and Y in (4) nd '"
missile load in (2), and without Y[v,er,d, local fection strengths When considering these loads, how may be exceeded under the effect of these concentrated loads, I
provided there will be no loss of function of any t
safety-related system.
Where any load reduces the effects of other-- loads, the corresponding coefficient for that load should be taken as 0.9, if it can be demonstrated that the load is always present or i
occurs simultaneously with other loads.
Otherwise, the coefficient for that load should be taken as zero.
Where the structural effect of differential settlement may be significant it should be included with the dead load, D.
IV.
Load Combinations and Acceptance Criteria for Category I Foundations In addition to the load combinations and acceptance criteria referenced above, all Category I foundations are also checked against sliding and overturning due to earthquakes,
winds, and tornadoes and against flotation due to floods in accordance with the following:
Minimum Factors of Safety Load Combination Overturning Sliding Flotation D
">E-1.5
-1,4-D+H+W l.5 1.5 D + H + E' 1.1 1.1 D+H+W II II
~
D.
t 1,2 D + F' Amendment I December 21, 1990 1
^"-
-2"
CESSAR 88Hncy,:u TABLE 3.8-5 (Cont'd)
(Sheet 10 of 10)
LOAD COMBINATIONS FOR CATEGORY I STRUCTURES Definitions:
I D
Dead Load F'
Buoyant Forces of Design Basis Flood H
Lateral Earth Pressure E
GBE-Senmic-t-oad-(E' SSESeismicLead-)
W Wind Load W
Tornado Load t
4 4
Amend:sent I December 21, 1990
CESSAR !annc,mou m. 1 - 13f17 f
3.8.4.3 Loads and Loading combinations The Category I structures are designed to maintain their function for the following loadings:
A.
Dead Loads The dead loads include all sustained loads during and after construction.
B.
operating Loads operating loads are those loads associated with the operation of the plant.
C.
Design Basis Accident Loads The Design Basis Accident Loads are those associated with the pressure increase in the annulus due to a temperature rise as a
result of the energy release inside the containment vessel during a loss-of-coolant accident.
D.
Wind Loads The wind load is based upon ANSI A58.1-1982 (Reference 6) and ASCE Papers 3269 and 4933 (References 7 and 8) using 130 I
mph as the fastest rile of wind for a 100 year recurrence pe.riod as defined in Section 3.3.1.
The normal and tornado wind loads considered in the design of the containment shield building are nonaxisynnetric loads.
The wind loads are analyzed by approximating the wind distribution on the containment shield building as defined in ASCE Paper 4933 by a Fourier Series.
The wind distribution curves used in the design are given in Section 3.3.1.
Individual harmonics are analyzed and combined to produce the force and noment resultants for the total series.
The wind loads on the Category I structures other than the shield building are analyzed using the methods da. fined in-ASCE Paper 3269.
E.
Tornado Loadings The tornado loadings are described in Section 3.3.2.
Amendment I
-Att.
l'- 16/17.
pC"C C A D orsicN w E=*sr & M R ' CERTIFICATIO N 1
1
-l F.
Snow and Ice Loads The containment shield building and the nuclear. system annex are designed for a snow and ice load ofv5'0; pounds..per square l
Toot.
1 G.
Soil and Water Pressure The containment shield building and the nuclear system annex are designed for the earth pressure and groundwater pressure _
i defined in Section 2.4.
H.
Seismic Loads See Section 3.7,
" Seismic Design," for the seismic loadings.
f i
Loading combinations used *for the design of Category I i
structures are shown in-Table 3.8-5.
h 3.8.4.4 Desian and Analysis Procedures i
The containment shield building is designed for the static and-dynamic loads listed in Section '3.8.4.3.
The shield building is modeled as a three dimensional finite element - structure with ANSYS or another suitable computer program.-
The forces and" I
moments determined by the analysis of the applied loads.and loading combinations are used for the design of the structure in 1
accordance with ACI 349-85.
The reactor building (including the steel containment vessel, internal structure and containment shield building) is designed to prevent possible overturning, sliding and flotation.
The forces and moments acting on the building which could cause these events are determined for the different
-loads and load combinations and are then compared to the corresponding forces and moments which resist overturning, sliding ~ or flotation.
Safety factors for the possible
' events.are determined ~ for i
comparison with the allowable safety factors listed in Table 3.8-5.
The nuclear system annex is modeled ' with beam and plate finite elements.
It is analyzed for the loads and load combinations l
found in Section 3.8.4.3.
The forces and moments calculated in the analysis are used to design the walls,
slabs, beams ' and columns using. either ACI 349-85 or ANSI /AISC N690-1984 as required.
i I
l
Att. 1 - 17/17 CESSAR nmncanou 3.8.4.5 Struciural Acceptance Criteria Analysis and design of the concrete internal structures, the shield building, sub-structures, the nuclear system annex and foundations use the ultimate strength design method in accordance with ACI 349-85.
Category I structural steel analysis and design are in accordance with ANSI /AISC N690-1984.
The concrete support for the spherical containment vessel is also analyzed and designed per ACI 349-85.
(Since the support has no pressure retaining function, it is not designed in accordance with the ASME Code. )
No increase in allowable stresses for concrete or steel is permitted due to seismic or wind loadings (per the NRC Standard Review Plan, Section 3.8.4, Part II.5).
g Three orthogonal components of earthquake loads (2 horizontal and i vertical) are considered simultaneously.
g A minimum additional eccentricity of 5% of the maximum building g
5 dimension at the level under consideration is assumed to account I
for accidental torsion (per the NRC Standard Review Plan Section 3.7.2, Part II.11).
4 In addition to satisfying the load combinations for structural adequacy against the design loadings, the load combinations to ensure safety factors against overturning, sliding, and flotation are checked to ensure overall stability.
The following events are checked as a minimum:
A.
The reactor building / nuclear system annex overturning about the toe of the foundation supported on soil.
B.
The reactor building / nuclear system annex foundation sliding on soil.
C.
Floating of the entire reactor building / nuclear system annex foundation base mat.
D.
The containment vessel slipping in the lower concrete support dish.
E.
The containment vessel overturning about the edge of the lower concrete support dish.
F.
The interior structure concrete slipping inside the conrainment vessel.
Att. 2 - 1\\1 EQUIPMENT AND LIVE LOADS i
EQUIP, LOADS FLOOR AREA FLOOR VALLS LIVE LOADS l
CONTROL AREA A 55 PSF 5 PSF 200 PSF CONTROL AREA B 55 PSF 5 PSF 200 PSF CVCS AREA 75 PSF 10 PSF 300 PSF DIESEL GEN. AREA 75 PSF 10 PSF 300 PSF FUEL POOL AREA 75 PSF 10 PSF 300 PSF h REACTOR BLDG.
50 PSF 10 PSF 200 PSF MAJOR WATER STORAGE AREAS OPERATING WATER LEVEL j
IN-CONTAINMENT REFUELING VTR STORAGE TANK EL. 82'+6' REFUELING CANAL EL. 144'+ 0' SPENT FUEL POOL EL. 144'+ 0' EMERGENCY FEEDVATER TANKS EL. 100'+0' O
\\\\ \\\\\\
NOTE
//(
DIESEL GEN.
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RDL'N KEY PLAN SHOVN
/
9 APPLIES TO THE RUE PDD AREA /
. ycif+ga
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/O
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FLOOR ELEVATIONS:
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c\\
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EL. 17 0' + 0, 8
KEY PLAN FOR FLOOR & WALL LOADS
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I-WEIGHTS OF REACTOR COOLANT SYSTEM COMPONENTS COMPONENT NAME COMPONENT WEIGHT R'EACTOR VESSEL-2,050,000.LB. (TOTAL OPERAT. VEIGHT, INCL, FUEL & INTERNALS)
-{
-STEAM GENERATOR 2,600,000 LB (FLOODED) 1,700,000 : LBS. (DRY VEIGHT)
~
1 PRESSURIZER 445,000 LB. (FLOODED)
REACTOR COOLANT PUMP & MOTOR
. 279,000 LB. (DRY VEIGHT) -+ 98.4 CU. FT. VATER = 285,150 _ LB.
REACTOR COOLANT SYSTEM PIPING 280,000 LB (DRY VEIGHT)
REACTOR ~ DRAIN TANK 33,880 LB. (FULL)
SAFETY INJECTION TANK 261,000. LB (FLOODED) g
' COMPONENT NAME COMPONENT WEIGHT POLAR CRANE-630,000 LB.
FUEL BRIDGE CRANE- (NUC. ANNEX' EL.146) 35,000 LB.
CASK HANDLING CRANE - (NUC. - ANNEX -EL.146) 433,400 LB..
X+
EQUI STAGING AREA CRANE (NUC. _ ANNEX EL.146) 270,000 LB.
,ct -
P DIESEL = GENERATOR (NUC. ANNEX EL'.50).
200,000 ' LB.
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1 Notes to pressurizer supports desian basis loads:
1.
Subscripts denote: v = vertical; h = horizontal; t = torsion b = bending; k - key load The direction of unsigned loads should be chosen to give the 2.
worst loading combination.
3.
Units are Inches, Kips, and Foot-Kips.
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Flooded weight is +400 Kips.
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Att. 7 - 3\\4 Aux Building Wind Pressure (psi)
Dist Above Windward l Leeward Side l
Roof Elev Ground p(psly p(psi)_
p(psil p(psi) 91 *-9" i
0 0.311
-0.20
-0.27 101'-9" l 10 0.31!
-0.20
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-0.21
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-0.23
-0.32 131'-9" !
40 0.39 l
-0.24
-0.34 141*-9" l 50 0.40 l
-0.25
-0.35 146*-0" !
54.25 0.42
-0.26
-0.36
-0.36 151'-9" 60 0.42
-0.26
-0.37 156*-0" 64.25 0.431
-0.27
-0.38
-0.38 161'-9" !
70 1
0.44I 0.27
-0.38 l
171'-9" 80 1
0.45i
-0.28
-0.39 '
181'-9" 90 0.46 l
-0.29 '
-0.41 -
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-0.41
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Angle El 157'-0" l El 211'-0" !
El 250'-6" El 265'-O'I O
I 0.71 l 0.16;
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0.58 0.12:
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-0.19
-0.48
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j 60
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-0.84
-1.191 l
i 75 1
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-1.251 i
105
-0.43
-0.96
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120 1
-0.321
-0.65 ;
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-0.38 l
-0.40 :
-1.061 150 i
-0.38 l
-0.34i
-0.92 l 165 l
-0.36 l
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Angle i
El 157*-0" i El 211'-0" 1 El 250'-6" (
El 265'-0" O
I 2.521 0.511
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2.07 0.38!
-2.52 l 30 1
0.51
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-2.26
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75 1
-3.15
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-3.63 90 i
-3.09
-3.54 !
-3.68 105 l
-1.51
-3.021
-3.60 120 l
-1.13
-2.031
-3.45 135 l
-1.37
-1.26 l
-3.09 150 l
-1.33
-1.08 !
-2.69 l 165 1
-1.29
-0.9 6 '
-2.39 ;
180 l
-0.88
-0.88 -
-2.34 Note:
Direction of wind is from Angle 0 to 180.
Concrete Reinforcing - Critical Areas Critical Area Section Elevation Col. Line/
Section Area Azimuth orientation 1
Shear _& Shield 1A 50 to 158 D-F @ 17 Looking North Building Wall 1B 50 to 115+6 E17 Looking East 1C 50 to 158 16-18, E-F Looking East 2
East Wall 0 Turb 2
40 to 93 B14 Looking South Building 4
3 Diesel Gen. Room Ext.
3A 40 to 93 N23 Looking East
& Int. Walls 3B 40 to 93 N25 Looking East 4
Subsphere Radial Wall 4
40 to 52 2250,R33-R65 5
Shear Wall and Slab 0 SA 40 to 158 K12-K13 Looking North Emerg. FDW Pump-Room 5B 40 to 130+6 K11 LookJ.ng North and CCW Pump Room SC 40 to 158 K10-K13 Looking East
-6 SCV Anchorage Region 6
70~to 92 K
Looking West 7
SCV Support Pedestal 7
50 to.62 K33 & R33 Looking West 8
S/G Wing Wall 0 IRWST 8
70 to 91+9 Along L1.5 Looking to Center 9
RX Cavity Walls 9
62 to 146 Centerline Plan & Elev.
10 Spent Fuel Refueling 10 93'to 117 T17-18 Looking East Canal' Wall 11 Main Steam Valve 11 106 to 130 H23-25 Looking East-House Wall i
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ML20044C006: Difference between revisions

Latest revision as of 13:52, 19 December 2024

Contents

Text

_

1.0 INTRODUCTION

1.0 INTRODUCTION

Subject:

Dear Mr. Esfandiari:

. u.1

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ML20044C006: Difference between revisions

Latest revision as of 13:52, 19 December 2024

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_

1.0 INTRODUCTION

1.0 INTRODUCTION

Subject:

Dear Mr. Esfandiari:

. u.1

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