ML110250783

MUAP-09002-NP, Rev. 2, Summary of Seismic and Accident Load Conditions for Primary Components and Piping

ML110250783
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Site:	05200021
Issue date:	12/28/2010
From:	Mitsubishi Heavy Industries, Ltd
To:	Office of New Reactors
References
UAP-HF-10360 MUAP-09002-NP, Rev. 2
Download: ML110250783 (126)

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping Non-Proprietary Version December 2010 2010 Mitsubishi Heavy Industries, Ltd.

All Rights Reserved

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

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Abstract i Table of Contents iii List of Tables v to vi List of Tables vii List of Acronym 1-1 Abstract 4-2 2nd Sentence 5-1 3rd Paragraph, 2nd Sentence 5-7 Section 5.4 6-3 Section 6.4 2nd Paragraph, 1st Sentence 6-3 Section 6.4 4th Paragraph 6-5 Section 6.5 1st to 3rd Sentences Add the sentences Change the page numbers Change the page number Change the page number Add the acronyms Add the items of analysis results Correct the word FW Correct the word Shown Change the contents as including preliminary analysis results Add the sentence of description for analysis model Add the descriptions for analysis methodology Add the new paragraphs

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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2010 MITSUBISHI HEAVY INDUSTRIES, LTD.

All Rights Reserved This document has been prepared by Mitsubishi Heavy Industries, Ltd. (MHI) in connection with the U.S. Nuclear Regulatory Commissions (NRC) licensing review of MHIs US-APWR nuclear power plant design. No right to disclose, use or copy any of the information in this document, other that by the NRC and its contractors in support of the licensing review of the US-APWR, is authorized without the express written permission of MHI.

This document contains technology information and intellectual property relating to the US-APWR and it is delivered to the NRC on the express condition that it not be disclosed, copied or reproduced in whole or in part, or used for the benefit of anyone other than MHI without the express written permission of MHI, except as set forth in the previous paragraph.

This document is protected by the laws of Japan, U.S. copyright law, international teaties and conventions, and the applicable laws of any country where it is being used.

Mitsubishi Heavy Industries, Ltd.

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Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

Abstract The purpose of this technical report is to present the seismic and accident loads for the US-APWR primary components and piping and the analysis methods used to produce them.

This report describes the development of the models, the modeling method and assumptions, the analysis approach, and the following analysis results:

Forces and moments of seismic loads and accident load for primary components and piping Loading Combination for components and piping design on service level C and D

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

i Table of Contents List of Tables iii List of Figures iv List of Acronyms vii

1.0 INTRODUCTION

1-1 2.0 ANALYSIS PROCESS 2-1 2.1 Seismic Analysis 2-1 2.2 Accident Analysis 2-1 3.0 SEISMIC DESIGN 3-1 4.0 DESIGN CONDITION OF ACCIDENT ANALYSIS 4-1 4.1 Consideration of Accident Design Conditions 4-1 4.2 Method of Accident Analysis 4-3 4.2.1 Blowdown Analysis 4-3 4.2.1.1 LOCA Blowdown 4-3 4.2.1.2 Main Steam Line Break Blowdown 4-4 4.2.2 Asymmetric Compartment Pressurization Analysis 4-6 4.2.2.1 Break Conditions 4-6 4.2.2.2 Nodalization Schemes 4-6 4.2.2.3 Calculated Pressure Response 4-6

5.0 DESCRIPTION

OF RCL ANALYSIS 5-1 5.1 Model Development of RCL 5-1 5.2 Seismic Analysis of RCL 5-2 5.3 Accident Analysis of RCL 5-5 5.3.1 Forcing Function of Dynamic Analysis 5-5 5.3.2 RCL Dynamic Analysis 5-6 5.4 Nozzle Loads from Piping Reaction Force 5-7 6.0 ANALYSIS OF REACTOR EQUIPMENT 6-1 6.1 Introduction 6-1 6.2 Seismic Loads 6-1 6.3 Hydraulic Loads in LOCA Events 6-2 6.4 Reactor Internals Dynamic Response Model 6-3 6.5 Response Analysis 6-5

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

ii Table of Contents (Contd) 7.0 ANALYSIS OF PRESSURIZER 7-1 7.1 Seismic Analysis 7-1 7.2 Accident Analysis 7-1 8.0

SUMMARY

OF PRIMARY COMPONENTS LOADS 8-1 8.1 RCL Member Forces and Support Loads 8-1 8.2 Loads Related to RV and Reactor Equipment Dynamic Response 8-1 8.3 Component Loads of PZR 8-3 9.0 COMPUTER PROGRAM 9-1

10.0 CONCLUSION

10-1

11.0 REFERENCES

11-1

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

iii List of Tables Table 5-1 Support Groups and Response Spectra of RCL Seismic Analysis 5-8 Table 8-1 List of Tables and Figures of RCL Seismic and Accident Loads 8-4 Table 8-2 Seismic Loads of Inlet and Outlet Nozzle on RV, SG, RCP 8-5 Table 8-3 Seismic Reaction Forces of Support Points 8-5 Table 8-4 Seismic Member Forces of MCP 8-6 Table 8-5 Seismic Loads of Nozzles on RV, SG, MCP 8-7 Table 8-6 Accident Loads of Inlet and Outlet Nozzle on RV, SG, RCP 8-8 Table 8-7 Accident Reaction Force of Support Points 8-8 Table 8-8 Accident Member Force of MCP 8-9 Table 8-9 Accident Loads of Nozzles on RV, SG, MCP 8-10 Table 8-10 Seismic and Accident Loads on Lower Reactor Internal Assembly 8-11 Table 8-11 Seismic and Accident Loads on Upper Reactor Internal Assembly 8-12 Table 8-12 Seismic and Accident Loads on Radial Support Keys 8-13 Table 8-13 Seismic and Accident Loads between Vessel / Reactor Internals Interface Loads 8-13 Table 8-14 Seismic and Accident Loads on RV Head Nozzles and IHP Lug 8-14 Table 8-15 Seismic and Accident Loads on CRDM 8-14 Table 8-16 Seismic Loads of PZR Component 8-15 Table 8-17 Seismic Loads of PZR Nozzles 8-15 Table 8-18 Accident Loads of PZR Nozzles 8-16

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

iv List of Figures Figure 2-1 Seismic Analysis Process 2-2 Figure 2-2 Accident Analysis Process 2-3 Figure 4-1 Blowdown Analysis Model for Broken Loop (B-Loop) 4-7 Figure 4-2 Blowdown Analysis Model for Broken Loop (B-Loop) 4-8 Figure 4-3 Blowdown Analysis Model for Intact A, C & D Loops (for all cases) 4-9 Figure 4-4 Blowdown Analysis Model for Downcomer Region (all cases) 4-10 Figure 4-5 Blowdown Analysis Model for Core Region (all cases) 4-11 Figure 4-6 Pressure at Break Region 4-12 Figure 4-7 Flow Rate at Break Region 4-12 Figure 4-8 Pressure at RV Outlet Nozzle 4-13 Figure 4-9 Pressure at RV Inlet Nozzle 4-13 Figure 4-10 Differential Pressure Between Core and Downcomer 4-14 Figure 4-11 Differential Pressure Between Downcomer 0 degree and 180 degree 4-14 Figure 4-12 Pressure at Break Region 4-15 Figure 4-13 Flow Rate at Break Region 4-15 Figure 4-14 Pressure at RV Outlet Nozzle 4-16 Figure 4-15 Pressure at RV Inlet Nozzle 4-16 Figure 4-16 Differential Pressure Between Core and Downcomer 4-17 Figure 4-17 Differential Pressure Between Downcomer 0 degree and 180 degree 4-17 Figure 4-18 Blowdown Analysis Model for Main Steam Line Break (A-Line) 4-18 Figure 4-19 Pressure at break point (A-line) 4-19 Figure 4-20 Fluid Density at break point (A-line) 4-19 Figure 4-21 Flow Rate at break point (A-line) 4-20 Figure 4-22 Nodalization Scheme for SG Compartment Analysis (1/5) 4-21 Figure 4-23 Nodalization Scheme for SG Compartment Analysis (2/5) 4-22 Figure 4-24 Nodalization Scheme for SG Compartment Analysis (3/5) 4-23 Figure 4-25 Nodalization Scheme for SG Compartment Analysis (4/5) 4-24 Figure 4-26 Nodalization Scheme for SG Compartment Analysis (5/5) 4-25 Figure 4-27 Nodalization Diagram for SG Compartment Analysis (1/2) 4-26 Figure 4-28 Nodalization Diagram for SG Compartment Analysis (2/2) 4-27 Figure 4-29 Pressure Transient at The Node V70 4-28 Figure 4-30 Pressure Transient at The Node V68 4-28

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

v List of Figures (Contd)

Figure 5-1 US-APWR RCL 5-9 Figure 5-2 Stick Mass Model for RV with Internals 5-10 Figure 5-3 Stick Mass Model for SG with Internals 5-11 Figure 5-4 Stick Mass Model for RCP with Internals 5-12 Figure 5-5 RCL Piping Model 5-13 Figure 5-6 RV Support and FE Structural Model 5-14 Figure 5-7 Spring Model of SG Upper Shell Support 5-15 Figure 5-8 Spring Model of SG Intermediate Shell Support 5-16 Figure 5-9 Stick Mass Spring Model for RCL 5-17 Figure 5-10 RCL Model for Accident Analysis 5-18 Figure 5-11 SG Lower Support and RCP Tie Rod 5-19 Figure 5-12 ISRS for RCL 5-20 Figure 6-1 Horizontal Hydraulic Loads on RV during LOCA 6-6 Figure 6-2 Vertical Hydraulic Loads on RV during LOCA 6-6 Figure 6-3 RV and Internals Dynamic Analysis Model 6-7 Figure 6-4 Model Details Inside The Core Barrel and Lower Plenum 6-8 Figure 6-5 Horizontal Displacement of RV during LOCA 6-9 Figure 6-6 Vertical Displacement of RV during LOCA 6-9 Figure 7-1 Stick Mass Model for PZR 7-2 Figure 7-2 Spring Model of PZR Upper Supports 7-3 Figure 7-3 ISRS for PZR 7-4 Figure 8-1 RV Nozzle Load 8-17 Figure 8-2 SG Nozzle Load 8-17 Figure 8-3 RCP Nozzle Load 8-18 Figure 8-4 Reaction Force of SG Support Points 8-19 Figure 8-5 Reaction Force of RCP Lower Support Points 8-20 Figure 8-6 MCP Member Forces 8-21 Figure 8-7 RV Nozzles 8-22 Figure 8-8 SG Nozzles 8-23 Figure 8-9 RCP Nozzle 8-24 Figure 8-10 MCP Nozzles 8-25 Figure 8-11 RV and Reactor Internals 8-26 Figure 8-12 Locations of Stress Evaluation on Lower Reactor Internal Assembly 8-27 Figure 8-13 Locations of Stress Evaluation on Upper Reactor Internal Assembly 8-28 Figure 8-14 CRDM Assembly 8-29 Figure 8-15 Axial Force vs Elevation on CRDM (SSE+LOCA) 8-30

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

vi List of Figures (Contd)

Figure 8-16 Shear Force vs Elevation on CRDM (SSE+LOCA) 8-30 Figure 8-17 Bending Moment vs Elevation on CRDM (SSE+LOCA) 8-31 Figure 8-18 Horizontal UCP Acceleration during SSE 8-32 Figure 8-19 Horizontal LCSP Acceleration during SSE 8-32 Figure 8-20 Horizontal UCP Acceleration during LOCA 8-33 Figure 8-21 Horizontal LCSP Acceleration during LOCA 8-33 Figure 8-22 Reaction Force of PZR Lower Elements 8-34 Figure 8-23 PZR Nozzles 8-35

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

vii List of Acronyms The following list defines the acronyms used in this document.

ASME American Society of Mechanical Engineers CB Core Barrel CIS Containment Internal Structure CRDM Control Rod Drive Mechanism CSDRS Certified Seismic Design Response Spectra CVCS Chemical and Volume Control System DCD Design Control Document DOFs Degrees of Freedom DVI Direct Vessel Injection FA Fuel Assembly FEs Finite Elements FW Feedwater GT Guide Tube ICIS In-core Instrumentation System IHP Integrated Head Package ISRS In-Structure Response Spectra ISM Independent Support Motion LBB Leak-before-break LCSP Lower Core Support Plate LOCA Loss of Coolant Accident MCP Main Coolant Piping MS Main Steam NR Neutron Reflector NRC U.S. Nuclear Regulatory Commission PCCV Prestressed Concrete Containment Vessel PZR Pressurizer R/B Reactor Building RCL Reactor Coolant Loop RCP Reactor Coolant Pump RCS Reactor Coolant System RG Regulatory Guide RHR Residual Heat Removal RV Reactor Vessel SG Steam Generator SI Safety Injection

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

viii List of Acronyms (Contd)

The following list defines the acronyms used in this document.

SRP Standard Review Plan SRSS Square Root of the Sum of the Squares SSE Safe Shutdown Earthquake SSI Soil-structure Interaction UCP Upper Core Plate UCS Upper Core Support USC Upper Support Columns

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

1-1

1.0 INTRODUCTION

This technical report describes the methods and results of the analysis methods used to determine the seismic and accident loads for the US-APWR primary system components.

These loads will be used as design inputs for stress analysis as described in Subsection 3.9.3 of the US-APWR design control document (DCD).

Section 2.0 describes the steps in the analysis process used to develop the seismic and accident loads for the primary system components and piping.

Section 3.0 describes the seismic design of US-APWR standard plant which is described on the technical report MUAP-10001, Rev.1, Seismic Design Bases of the US-APWR Standard Plant (Reference 1).

Section 4.0 defines the accident design conditions including postulated accidents associated with level C and D service conditions and analyses of associated loads including initial boundary conditions, modeling, and results. The analysis of asymmetric compartment pressurization loads is included.

Section 5.0 presents analysis of the reactor coolant loop (RCL) subject to the seismic and accident loads developed in the preceding sections of the report. The RCL model and methods of dynamic analysis are described.

Sections 6.0 and 7.0 describe the analysis to define the dynamic responses of the reactor equipment and pressurizer (PZR), respectively, to the seismic and postulated accident loads.

Section 8.0 provides a summary of the seismic and accident loads for the component nozzles, supports points, and members of the main coolant system. It provides the loads at the locations to be included in the design stress analyses.

Section 9.0 describes the computer codes used in the analyses described in the report.

Finally, Section 10.0 presents the overall conclusions of the report.

Seismic and accident loads are applied the stress analysis and evaluation in accordance with American Society of Mechanical Engineers (ASME) Code,Section III (Reference 2) which requires stress evaluation on Level C and D Service conditions. Summary of stress analysis and related information of primary components and piping will be provided technical reports in a separate.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

2-1 2.0 ANALYSIS PROCESS 2.1 Seismic Analysis The seismic analysis process depicted in Figure 2-1 shows the steps used to generate the safe shutdown earthquake (SSE) loads in the primary components, piping, and the support structure.

2.2 Accident Analysis The accident analysis process depicted in Figure 2-2 shows the steps used to generate the RCL pipe rupture forces in the primary components, loop piping, and the support structures.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

2-2 Figure 2-1 Seismic Analysis Process

Input

Analysis

Output Design Seismic Loads for Class 1 Components and piping CSDRS (Certified Seismic Design Response Spectra)

Seismic Analysis RCL-R/B-CIS-PCCV Coupled Model (ANSYS)

RV In-Structure Response Spectra Seismic Analysis PZR Stick Model (ANSYS)

Acceleration, Time History Seismic Analysis Reactor Equipment Model (ANSYS)

RCP MCP SG PZR RV RI CRDM FA Seismic Analysis RCL Stick Model (ANSYS)

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

2-3 Figure 2-2 Accident Analysis Process

• 1 : Reactor Coolant Loop Force from RCL is provided for RV LOCA Analysis

• 2 : RV Displacement is provided for RCL Dynamic Analysis Design Accident Loads for Class 1 Components and piping

Input

Analysis

Output Break Condition Blowdown Analysis (MULTIFLEX Code)

• 1 Asymmetric Compartment Pressurization Analysis (GOTHIC Code)

RV & Internals Hydraulic Loads RCL Thrust Loads Jet Impingement Loads Asymmetric Cavity Pressure Loads Reactor Equipment System Dynamic Analysis (ANSYS)

RCL Dynamic Analysis (ANSYS)

• 2 RV RCP MCP SG PZR Design Load (Calculating)

RV RI CRDM FA

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

3-1 3.0 SEISMIC DESIGN This section provides the seismic design for the US-APWR primary components and piping system.

The seismic design of US-APWR standard plant structure is based on the coupled system model of the RCL and the building structures, including the reactor building (R/B), the prestressed concrete containment vessel (PCCV), and containment internal structures (CIS),

considering the effects of soil-structure interaction (SSI), which is described on the technical report MUAP-10001, Rev.1, Seismic Design Bases of the US-APWR Standard Plant (Reference 1).

On the other hand, the seismic design of the primary components and piping system is based on the independent support motion (ISM) method with the decoupled RCL analysis model, as presented in section 5.2 on this report.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

4-1 4.0 DESIGN CONDITION OF ACCIDENT ANALYSIS 4.1 Consideration of Accident Design Conditions DCD 3.9.1 identifies the following events as Level C and D service conditions.

Level C

• Small Loss of Coolant Accident (LOCA)

• Small Main Steam (MS) Line Break

• Complete Loss of Flow

• Small Feedwater (FW) Line Break

• Steam Generator (SG) Tube Rupture Level D

• Large LOCA

• Large Main Steam Line Break

• Large Feedwater Line Break

• Reactor Coolant Pump (RCP) Locked Rotor

• Control Rod Ejection US-APWR has applied the leak-before-break (LBB) methodology. As the result it was confirmed that the main coolant piping (MCP), accumulator lines and surge line satisfied the LBB criteria. (Reference 3,4,5) The postulated pipe break events that are evaluated for the reactor coolant system (RCS) loop components are as follows.

• Hot Leg Branch Line break at the 10 inch Schedule 160 Residual Heat Removal (RHR)/

Safety Injection (SI) line nozzle

• Cold Leg Branch line break at the 8 inch Schedule 160 RHR return line nozzle

• Feedwater Line break at the SG FW nozzle

• Main Steam Line break outside PCCV Following portions must be protected against mechanical loads of the LOCA and secondary side pipe rupture (MS line break and FW line break).

a. LOCA (a) Intact loop including branch piping and support components.

(b) Intact leg main coolant pipe of affected loop, SG support components and RCP support components in order to maintain the flow path.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

4-2 (c) Intact leg safety injection line connected to intact leg of affected loop in order to maintain safety injection.

(d) MS line and FW line in order to prevent simultaneous rupture of secondary side.

b. All of primary side must be protected against secondary side pipe rupture.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

4-3 4.2 Method of Accident Analysis 4.2.1 Blowdown Analysis The blowdown analysis was performed to provide the hydraulic transient input for each postulated pipe break event except FW line break accident. The thrust force due to pipe break at FW nozzle was needed in the dynamic analysis of RCS loop described in the section 5.3.1, but it was calculated according to the simplified method given in Appendix B of ANSI/ANS-58.2-1988 (Reference 6).

4.2.1.1 LOCA Blowdown (1) Introduction As shown in Figure 2-2, the blowdown analysis of a postulated pipe break accident provides the hydraulic transient input for primary coolant system stress evaluation. The blowdown analysis was performed using the MULTIFLEX hydraulic depressurization analysis code that is described in detail in Section 9.0, Computer Programs. The specification of the procedures, methodology and result for this blowdown analysis are described below.

(2) Condition (a) Analysis Cases The sizes and locations of the break for each of the analysis cases are:

Cold Leg RHR Return Line 8 inches Nozzle Break, and Hot Leg RHR and SIS Line 10 inches Nozzle Break which are consistent with the description in subsection 4.1.

(b) Initial Condition The initial condition for blowdown analysis was at 102% power in accordance with NUREG-0800 Standard Review Plan (SRP) 3.6.2. (Reference 7)

(c) Analysis Model Figure 4-1 shows the broken loop nodal diagram for the cold leg RHR Return line 8 inches nozzle break. Figure 4-2 shows the broken loop nodal diagram for the hot leg RHR and SIS line 10 inches nozzle break. Figure 4-3 shows the intact loop nodal diagram. Figure 4-4 shows the downcomer nodal diagram. Figure 4-5 shows the nodal diagram for the inside of the core barrel (CB).

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

Mitsubishi Heavy Industries, LTD.

4-4 (3) Result The blowdown hydraulic analysis for each case was performed for 500 milliseconds. Figures 4-6 through 4-17 show the results of the blowdown analysis for each case. These figures show the pressure and mass flow rate at the break region, the pressure at reactor vessel (RV) outlet and inlet nozzles, the differential pressure between the core and the downcomer regions, and the differential pressure between the downcomer regions at 0o and 180o.

4.2.1.2 Main Steam Line Break Blowdown (1) Introduction The blowdown analysis of the postulated Main Steam (MS) Line break accident also provides the hydraulic transients which are used as the forcing function for the RCL dynamic analysis.

The blowdown analysis was performed using the M-RELAP5 code that is described in section 9.0 Computer Programs. The specification of the procedures, methodology and results for this blowdown analysis are described below:

(2) Condition (a) Analysis Cases The size and location of the break for the analysis case is:

Main steam line break outside Containment Vessel (PCCV) which is consistent with the description in subsection 4.1.

(b) Initial Condition The initial condition for blowdown analysis was at 102% power in accordance with NUREG-0800 SRP 3.6.2. (Reference 7)

(c) Analysis Model There are 4 MS lines (A, B, C and D lines) in the US-APWR SG secondary side. However, the layout of A and C lines, and that of B and D lines are symmetric each other. The length of A line is different from that of B line. Therefore, it is necessary to consider the only A and B lines to evaluate the hydraulic transients for all of the MS lines in this analysis.

Figure 4-18 shows the typical MS line nodal diagram between the SG outlet nozzle and the break point for this analysis. In this model, the MS line was represented by pipe components with single junctions. Each pipe component was subdivided into the small control volumes so that the depressurization wave propagation due to the piping break can be simulated accurately. The SG and the break region were modeled as the boundary conditions, which were connected to the each side of the MS line. The loss coefficient at elbows and the wall friction loss was not considered in order to evaluate the break flow conservatively.

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4-5 (d) Result The blowdown analysis was performed for 400 milliseconds. Figures 4-19 through 4-21 show the typical results of the blowdown analysis including the pressure, fluid density, and mass flow rate at the break point of the MS line.

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4-6 4.2.2 Asymmetric Compartment Pressurization Analysis The asymmetric pressurization due to a postulated piping break was performed to provide the load conditions for components. The asymmetric pressurization analysis was performed by using the GOTHIC code. An overview of the GOTHIC code is described in section 9.0.

4.2.2.1 Break Conditions The analysis cases break sizes and locations, consistent with the description in subsection 4.1, are : RHR pump inlet line 10 inches break RHR pump outlet line 8 inches break Feedwater line 16 inches break Because the MS nozzle is located at open space without surrounding walls, the asymmetric pressurization due to MS nozzle break is not generated. Therefore MS nozzle break was excluded from the piping postulated break.

4.2.2.2 Nodalization Schemes A separate GOTHIC evaluation model was prepared for the SG compartment. In this model, the compartment was divided into nodes, with paths defined to model the transfer of mass and energy between nodes during the analyzed transient. The compartment nodalization scheme was selected so that nodal boundaries are basically at the location of flow obstructions or geometry changes within the compartment. These discontinuities create pressure differentials across nodal boundaries.

Annular configurations were nodalized circumferentially when asymmetric pressure distribution was presumed.

The nodalization scheme for US-APWR the SG compartment pressure analysis is shown in Figure 4-22 to Figure 4-26. The SG compartment was azimuthally divided into 4 sectors around the SG and the RCP, and vertically divided into 8 sectors for the SG region, and 6 sectors for the RCP region. The region close to FW line nozzle was divided in smaller nodes.

The vertical nodal boundaries are basically at the location of flow obstructions (gratings) or geometry changes. The GOTHIC nodalization for the SG compartment analysis is shown in Figure 4-27 and Figure 4-28. A total of 97 nodes, including the containment atmosphere and other compartment, are used for the SG compartment analyses.

The flow area of each vent path was conservatively estimated considering the flow obstruction by main components including margin. The friction length was conservatively estimated considering the length of the estimated flow line plus margin. The loss coefficient was conservatively estimated considering the effect of obstruction, contraction and expansion plus margin.

4.2.2.3 Calculated Pressure Response The calculated pressure transients at typical nodes in the case of FW line nozzle break are shown in Figure 4-29 and Figure 4-30. The selected nodes are close to the break point and connecting with the SG. The node including the break point is V71. The load to the SG due to the asymmetric pressurization was calculated with the pressure transient data and area of all nodes contacting the SG.

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4-7 Figure 4-1 Blowdown Analysis Model for Broken Loop (B-Loop):

(Cold Leg RHR Return Line 8 inches Nozzle Break )

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4-8 Figure 4-2 Blowdown Analysis Model for Broken Loop (B-Loop):

(Hot Leg RHR and SIS Line 10 inches Nozzle Break)

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4-9 Figure 4-3 Blowdown Analysis Model for Intact A, C & D Loops (for all cases)

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4-10 Figure 4-4 Blowdown Analysis Model for Downcomer Region (all cases)

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4-11 Figure 4-5 Blowdown Analysis Model for Core Region (all cases)

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4-12 Figure 4-6 Pressure at Break Region (Cold Leg RHR Return Line 8 inches Nozzle Break)

Figure 4-7 Flow Rate at Break Region (Cold Leg RHR Return Line 8 inches Nozzle Break)

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4-13 Figure 4-8 Pressure at RV Outlet Nozzle (Cold Leg RHR Return Line 8 inches Nozzle Break)

Figure 4-9 Pressure at RV Inlet Nozzle (Cold Leg RHR Return Line 8 inches Nozzle Break)

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4-14 Figure 4-10 Differential Pressure Between Core and Downcomer (Cold Leg RHR Return Line 8 inches Nozzle Break)

Figure 4-11 Differential Pressure Between Downcomer 0 degree and 180 degree (Cold Leg RHR Return Line 8 inches Nozzle Break)

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4-15 Figure 4-12 Pressure at Break Region (Hot Leg RHR and SIS Line 10 inches Nozzle Break)

Figure 4-13 Flow Rate at Break Region (Hot Leg RHR and SIS Line 10 inches Nozzle Break)

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4-16 Figure 4-14 Pressure at RV Outlet Nozzle (Hot Leg RHR and SIS Line 10 inches Nozzle Break)

Figure 4-15 Pressure at RV Inlet Nozzle (Hot Leg RHR and SIS Line 10 inches Nozzle Break)

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4-17 Figure 4-16 Differential Pressure Between Core and Downcomer (Hot Leg RHR and SIS Line 10 inches Nozzle Break)

Figure 4-17 Differential Pressure Between Downcomer 0 degree and 180 degree (Hot Leg RHR and SIS Line 10 inches Nozzle Break)

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4-18 Figure 4-18 Blowdown Analysis Model for Main Steam Line Break (A-Line)

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4-19 Figure 4-19 Pressure at break point (A-line)

Figure 4-20 Fluid Density at break point (A-line)

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4-20 Figure 4-21 Flow Rate at break point (A-line)

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4-21 Figure 4-22 Nodalization Scheme for SG Compartment Analysis (1/5)

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4-22 Figure 4-23 Nodalization Scheme for SG Compartment Analysis (2/5)

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4-23 Figure 4-24 Nodalization Scheme for SG Compartment Analysis (3/5)

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4-24 Figure 4-25 Nodalization Scheme for SG Compartment Analysis (4/5)

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4-25 Figure 4-26 Nodalization Scheme for SG Compartment Analysis (5/5)

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4-26 Figure 4-27 Nodalization Diagram for SG Compartment Analysis (1/2)

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4-27 Figure 4-28 Nodalization Diagram for SG Compartment Analysis (2/2)

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4-28 Figure 4-29 Pressure Transient at The Node V70 (FeedWater line 16 inches break)

Figure 4-30 Pressure Transient at The Node V68 (FeedWater line 16 inches break)

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5-1

5.0 DESCRIPTION

OF RCL ANALYSIS 5.1 Model Development of RCL Figure 5-1 shows a schematic of the US-APWR RCL. The US-APWR is a four loop plant with four safety trains. Each RCL connecting to the RV includes the SG, RCP, and the loop piping.

The loop piping consists of hot leg, crossover leg the and cold leg piping in which the coolant flows from the RV to SG, from the SG to RCP, and from RCP back to RV, respectively.

The RCL analysis model consists of the RV, SG, RCP, MCP, and component supports, as applicable, for each loop. The RCL piping and support system was modeled with three-dimensional finite elements (FEs) representing the components, pipes, and supports as beam elements, masses, and springs with imposed boundary conditions.

The RCL of the US-APWR has four loops, which are modeled as combination of RV, SG, RCP and MCP. These combined system models include both the translational and rotational stiffness, mass characteristics of RCL piping and components, and the stiffness of supports.

The analytical models of the individual components of RCL are shown in Figures 5-2, 5-3, 5-4 and 5-5.

The RV support system consists of eight steel support pads which are integrated with the inlet and outlet nozzle forgings. The support pads are placed on brackets, which are supported by an embedded steel structure on the primary shield wall. The supports allow radial thermal growth of RCS and RV. Figure 5-6 shows RV support configuration and the FE model of the support ring utilized to obtain the support stiffness coefficients.

The SG support system consists of an upper shell support structure, an intermediate shell support structure, and a lower support structure. The upper and intermediate shell supports are lateral restraints (snubbers) attached to structural steel brackets, while the lower support structure is constructed entirely of structural steel and provides both vertical and lateral support. Four pinned-end columns support the vertical loads of the SG. Each RCP support system consists of a lateral support structure, and three pinned-end structural columns. Both support structures were designed considering thermal expansion of connected piping.

Figures 5-7 and 5-8 show the FE models utilized to develop the support stiffnesses of the upper and intermediate shell supports of the SGs.

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5-2 5.2 Seismic Analysis of RCL (1) Analysis Method Figure 5-9 shows the analytical model of the entire RCL including the individual components of the four loops and the support representation. The fixed end nodes of RCL support springs is excited by the nodes associated with the CIS in the seismic event. For example, fixed end nodes which represent the ends of RV support springs at RV inlet and outlet nozzle are all excited by the nodal acceleration of IC03 CIS node at that elevation, the fixed end nodes of SG intermediate shell support is excited by the nodal acceleration of IC05, the fixed end nodes of SG upper shell support is excited by the another nodal acceleration of IC61, or IC62. So, the RCL is multiply supported to the CIS, and multiple support excitation response spectrum analysis technology is applied, as follows; The seismic analysis of the decoupled RCL model utilizes the independent support motion (ISM) method. The basic equations of motion for multiply supported system are solved, utilizing the response spectra for the three directional building responses applied to the each fixed end node of RCL model, separately.

The response of a multi degree-of-freedom linear system subjected to seismic excitation is represented by the following equation of motion:

}

]{

[

]

][

[

}

]{

[

}

]{

[

}

]{

[

b X

Keb K

M X

K X

C X

M 1

(5-1) where

M mass matrix (nxn),

C damping matrix (nxn)

K stiffness matrix (nxn)

x column vector of relative displacements (nx1)

x column vector of relative velocities (nx1)

x column vector of relative accelerations (nx1)

]

[Keb = boundary stiffness coupling matrix between the degree of freedom node of RCL and the corresponding fixed building structural node

}

{ b X

= column vector of accelerations for each fixed building node Equation of multiple support excitation motion (5-1) is expressed in the following form by the modal transformation.

}

]{

[

}

]{

[

}

]{

2

[

}

{

2 b

X q

q h

q j

i i

i i

(5-2) where

]

[

j i

participation factors in mode i for the dependent support points in group j

j T

j i

I Keb K

M 1

1

]

[

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5-3 The seismic response of multiply supported RCL system was evaluated, based on the ISM method, of NUREG-1061, Vol.4 (Reference 8), and SRP 3.7.2, Rev.3 (Reference 9), SRP 3.7.3, Rev.3 (Reference 10) and Regulatory Guide (RG) 1.92, Rev.2 (Reference 11) in the following way.

a. For Inertial or Dynamic Components

1) Group responses for each direction were combined by the absolute sum method.

2) Modal and directional responses were combined by the square root of the sum of the square (SRSS) method in the range of frequencies up to 50 Hz without considering closely spaced frequencies.

b. For Missing Mass Effects

1) Static analysis for the applied load that equals the missing mass multiplied by the spectrum ZPA, that is, ZPA Dj M

)

1(

, where

N i

i i

Dj

c. For the Pseudo-static Components

1) For each group, the maximum relative displacements of building nodes were evaluated for each exciting direction.

2) Forced displacement analysis of RCL, applied to the maximum displacements corresponding to the fixed end node of the RCL support, was performed.

d. For the total Response

1) Dynamic responses, missing mass effects, and the pseudo-static components were combined by the SRSS rule.

Combined response (Acceleration, Velocity, Displacement, Support Load or Element Force) for each N-S, E-W, and Vertical excitation direction, is as follows;

]

)

(

)

(

)

[(

2 max 2

max 2

max UD i

EW i

NS i

i F

F F

SQRT R

(5-3) where i ; suffix for the degree of freedom of each x, y, z, x, y, z direction iR ; combined response for i-th direction (ex. Fx, Fy, Fz, Mx, My, Mz)

NS iF

)

(

max

Maximum i-th directional response for N-S excitation EW iF

)

(

max

Maximum i-th directional response for E-W excitation UD iF

)

(

max

Maximum i-th directional response for Vertical excitation

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5-4 (2) Support Groups and Response Spectra The relation between the fixed end nodes of RCL model and groups corresponding to the fixed building nodes, where each response spectrum is defined in the ISM analysis, is presented in Table 5-1.

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5-5 5.3 Accident Analysis of RCL The postulated pipe break conditions for the RCS loop are as follows

• Hot Leg Branch Line break at the 10 inches RHR/SI line nozzle

• Cold Leg Branch Line break at the 8 inches RHR return line nozzle

• FW Line break at SG FW nozzle

• MS Line break outside PCCV The design loads were set based on the dynamic analysis of RCS loop for these pipe break conditions. The dynamic analysis method is shown below.

5.3.1 Forcing Function of Dynamic Analysis Hydraulic forcing functions considered in the postulated RCS pipe break events are as follows,

• thrust forces that include the jet force at the break point and system internal hydraulic forces

• jet impingement force from the ruptured piping

• asymmetric compartment pressure force based on the compartment pressure analysis Additionally, since the RV is oscillated by the internal hydraulic forces in case of the pipe break at the branch line nozzle of hot leg and cold leg, RV dynamic motion is also considered for these cases in RCS loop dynamic analysis.

a. Calculation of thrust force Hydraulic forces acts on each part of the RCL system by jet thrust force from the break location, or flow change in a system.

The time history hydraulic forces were calculated using the pressure transient, flow rate, and other coolant property obtained by blowdown analysis described in the section 4.2.1. the jet blowdown force is calculated by the following equation.

A G

A Pa P

A u

A Pa P

Fj

2 2

)

(

)

(

)

(

where:

j F : jet force P : fluid pressure at the break plane area Pa : ambient pressure A : break plane area u : flow rate

mass density G : mass flow rate.

u G

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5-6 The system internal hydraulic forces were calculated at various locations of the RCS loop, such as elbow, RCP, SG plenum using time history hydraulic property of P, G, and control volume surface area.

b. Calculation of jet impingement force The jet impingement force is calculated according to the equation given in Appendix D of ANSI/ANS-58.2-1988. (Reference 6) e O

T j

imp A

P C

K F

K F

where:

imp F

jet impingement force jF

e O

T A

P C

K

= shape factor T

C = steady state thrust coefficient O

P = initial pressure eA = break plane area

c. Asymmetric compartment pressure force The asymmetric compartment pressure force acting on the SG and the RCP was calculated based on the pressure time history result of the compartment pressurization analysis described in section 4.2.2. This force was applied to the broken loop of four loop RCL model.

d. RV dynamic motion The RV was oscillated by the internal hydraulic forces in the pipe break cases at the branch line nozzle of hot leg and cold leg.

Therefore RV dynamic motion was also considered for these pipe break cases. Time history displacement data at RV center was loaded in the RCS loop dynamic analysis.

5.3.2 RCL Dynamic Analysis The RCL was vibrated by the hydraulic force which acts at the piping rupture. The analysis method used to calculate the load is as follows.

The RCL structural model was created using the ANSYS code. A four loop structural model centering on RV model was developed. Considering the symmetry of the four loops, the break can be postulated in any one loop.

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5-7 RCL model consists of the following structures.

- Reactor vessel (RV)

- Steam generator (SG)

- Reactor coolant pump (RCP)

- Main coolant pipes (MCP)

- Primary component support Three-dimensional beam element and pipe element was used in the model. The support structures were modeled by spring elements. The four loop RCL model is shown in Figure 5-10. Non-linear spring element was used for SG lower lateral support considering those bumper support structures as shown in Figure 5-11.

Time history direct integration method was applied to obtain the dynamic load of the RCL in the analysis. Forcing functions described in section 5.3.1 were assumed to act on the structural model.

- Direct integration

Newmark beta method

- Break opening time

0.001 second

- Integration time interval : 0.0001 second

- Analysis time

0.5 second Rayleigh damping was applied to the model. Critical damping ratios were set 3% to the applicable RG 1.61, Rev.1 (Reference 12) 5.4 Nozzle Loads from Piping Reaction Force The RCL components/piping nozzle loads were conservatively determined as estimated loads reflecting the RCL branch piping and MS piping preliminary analysis results. These estimated loads will be confirmed, in the technical reports (Reference 13), to envelope the actual loads obtained from the analysis of RCL branch piping and MS piping.

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5-8 Table 5-1 Support Groups and Response Spectra of RCL Seismic Analysis RCL Node Number of Same Support Group GroupNo.

Location Spectrum LOOP-A LOOP-B LOOP-C LOOP-D 2203 4203 6203 8203 2206 4206 6206 8206 2209 4209 6209 8209 2212 4212 6212 8212 2303 4303 6303 8303 2306 4306 6306 8306 1

Base Floor of SG,RCP Columns ISRS of IC02OR 2309 4309 6309 8309 2601 4601 6601 8601 2602 4602 6602 8602 2603 4603 6603 8603 2

RV Support ISRS of IC03OR 2604 4604 6604 8604 2760 6760 2761 6761 6766 8766 6767 8767 4768 8768 4769 8769 3

SG,RCP Lower Support ISRS of IC04OR 2811 4811 6811 8811 2762 4762 6762 8762 4

SG Intermediate Shell Support ISRS of IC05OR *)

2763 4763 6763 8763 2764 4764 6764 8764 5

SG Upper Shell Support ISRS of ICSGOR 2765 4765 6765 8765

• note) Response spectra of group 4 are presented in Figure 5-12.

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5-9 Figure 5-1 US-APWR RCL STEAM GENERATOR REACTOR COOLANT PUMP REACTOR VESSEL CROSSOVER LEG COLD LEG HOT LEG

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5-10 Figure 5-2 Stick Mass Model for RV with Internals 705 702 706 707 708 703 704 710 709

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5-11 Figure 5-3 Stick Mass Model for SG with Internals 131 140 139 137 133 135 121 134 129 132 130 120

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5-12 Figure 5-4 Stick Mass Model for RCP with Internals 167 168 300 169 174 175 176 360 361 177 364 179 374 180 377 376 375 383 365 369 178 385 384 366

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5-13 Figure 5-5 RCL Piping Model 107 109 111 112 113 153 155 157 159 160 161 181 183 185 186 189 191 (194)

(117) 115 141 142 143 145,147,149,151,152 165,(167) 163 RCP SG Main Coolant Piping Global Layout Hot Leg Cold Leg Crossover Leg RV RCP Cold Leg (Side View) 181 183 186 191 185 (194) 189 SG RV Hot Leg 115 111 109 107 (117) 112 (Side View) 113 RCP SG Crossover Leg 141 152 155 161 165 145 142 (167) 159 157 160 147 151 (Side View) 143 163 153 149 1014 1007 1004 1012 1004 1007 RV 1013 1015 1014 1013 1005 1005 1012 1015

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5-14 Figure 5-6 RV Support and FE Structural Model Krh Krv RV (2196)

Krv Krh (4196)

(8196)

(6196)

Krh Krh Krh Krh Krh Krh Krh (6105)

Support Point (2105)

(4105)

(8105)

Krv Krv Krv Krv Krv Krv Krv Support Point Support Point Support Point RV Outlet Nozzle RV Outlet Nozzle RV Outlet Nozzle RV Outlet Nozzle Support Point RV Inlet Nozzle Support Point RV Inlet Nozzle Support Point RV Inlet Nozzle Support Point RV Inlet Nozzle

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5-15 Figure 5-7 Spring Model of SG Upper Shell Support (RV dir.)

22.5° 22.5° Kuy Kux

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5-16 Figure 5-8 Spring Model of SG Intermediate Shell Support (RV dir.)

Kmy Kmx

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5-17 Figure 5-9 Stick Mass Spring Model for RCL 705 704 703 702 706 707 708 709 710 2102 2105 2603 2604 2107 2109 3004 2111 2112 2113 2115 2117 2130 2141 2142 2143 2145 2147 2149 2151 2152 2153 2155 2157 3007 2159 2160 2161 2163 2165 2167 2168 2300 2173 2181 3014 3012 2183 2185 2186 2189 2191 2194 2196 2601 2602 2198 2129 2131 2132 2134 2121 2135 2133 2137 2139 2140 2120 2125 2201 2202 2203 2126 2204 2205 2206 2127 2207 2208 2209 2128 2210 2211 2212 2760 2761 2762 2763 2764 2765 2169 2175 2176 2360 2361 2180 2174 2170 2301 2302 2303 2171 2304 2305 2306 2172 2307 2308 2309 2811 4102 4105 4603 4604 4107 4109 5004 5005 4111 4112 4113 4115 4117 4130 4141 4142 4143 4145 4147 4149 4151 4152 4153 4155 4157 5007 4159 4160 4161 4163 4165 4167 4168 4300 4173 4181 5013 5014 4183 4185 4186 4189 4191 4194 4196 4601 4602 4198 4129 4131 4132 4134 4121 4135 4133 4137 4139 4140 4120 4125 4201 4202 4203 4126 4204 4205 4206 4127 4207 4208 4209 4128 4210 4211 4212 4762 4763 4764 4765 4169 4175 4176 4360 4361 4364 4180 4174 4170 4301 4302 4303 4171 4304 4305 4306 4172 4307 4308 4309 4811 6102 6105 6603 6604 6107 6109 7004 6111 6112 6113 6115 6117 6130 6141 6142 6143 6145 6147 6149 6151 6152 6153 6155 6157 7007 6159 6160 6161 6163 6165 6167 6168 6300 6173 6181 7013 7014 6183 6185 6186 6189 6191 6194 6196 6601 6602 6198 6129 6131 6132 6134 6121 6135 6133 6137 6139 6140 6120 6125 6201 6202 6203 6126 6204 6205 6206 6127 6207 6208 6209 6128 6210 6211 6212 6760 6761 6762 6763 6764 6765 6169 6176 6360 6364 6180 6174 6170 6301 6302 6303 6171 6304 6305 6306 6172 6307 6308 6309 6811 8102 8105 8603 8604 8107 8109 9004 8111 8112 8113 8115 8117 8130 8141 8142 8143 8145 8147 8149 8151 8152 8153 8155 8157 9007 8159 8160 8161 8163 8165 8167 8168 8300 8173 8181 9014 8183 8185 8186 8189 8191 8194 8196 8601 8602 8198 8129 8131 8132 8134 8121 8135 8133 8137 8139 8140 8120 8125 8201 8202 8203 8126 8204 8205 8206 8127 8207 8208 8209 8128 8210 8211 8212 8762 8763 8764 8765 8169 8175 8176 8360 8364 8180 8174 8170 8301 8302 8303 8171 8304 8305 8306 8172 8307 8308 8309 8811 712 711 3005 7005 9005 3013 5012 7012 9012 9013 4768 4769 6766 6767 8768 8769 8767 8766 3015 5015 7015 9015 Global Coordinate Systems Y

X Z

PN CROSS OVER LEG COLD LEG HOT LEG Reactor Vessel Reactor Coolant Pump Steam Generator LOOP-B LOOP-A LOOP-C LOOP-D 2364 2365 2385 2178 2369 2374 2179 2383 2375 2376 2377 2177 2366 2384 4177 4364 4384 4385 4178 43694374 4365 4179 4376 4377 4366 6167 6300 6175 6361 6377 6364 6766 6384 6385 6178 6369 6374 6365 6179 6383 6375 6376 6366 6177 8361 8177 8364 8179 8365 8384 8385 8178 8369 8374 8383 8375 8376 8377 8366 Mass Point Fixed edge Support Node Point 4383 4375

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5-18 Figure 5-10 RCL Model for Accident Analysis 705 704 703 702 706 707 708 709 710 2102 2105 2603 2604 2107 2109 3004 2111 2112 2113 2115 2117 2130 2141 2142 2143 2145 2147 2149 2151 2152 2153 2155 2157 3007 2159 2160 2161 2163 2165 2167 2168 2300 2173 2181 3014 3012 2183 2185 2186 2189 2191 2194 2196 2601 2602 2198 2129 2131 2132 2134 2121 2135 2133 2137 2139 2140 2120 2125 2201 2202 2203 2126 2204 2205 2206 2127 2207 2208 2209 2128 2210 2211 2212 2762 2763 2764 2765 2169 2175 2176 2360 2361 2180 2174 2170 2301 2302 2303 2171 2304 2305 2306 2172 2307 2308 2309 2811 4102 4105 4603 4604 4107 4109 5004 5005 4111 4112 4113 4115 4117 4130 4141 4142 4143 4145 4147 4149 4151 4152 4153 4155 4157 5007 4159 4160 4161 4163 4165 4167 4168 4300 4173 4181 5013 5014 4183 4185 4186 4189 4191 4194 4196 4601 4602 4198 4129 4131 4132 4134 4121 4135 4133 4137 4139 4140 4120 4125 4201 4202 4203 4126 4204 4205 4206 4127 4207 4208 4209 4128 4210 4211 4212 4762 4763 4764 4765 4169 4175 4176 4360 4361 4364 4180 4174 4170 4301 4302 4303 4171 4304 4305 4306 4172 4307 4308 4309 4811 6102 6105 6603 6604 6107 6109 7004 6111 6112 6113 6115 6117 6130 6141 6142 6143 6145 6147 6149 6151 6152 6153 6155 6157 7007 6159 6160 6161 6163 6165 6167 6168 6300 6173 6181 7013 7014 6183 6185 6186 6189 6191 6194 6196 6601 6602 6198 6129 6131 6132 6134 6121 6135 6133 6137 6139 6140 6120 6125 6201 6202 6203 6126 6204 6205 6206 6127 6207 6208 6209 6128 6210 6211 6212 6761 6762 6763 6764 6765 6169 6176 6360 6364 6180 6174 6170 6301 6302 6303 6171 6304 6305 6306 6172 6307 6308 6309 6811 8102 8105 8603 8604 8107 8109 9004 8111 8112 8113 8115 8117 8130 8141 8142 8143 8145 8147 8149 8151 8152 8153 8155 8157 9007 8159 8160 8161 8163 8165 8167 8168 8300 8173 8181 9014 8183 8185 8186 8189 8191 8194 8196 8601 8602 8198 8129 8131 8132 8134 8121 8135 8133 8137 8139 8140 8120 8125 8201 8202 8203 8126 8204 8205 8206 8127 8207 8208 8209 8128 8210 8211 8212 8762 8763 8764 8765 8169 8175 8176 8360 8364 8180 8174 8170 8301 8302 8303 8171 8304 8305 8306 8172 8307 8308 8309 8811 712 711 3005 7005 9005 3013 5012 7012 9012 9013 3015 5015 7015 9015 Global Coordinate Systems Y

X Z

PN CROSS OVER LEG COLD LEG HOT LEG Reactor Vessel Reactor Coolant Pump Steam Generator LOOP-B LOOP-A LOOP-C LOOP-D 2364 2365 2385 2178 2369 2374 2179 2383 2375 2376 2377 2177 2366 2384 4177 4364 4384 4385 4178 43694374 4365 4179 4376 4377 4366 6167 6300 6175 6361 6377 6364 6384 6385 6178 6369 6374 6365 6179 6383 6375 6376 6366 6177 8361 8177 8364 8179 8365 8384 8385 8178 8369 8374 8383 8375 8376 8377 8366 Mass Point Fixed edge Support Node Point 4383 4375 2768 2769 2767 2760 2761 4768 4769 4760 4761 4766 4767 6760 6767 6768 6769 8768 8769 8767 8766 8761 8760 BREAK POINT Support (Compression only) 2766 6766

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5-19 Figure 5-11 SG Lower Support and RCP Tie Rod Cold Leg CL.

Cross CL.

Over Leg RCP Tie Rod (170)

Kpt SG HOT LEG CL.

SG Lower Lateral Support(2)

(126)

Support(1)

SG Lower Lateral (125)

Support(3)

SG Lower Lateral (126)

Support(4)

SG Lower Lateral (127)

Support(5)

SG Lower Lateral (127)

Support(6)

SG Lower Lateral (128)

K5y K6y K6x K7x K7y K8y

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5-20 Figure 5-12 ISRS for RCL N-S Direction (3 % Damping)

E-W Direction (3 % Damping)

Vertical Direction (3 % Damping) 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 0.1 1

10 100 Frequency [Hz]

Acceleration [g]

IC05OR 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 0.1 1

10 100 Frequency [Hz]

Acceleration [g]

IC05OR 0.0 1.0 2.0 3.0 4.0 5.0 0.1 1

10 100 Frequency [Hz]

Acceleration [g]

IC05OR

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6-1 6.0 ANALYSIS OF REACTOR EQUIPMENT 6.1 Introduction The review requirements of SRP 3.9.2, Rev. 3 (Reference 14), create a need for a detailed discussion of the reactor internals, design criteria and dynamic analyses methodology for the combined the seismic and postulated pipe rupture events under the ASME Level D (faulted) service conditions. The results of the analyses are required to meet the stress limits of the ASME Code,Section III (Reference 2), Subsection NG for the core support structures, and the functional requirements of the reactor internals design specification. Meeting the requirements of the ASME Code,Section III (Reference 2) and the design specification provides assurance of the structural and functional integrity of the reactor internals under the ASME Level D service conditions, combined loads of the seismic and pipe rupture events.

The time history displacements of RV obtained by this analysis were used as the input of the RCS, and Direct Vessel Injection (DVI) piping dynamic response analysis described in Section

5. The time histories of reactor internals acceleration response were used to dynamic analysis of the fuel assembly. The maximum element forces and reaction forces on the RV, the reactor internals and the control rod drive mechanism (CRDM) were used for the stress analysis of the core support structures, RV and CRDM.

6.2 Seismic Loads The seismic loads for the reactor equipments were obtained from the time history analysis with the finite element model of the coupled RCL-Building as discussed in Section 5.2. The time history data of the earthquake response at the vessel support and operating floor level nodes were used as the inputs. The analysis was performed for three orthogonal (two horizontal and one vertical) components of earthquake ground motion.

Additional loading input to the seismic analysis were vertical pressure loadings converted to nodal point external loads, and the vertical weights of the reactor internals and interfacing components, as inputs of density on the beams with spring effects or mass nodal points.

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6-2 6.3 Hydraulic Loads in LOCA Events The Hydraulic loads on the RV and internals during the blowdown phase of the LOCA events were calculated from the dynamic pressure inside the vessel obtained by MULTIFLEX code as discussed in Section 4.2. The calculation methods of the horizontal and vertical part for hydraulic loads are discussed below. In addition, the thrust forces from the RCL analysis were combined with the hydraulic loads as the input.

(1) Horizontal Loads At the beginning of blow-down phase of the LOCA events, the region of the downcomer annulus close to the break depressurizes rapidly, but because of the finite speed of sound, the opposite side of the CB remains at a high pressure. This pressure difference in the downcomer causes asymmetric horizontal loads on the RV and the CB. The depressurization wave propagates around the downcomer annulus and goes up though the core, causing the CB differential pressure to be reduced and decreasing, the resulting hydraulic forces.

For the forces calculation, the CB was divided into [ ] force segments in the axial direction and the [ ] in the circumferential direction. The horizontal time history force on each segment was calculated by the circumferential integration of the downcomer pressure distribution around the CB obtained from the blowdown analysis. Dynamic loads on the RV were also calculated in the same manner of the CB.

The result of total horizontal force on the CB is shown in Figure 6-1 for the LOCA event of the 8 inches pipe rupture connecting to the cold leg.

(2) Vertical Loads The depressurization wave also causes the vertical differential pressure across the reactor internals and fuel assemblies.

The vertical dynamic loads on each structural element were calculated by multiplying the pressure time histories obtained from the blow-down analysis and the projection area of each element.

The results of total vertical force on the RV and its internals are shown in Figure 6-2 for the LOCA event of the 8 inches pipe rupture connecting to the cold leg.

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6-3 6.4 Reactor Internals Dynamic Response Model The dynamic analysis methodology was based on the models using the general purpose FE computer code ANSYS. The reactor equipment model was a three-dimensional non linear dynamic FE computer model representing the reactor internals and the support system and was used to determine the maximum accelerations, displacements, and member forces. The details of the seismic and LOCA dynamic analysis model are discussed below.

Both the seismic and LOCA dynamic analyses models consist of beam elements, linear and non-linear springs, gaps, hydrodynamic mass matrices, and stiffness matrices. The only shell element modeled was for the diffuser plates. The model includes representation of the RV support system, inlet and outlet piping nozzles, the CRDM, integrated head package, in-core instrumentation columns, and fuel assembly nozzles and grids.

Figures 6-3 and 6-4 show a typical model of the reactor internals used for the seismic and the LOCA dynamic analysis. The physical geometry and the material properties (density, modulus of elasticity, Poissons ratio) of the reactor internals were represented by the beam elements. The reactor internals and interfacing structures were connected or represented by the mass inertia effects, stiffness matrices, and hydro-dynamic matrices, springs, and/or impact elements including gap and damping (including coexistence of viscous and Coulomb damping).

The nodal point DOFs, and the damping coefficients of the reactor internals and surrounding structures were selected such that the most dominant frequencies were represented in the seismic - LOCA response. This forms the bases for establishing any directional decoupling and system structural partitioning in the seismic-LOCA system models.

The main structures were modeled by the beam elements. The structures using beam elements are:

Reactor Vessel (RV)

Core Barrel (CB)

Control Rod Drive Mechanism (CRDM)

Integrated Head Package (IHP)

Upper Core Support (UCS)

Upper Core Plate (UCP)

Lower Core Support Plate (LCSP)

Guide Tube (GT)

Upper Support Columns (USC)

In-core Instrumentation System (ICIS)

Neutron Reflector (NR)

Fuel Assembly (FA)

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6-4 The structural interfaces were modeled by the spring, gap, and impact elements. The structures using linear spring elements are:

RV supports in tangential direction (8 locations)

Holddown Spring in vertical direction Fuel Assembly Holddown Spring in vertical direction The structures modeled by the non-linear spring, gap, and impact elements are:

RV supports in vertical direction (8 locations)

Radial keys between the RV and CB in tangential direction (6 locations)

UCP guide pins between the UCP and CB in tangential direction (4 locations)

CB outlet nozzles between the CB and RV in radial direction (4 locations)

Between CB flange and RV flange in radial direction (16 locations)

Between UCS flange and RV flange in radial direction (16 locations)

Between FA and NR in horizontal direction Between FAs in horizontal direction FA nozzle and UCP and LCSP in vertical direction The RCS loops were not explicitly modeled in the seismic and LOCA dynamic models; however they were simulated by stiffness matrices connected to the RV nozzle center location.

The fluid-structural effects were accounted for between the RV and CB, CB and NR, UCS and RV by hydrodynamic mass matrices. The effects of friction between the CB flange and RV, the UCS flange and RV were accounted for by the friction elements. The shell elements were used for modeling the diffuser plates. The core region was divided to five regions, one center region and four outer regions. The beam elements were used for representing the UCS, UCP and LCSP vertical vibration and out of plane stiffness.

Fluid-structure interaction effects were accounted for by matrices developed for that purpose.

The Hydrodynamic masses are calculated for the following locations in the seismic analysis model.

(1) Between RV and CB in two horizontal directions (2) Between CB and NR in two horizontal directions (3) Between UCS and RV head in vertical direction Validation that the reactor internals dynamic models were representative is made by the comparison of a simulation analysis of 1/5 scale model test and the test results as discussed in Reference 15.

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6-5 6.5 Response Analysis The time history seismic accelerations developed from the RCL seismic analysis were applied to the RV supports and IHP locations. The hydraulic forces acting on the reactor internal structures during normal operation were accounted for in the seismic dynamic analysis by applying the loads as steady-state uniform loadings.

The LOCA time history forcing functions from the MULTIFLEX analysis were applied to the RV and CB in the horizontal direction and the vertical forces are applied to the RV, UCS, CB, LCSP, UCP and FAs.

The effects of flow upon both the lumped mass and flexibility properties in the LOCA dynamic system model were accounted for in the model because the MULTIFLEX results used as input to the LOCA dynamic system model included fluid-structural interaction. In the LOCA dynamic analysis, the hydrodynamic mass matrices between RV and CB were deleted because the hydrodynamics mass effect were included in the pressure force as the output of blow-down analysis code, MULTIFLEX.

(1) Calculation Method The time history analysis with the direct time integration method was applied to both SSE and the LOCA response analysis. The 4% of critical damping ratio was used with the Raleigh damping method. The dead weights of structures and vertical interface loads such as the hold down spring force were also taken into the response analyses.

(2) Response Combination The outputs of the SSE and the LOCA response analysis are time-history accelerations, displacements (absolute and relative), and member forces (forces and moments). The maximum member forces were input into reactor internals component static FE models and the maximum stress intensities were compared to the ASME Code,Section III. (Reference 2)

The interface loads and displacements results were compared to the allowable interface load and displacement limits specified in Table 3.9-2 of the DCD Revision 1 Subsection 3.9.2.

The LOCA dynamic system analyses results confirm that the adequacy of the structural design of the reactor internals can withstand the dynamic loadings of the most severe LOCA in combination with the SSE.

(3) Results of Response Analysis The horizontal and vertical displacements at the elevation of inlet nozzle center are shown in Figure 6-5 and Figure 6-6 as examples of the RV response analysis result.

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6-6 Figure 6-1 Horizontal Hydraulic Loads on RV during LOCA Figure 6-2 Vertical Hydraulic Loads on RV during LOCA

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6-7 Figure 6-3 RV and Internals Dynamic Analysis Model CRDM IHP Neutron Reflector Core Barrel Reactor Vessel

[SM]

Reactor Vessel Support RV FA FA FA Impact Spring Linear Spring Hydrodynamic Mass Coupling

[SM]

Stiffness Matrix Beam Element Reactor Vessel RV Core Barrel CB Neutron Reflector NR Fuel Assembly FA

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6-8 Figure 6-4 Model Details Inside The Core Barrel and Lower Plenum

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6-9 Figure 6-5 Horizontal Displacement of RV during LOCA Figure 6-6 Vertical Displacement of RV during LOCA

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7-1 7.0 ANALYSIS OF PRESSURIZER The PZR stick mass spring model consists of lumped masses, pipe elements, beam elements and spring elements, as shown in Figure 7-1. The structural model in the figure is a three dimensional two-stick model. The numbers in the figure shows the node numbers. One stick represents the stiffness and mass properties of the PZR pressure boundary shell, and the second stick at the bottom represents the skirt. Both sticks were tied in the double line between node 3 and node 113, which represents the rigid connection. The total number of DOFs is adequate to represent the dynamic behavior of the PZR component. The model includes spring elements at the interface point of the upper lateral support at node 11, which represents the bumper stiffness and the local shell flexibility, as shown in Figure 7-2.

7.1 Seismic Analysis This section summarizes the seismic analysis of the PZR component to obtain the SSE earthquake loadings for the element stress evaluation. Seismic loadings of the PZR component were evaluated by the response spectrum analysis method, using the PZR structural model shown in Figure 7-1. In the response spectrum analysis, three orthogonal earthquakes components(X, Y, Z directions) were applied with SRSS combination.

The in-structure response spectra (ISRS) at IC15OR and ICP2OR referred to the CIS model of Figure 3-1, was developed with a 3 % damping for the analysis, as shown in Figure 7-3.

These 7.2 Accident Analysis This section summarizes the accident analysis of the PZR component. Pipe rupture at RCL accident condition are postulated in the RCL branch piping, MS piping or FW piping considering LBB methodology as stated in Section 4.1. Significant Accident load of the PZR is generated by surge line vibration caused by MCP dynamic response in the LOCA and other reaction force such as the PZR spray line is negligible small. Consequently component loads other than surge line nozzle do not occur in accident condition.

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7-2 Figure 7-1 Stick Mass Model for PZR

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7-3 Figure 7-2 Spring Model of PZR Upper Supports Kn Ke PN Global Cordinate Systems X

Y Z

Global Coordinate System

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7-4 Figure 7-3 ISRS for PZR N-S Direction (3 % Damping)

E-W Direction (3 % Damping)

Vertical Direction (3 % Damping) 0.0 4.0 8.0 12.0 16.0 20.0 0.1 1

10 100 Frequency [Hz]

Acceleration [g]

IC15OR ICP2OR Envelope 0.0 4.0 8.0 12.0 16.0 20.0 0.1 1

10 100 Frequency [Hz]

Acceleration [g]

IC15OR ICP2OR Envelope 0.0 0.5 1.0 1.5 2.0 2.5 3.0 0.1 1

10 100 Frequency [Hz]

Acceleration [g]

IC15OR

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8-1 8.0

SUMMARY

OF PRIMARY COMPONETNS LOADS 8.1 RCL Member Forces and Support Loads (1) Seismic Loads The component nozzle load, the reaction force of support point, and the member forces of the MCP at SSE design condition are presented in the tables and figures of Table 8-1.

(2) Accident Loads The force tables for the accident loads are presented in a same manner in Table 8-1. The accident design loads were determined conservatively based on the result of RCS loop dynamic analysis described in the section 5.3. The accident design loads are presented by the Table 8-6 through Table 8-9.

The load in each table is presented as the maximum umbrella design load for the following break conditions.

Hot Leg Branch Line break at the 10 inches RHR/SI line nozzle Cold Leg Branch Line break at the 8 inches RHR return line nozzle FW Line break at SG FW nozzle MS Line break outside PCCV 8.2 Loads Related to RV and Reactor Equipment Dynamic Response The general assembly of the RV and reactor internals is shown in Figure 8-11. The SRSS of the SSE and the LOCA loads obtained from the dynamic analysis of the RV are summarized as follows.

(1) Member Forces of Core Support Structures The member forces on core support structures are summarized in from Table 8-10 to Table 8-12. The dynamic output parameters are used into the detailed structures component static FE model and the maximum stress intensities are calculated. The results are compared to the ASME Code,Section III Level D service limits. Locations that are likely to be sources of high stresses during a Level D event are structures that have to transmit high loads and have limiting minimum thicknesses. Critical core support structure locations identified in Figures 8-12 and 8-13 include:

Core Barrel Flange Discontinuity Upper/Lower Core Barrel Discontinuity Lower Core Barrel/Lower Core Support Plate Discontinuity Radial Support Key

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8-2 Lower Core Support Plate Upper Core Support UCS Flange/Skirt Discontinuity Top Slotted Column Extension Top Slotted Column Top Slotted Column Fastener Upper Core Support Extension Upper Core Support Column Upper Core Support Column Fastener.

(2) The RV and reactor internals Interface Loads The interface loads between the RV and reactor internals are summarized in Table 8-13.

(3) Member Forces on the RV Head The Member forces on the RV Head for the CRDM and instrumentation systems are summarized in Table 8-14.

(4) Member Forces on CRDM The CRDM assembly is shown in Figure 8-14. The member force on the CRDM are summarized in Table 8-15. The distributions of the axial force, the shear force and the bending moment on the CRDM are shown in Figure 8-15 through Figure 8-17.

(5) Fuel Assembly Loads The accelerations of the upper core plate and the lower core support plate during SSE are shown in Figure 8-18 and Figure 8-19 as examples of the RV response analysis result.

The horizontal accelerations of the upper core plate and the lower core support plate during the LOCA event are shown in Figure 8-20 and Figure 8-21 as examples of the RV response analysis result.

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8-3 8.3 Component Loads of PZR (1) Seismic Loads Seismic loads of the PZR consist of component load and nozzle load. Component loads are shown in Table 8-16, is provided from dynamic analysis results of the PZR stick model described in section 7.1. Nozzle loads are shown in Table 8-17, is determined as design loads described in section 5.4.

(2) Accident Loads Accident loads of the PZR consist of nozzle load at postulated pipe rupture. Nozzle loads are shown in Table 8-18, is determined as design loads described in section 5.4.

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8-4 Table 8-1 List of Tables and Figures of RCL Seismic and Accident Loads Load Type Location Seismic Load Accident Load Coordinate System RV Inlet and Outlet Nozzle Figure 8-1 SG Inlet and Outlet Nozzle Figure 8-2 Nozzle Load RCP Inlet and Outlet Nozzle Table 8-2 Table 8-6 Figure 8-3 SG Lower Support SG Intermediate Shell Support SG Upper Shell Support Figure 8-4 Reaction Force of Support Point RCP Lower Support Table 8-3 Table 8-7 Figure 8-5 Hot Leg Cross Over Leg Member Forces of MCP Cold Leg Table 8-4 Table 8-8 Figure 8-6 RV Vent Nozzle RV DVI Nozzle Figure 8-7 SG MS Nozzle SG FW Nozzle Figure 8-8 RCP Seal Water Injection Nozzle Figure 8-9 MCP Pressurizer Surge Nozzle MCP Pressurizer Spray Nozzle MCP CVCS Letdown and Loop Drain Nozzle MCP Charging Nozzle MCP RHRS/SIS Nozzle MCP Accumulator Tank Nozzle Nozzle Load RHRS Return Nozzle Table 8-5 Table 8-9 Figure 8-10

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8-5 Table 8-2 Seismic Loads of Inlet and Outlet Nozzle on RV, SG, RCP Force (kips)

Moment (kips-in)

Node Location Coordinate System Fx Fy Fz Mx My Mz 194 RV Inlet Nozzle 107 RV Outlet Nozzle Figure 8-1 117 SG Inlet Nozzle 141 SG Outlet Nozzle Figure 8-2 167 RCP Inlet Nozzle 181 RCP Outlet Nozzle Figure 8-3 Table 8-3 Seismic Reaction Forces of Support Points Force (kips)

Moment (kips-in)

Node Location Coordinate System Fx Fy Fz Mx My Mz 125 SG Lower Support Point(1) 126 SG Lower Support Point(2) 127 SG Lower Support Point(3) 128 SG Lower Support Point(4) 132 SG Intermediate Shell Support Point 137 SG Upper Shell Support Point Figure 8-4 170 RCP Lower Support Point(1) 171 RCP Lower Support Point(2) 172 RCP Lower Support Point(3)

Figure 8-5

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8-6 Table 8-4 Seismic Member Forces of MCP Force (kips)

Moment (kips-in)

Node Location Coordinate System Fx Fy Fz Mx My Mz 107 Hot Leg 109 Hot Leg 111 Hot Leg 112 Hot Leg 113 Hot Leg 115 Hot Leg 117 Hot Leg 141 Cross Over Leg 142 Cross Over Leg 143 Cross Over Leg 145 Cross Over Leg 147 Cross Over Leg 149 Cross Over Leg 151 Cross Over Leg 152 Cross Over Leg 153 Cross Over Leg 155 Cross Over Leg 157 Cross Over Leg 159 Cross Over Leg 160 Cross Over Leg 161 Cross Over Leg 163 Cross Over Leg 165 Cross Over Leg 167 Cross Over Leg 181 Cold Leg 183 Cold Leg 185 Cold Leg 186 Cold Leg 189 Cold Leg 191 Cold Leg 194 Cold Leg Figure 8-6

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8-7 Table 8-5 Seismic Loads of Nozzles on RV, SG, MCP Force (kips)

Moment (kips-in)

Location Coordinate System Fx Fy Fz Mx My Mz RV Vent Nozzle RV DVI Nozzle Figure 8-7 SG MS Nozzle SG FW Nozzle Figure 8-8 RCP Seal Water Injection Nozzle Figure 8-9 MCP Pressurizer Surge Nozzle MCP Pressurizer Spray Nozzle MCP CVCS Letdown and Loop Drain Nozzle MCP Charging Nozzle MCP RHRS/SIS Nozzle MCP Accumulator Tank Nozzle RHRS Return Nozzle Figure 8-10

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-8 Table 8-6 Accident Loads of Inlet and Outlet Nozzle on RV, SG, RCP Force (kips)

Moment (kips-in)

Node Location Coordinate System Fx Fy Fz Mx My Mz 194 RV Inlet Nozzle 107 RV Outlet Nozzle Figure 8-1 117 SG Inlet Nozzle 141 SG Outlet Nozzle Figure 8-2 167 RCP Inlet Nozzle 181 RCP Outlet Nozzle Figure 8-3 Table 8-7 Accident Reaction Force of Support Points Force (kips)

Moment (kips-in)

Node Location Coordinate System Fx Fy Fz Mx My Mz 125 SG Lower Support Point(1) 126 SG Lower Support Point(2) 127 SG Lower Support Point(3) 128 SG Lower Support Point(4) 132 SG Intermediate Shell Support Point 137 SG Upper Shell Support Point Figure 8-4 170 RCP Lower Support Point(1) 171 RCP Lower Support Point(2) 172 RCP Lower Support Point(3)

Figure 8-5

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-9 Table 8-8 Accident Member Force of MCP Force (kips)

Moment (kips-in)

Node Location Coordinate System Fx Fy Fz Mx My Mz 107 Hot Leg 109 Hot Leg 111 Hot Leg 112 Hot Leg 113 Hot Leg 115 Hot Leg 117 Hot Leg 141 Cross Over Leg 142 Cross Over Leg 143 Cross Over Leg 145 Cross Over Leg 147 Cross Over Leg 149 Cross Over Leg 151 Cross Over Leg 152 Cross Over Leg 153 Cross Over Leg 155 Cross Over Leg 157 Cross Over Leg 159 Cross Over Leg 160 Cross Over Leg 161 Cross Over Leg 163 Cross Over Leg 165 Cross Over Leg 167 Cross Over Leg 181 Cold Leg 183 Cold Leg 185 Cold Leg 186 Cold Leg 189 Cold Leg 191 Cold Leg 194 Cold Leg Figure 8-6

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-10 Table 8-9 Accident Loads of Nozzles on RV, SG, MCP Force (kips)

Moment (kips-in)

Location Coordinate System Fx Fy Fz Mx My Mz RV Vent Nozzle RV DVI Nozzle Figure 8-7 SG MS Nozzle SG FW Nozzle Figure 8-8 RCP Seal Water Injection Nozzle Figure 8-9 MCP Pressurizer Surge Nozzle MCP Pressurizer Spray Nozzle MCP CVCS Letdown and Loop Drain Nozzle MCP Charging Nozzle MCP RHRS/SIS Nozzle MCP Accumulator Tank Nozzle RHRS Return Nozzle Figure 8-10

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-11 Table 8-10 Seismic and Accident Loads on Lower Reactor Internal Assembly Force (lbf)

Member Load Shear axial Moment (lbf-in)

Memo Seismic Accident Lower Core Support Plate SRSS Vertical Force on LCSP Seismic Accident Core Barrel Flange SRSS Seismic Accident Core Barrel Middle SRSS Seismic Accident Core Barrel Bottom SRSS CB/LCSP Discontinuity

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-12 Table 8-11 Seismic and Accident Loads on Upper Reactor Internal Assembly Force (lbf)

Member Load Shear axial Moment (lbf-in)

Memo Seismic Accident Upper Core Plate SRSS Vertical Direction Seismic Accident Upper Core Support (Plate)

SRSS Vertical Direction Seismic Accident Upper Core Support (Flange)

SRSS Seismic Accident Upper Core Support (Bottom of Skirt)

SRSS Seismic Accident Upper Core Support Column (Top)

SRSS Seismic Accident Upper Core Support Column (Bottom)

SRSS Seismic Accident Top Slotted Column (Top)

SRSS Seismic Accident Top Slotted Column (Bottom)

SRSS

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-13 Table 8-12 Seismic and Accident Loads on Radial Support Keys Force (lbf)

Member Load Horizontal Seismic Accident Radial Support Key SRSS Note: Horizontal Force per One Key Table 8-13 Seismic and Accident Loads between Vessel / Reactor Internals Interface Loads Force (lbf)

Member Load Horizontal Vertical Seismic Accident Core Barrel Flange I / F SRSS Seismic Accident Radial Support SRSS Seismic Accident USP Flange I / F SRSS Note: Contact surface angle for Horizontal Force by Accident Load on Core Barrel Flange is [112.5°]

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8-14 Table 8-14 Seismic and Accident Loads on RV Head Nozzles and IHP Lug Force (lbf)

Moment (lbf-in)

Member Load Fx Fy Fz Mx My Mz Seismic Accident CRDM Nozzle SRSS Seismic Accident ICIS Nozzle SRSS Seismic Accident TC Nozzle SRSS Seismic Accident RV Water Level Instrumentation Nozzle SRSS Seismic Accident IHP Lug SRSS Note: Loads for RV Head nozzles are defined at the elevation on the outer surface of the vessel head.

Table 8-15 Seismic and Accident Loads on CRDM Force (lbf)

Member Load Shear axial Moment (lbf-in)

Seismic Accident Bottom of Rodtravel Housing SRSS Seismic Accident Bottom of Latch Housing (Weld Junction)

SRSS

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-15 Table 8-16 Seismic Loads of PZR Component Forces (kips)

Moment (kips-in)

Part Node Fx Fy Fz Mx My Mz Upper Support Point 11 Lower Head 3

Skirt 111 Note : Location of Upper Support Point of Node 11 is presented in Figure 7-1.

Table 8-17 Seismic Loads of PZR Nozzles Forces (kips)

Moment (kips-in)

Part Coordinate System Fx Fy Fz Mx My Mz Surge Nozzle Spray Line Nozzle Safety Valve Nozzle Safety Depressurization Valve Nozzle Figure 8-23

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-16 Table 8-18 Accident Loads of PZR Nozzles Forces (kips)

Moment (kips-in)

Part Coordinate System Fx Fy Fz Mx My Mz Surge Nozzle Spray Line Nozzle Safety Valve Nozzle Safety Depressurization Valve Nozzle Figure 8-23

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-17 Figure 8-1 RV Nozzle Load Figure 8-2 SG Nozzle Load SG CL.HOT LEG CROSS OVER LEG CL.

Y Y

X X

Z Z

A A

RV RV Outlet Nozzle (107)

X X

Y Y

RV Inlet Nozzle (194)

Z Z

Y X

Z SG Inlet Nozzle (117)

Z Y

X SG Outlet Nozzle (141)

SG A-A

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-18 Figure 8-3 RCP Nozzle Load RCP Y

Z X

CROSS CL.

CL.COLD LEG Y

Z X

RCP Inlet Nozzle (167)

RCP Outlet Nozzle (181)

RCP Z

X Y

Z X

Y OVER LEG

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-19 Figure 8-4 Reaction Force of SG Support Points CL. HOT LEG Y

X SG (137)

SG Upper Shell Support Point CL. HOT LEG Y

X SG (132)

SG Intermediate Shell Support Point 36°30' 53°30' SG C HOT LEG L.

X X

X X

Y Y

Y Y

Z Z

Z Z

(128)

Point(4)

(125)

SG Lower Support (126)

SG Lower Support (127)

SG Lower Support Point(3)

Point(1)

Point(2)

SG Lower Support

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-20 Figure 8-5 Reaction Force of RCP Lower Support Points RCP Lower Support Point(2)

(171)

RCP Lower Support Point(1)

(170)

RCP Lower Support Point(3)

(172)

L.COLD LEG C

CROSS L.

OVER LEG C

X Y

Z X

X Y

Y Z

Z

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-21 Figure 8-6 MCP Member Forces

[NOTE] (1)+X-local axis is set in the coolant flow direction.

(2)Right-handed coordinate system is addopted.

(3)As a rule, piping load shows the load generated at the starting point of the downstream member among evaluation points.

X Y

Z X

Y Z

X Y

Z X

Y Z

X Y

Z X

Y Z

X Y

Z Y

Y Y

Y Y

Y Y

Y Y

Y Y

Y Y

Y X

X X

X X

X X

X X

X X

X X

X Z

Z Z

Z Z

Z Z

Z Z

Z Z

Z Z

Z 107 109 111 112 113 153 155 157 159 160 161 181 183 185 186 189 191 (194)

(117) 115 141 142 143 Z

X Y

145,147,149,151,152 165,(167)

Y X

Z 163 RCP SG Main Coolant Piping Global Layout HOT LEG COLD LEG CROSS OVER LEG RCP COLD LEG (Side View) 181 183 X

X Z

Z Y

Y Y

X Z

SG RV HOT LEG Y

Z Y

Z Z

115 111 109 107 X

X Y

X Y

X Z

Y Z

X (117) 112 (Side View) 113 X

Z Y

RCP SG CROSS OVER LEG 141 152 155 161 165 145 142 (167) 159 157 160 147 151 (Side View) 143 163 153 149 Z

Z Z Z Z Z

Z Z Z

Z Z

Z Z

Z Z

Z Z

X X

X X

X X

X X X

X X

X X

X X

X X

Y Y

Y Y

Y Y

Y Y

Y Y

Y Y

Y Y

Y Y

Y 1014 1007 1004 Y

Z X

X Y

Z Y

X Z

1014 1004 1007 Z

Y X

Y X

Z X

Y RV Y

X Z

1013 Y

X Z

1012 X

Y X

Y 1012 1013 Z

Z Z

X Y

Z 1005 Y

Z X

1005 On the other hand the point number added by ( ) shows the load generated at the end point of the upstream member.

Y Y

Y X

X X

X X

Z Z Z RV 186 191 185 (194) 189 X X X X X Z Z Z Z Z Y Y Y Y

Y X

Y Z

X Z

1015 1015 Y

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-22 Figure 8-7 RV Nozzles Z

X Y

Z X

Y Vent Nozzle DVI Nozzle

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-23 Figure 8-8 SG Nozzles Z

X Y

X Y

Z MS Nozzle FW Nozzle

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-24 Figure 8-9 RCP Nozzle Z

X Y

Seal Water Injection Nozzle

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-25 Figure 8-10 MCP Nozzles Z

X Y

Accumulator Tank Nozzle Z

X Y

Pressurizer Surge Nozzle Pressurizer Spray Nozzle Charging Nozzle Z

X Y

CVCS Letdown and Loop Drain Nozzle RHRS/SIS Nozzle RHRS Return Nozzle RCL piping circumferential direction is Z direction.

RCL piping circumferential direction is Y direction.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-26 Figure 8-11 RV and Reactor Internals ICIS NOZZLE UPPER CORE SUPPORT CORE BARREL FLANGE INLET NOZZLE SUPPORT COLUMN UPPER CORE BARREL SUPPORT PLATE RADIAL SUPPORT KEY LOWER CORE UPPER CORE PLATE TOP SLOTTED COLUMN OUTLET NOZZLE HOLD-DOWN SPRING RVWL NOZZLE THERMOCOUPLE NOZZLE CRDM NOZZLE Z

Y X

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-27 Figure 8-12 Locations of Stress Evaluation on Lower Reactor Internal Assembly Z

Y X

SUPPORT KEY DISCONTINUITY FLANGE/UPPER CB DISCONTINUITY UPPER/LOWER CB RADIAL DISCONTINUITY LOWER CB/LCSP CORE INLET HOLES LCSP LIGAMENTS BETWEEN

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-28 Figure 8-13 Locations of Stress Evaluation on Upper Reactor Internal Assembly Z

Y X

GUIDE TUBE HOLES TSC EXTENSION DISCONTINUITY UCS PLATE/SKIRT TSC BODY TSC BOTTOM FASTENER USC EXTENSION USC BODY USC BOTTOM FASTENER UCS PLATE LIGAMENTS BETWEEN DISCONTINUITY UCS FLANGE/SKIRT UCP PLATE LIGAMENTS BETWEEN CORE OUTLET FLOW HOLES

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-29 Figure 8-14 CRDM Assembly Bottom of Latch Housing (Weld Joint)

Bottom of Rod Travel Housing Z

Y X

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-30 Figure 8-15 Axial Force vs Elevation on CRDM (SSE+LOCA)

Figure 8-16 Shear Force vs Elevation on CRDM (SSE+LOCA)

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-31 Figure 8-17 Bending Moment vs Elevation on CRDM (SSE+LOCA)

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-32 Figure 8-18 Horizontal UCP Acceleration during SSE Figure 8-19 Horizontal LCSP Acceleration during SSE

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-33 Figure 8-20 Horizontal UCP Acceleration during LOCA Figure 8-21 Horizontal LCSP Acceleration during LOCA

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-34 Figure 8-22 Reaction Force of PZR Lower Elements 4

3 2

113 111 112 113 Global Cordinate Systems X

Y Z

PN Global Coordinate System

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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8-35 Figure 8-23 PZR Nozzles X

Y Z

X Y

Z Spray Line Nozzle Safety Valve Nozzle Safety Depressurization Valve Nozzle Z

X Y

Surge Nozzle

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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9-1 9.0 COMPUTER PROGRAM

- MULTIFLEX Code The MULTIFLEX computer code was used for the blowdown analysis in the primary system hydraulic load evaluation of the postulated pipe break. MULTIFLEX code is a computer program which calculates the transient of pressure, flow rate and density during the initial phase of the blowdown in a complex system such as the primary coolant system of a PWR.

MULTIFLEX code includes mechanical structure models and their interactions with thermal hydraulic system. The general characteristics of MULTIFLEX are shown in the following; (a) The complex system is modeled with one-dimensional hydraulic piping.

(b) The flow conditions within the system are calculated by solving the method of characteristics.

(c) MULTIFLEX includes heat transfer models of the core and the SG, and also simulates various boundary condition of the PWR system including the core.

The calculated results of MULTIFLEX (pressure, flow rate and so on) are used in the RV internals load evaluation and the RCL mechanical load evaluation. MULTIFLEX code includes the fluid-structure interaction model, which enable to simulate the effect of flow path area change due to the motion of the CB. During blowdown phase of the LOCA, the CB walls surrounding a hydraulic flow path deviate from their neutral position depending on the force differential on the wall. In the calculation, the wall displacements are represented by those of 1-dimensional mass points which are described by mechanical equations of vibration. Ten of the mass points are used for modeling of the CB.

MULTIFLEX code had been approved by NRC (Reference 16), and the analysis model described in this report is the same as that of US conventional plants evaluated by Westinghouse.

- M-RELAP5 Code The M-RELAP5 computer code was used for the SB-LOCA blowdown analysis. This code model is described in Reference 17.

The M-RELAP5 is a code developed by MHI based on RELAP5-3D code. One of the main improvements is an addition of Moody critical flow model. The Moody critical flow model is used in the calculation of the break flow out of the highly pressurized RV during the blowdown phase of the postulated pipe break. The Moody model is acceptable to critical flow phenomena, which is shown in ANSI/ANS-58.2-1988. (Reference 6)

RELAP5-3D code, which is the original source of M-RELAP5, has been developed by Idaho National Laboratory. RELAP5 is a highly generic code that, in addition to calculating the behavior of the RCS during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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9-2 The code models the coupled behavior of the RCS and the core for the LOCA and operational transients such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow.

The code includes many generic component models from which general systems can be simulated. The component models include pumps, valves, pipes, heat releasing or absorbing structures, reactor kinetics, electric heaters, jet pumps, turbines, separators, annuli, PZRs, feedwater heaters, ECC mixers, accumulators, and control system components. In addition, special process models are included for effects such as form loss, flow at an abrupt area change, branching, choked flow, boron tracking, and noncondensable gas transport.

A set of equation of RELAP5 gives a two-fluid system simulation using a nonequilibrium, nonhomogeneous, six-equation representation. The presence of boron and noncondensable gases is also simulated using separate equations for each. Constitutive models represent the interphase drag, the various flow regimes in vertical and horizontal flow, wall friction, and interphase mass transfer.

- GOTHIC Code The GOTHIC computer code was used for the asymmetric pressurization analysis. GOTHIC is a general purpose thermal-hydraulics code for performing design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings (Reference 18).

GOTHIC solves the conservation equations for mass, momentum and energy for multi-component, multi-phase flow in lumped parameter and/or multi-dimensional geometries.

The phase balance equations are coupled by mechanistic models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film/drop flow, as well as single phase flows. The interface models allow for the possibility of thermal non-equilibrium between phases and unequal phase velocities, including countercurrent flow.

GOTHIC includes full treatment of the momentum transport terms in multi-dimensional models, with optional models for turbulent shear and turbulent mass and energy diffusion.

The principal element of a model is a control volume, which is used to model the space within a building or subsystem that is occupied by fluid. The fluid may include non-condensing gases, steam, drops or liquid water. GOTHIC features a flexible nodal scheme that allows computational volumes to be treated as lumped parameter (single node) or one-, two-or three-dimensional, or any combination of these within a single model.

Flow paths model hydraulic connections between any two computational cells, which includes lumped parameter volumes and cells in subdivided volumes. Flow paths are also used to connect boundary conditions to computational cells where mass, momentum and energy can be added or removed. A separate set of momentum equations (one for each phase) is solved for each flow path.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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9-3 Initial conditions allow the user to specify the state of the fluid and solid structures within the modeled region at the start of a transient. These include the initial temperature and composition of the atmosphere.

Additional resources available to expand the realm of situations that can be modeled by GOTHIC include functions, control variables, trips and material properties.

Using a conservative model prescription, GOTHIC predicts the time dependent compartment differential pressure.

- ANSYS Code The ANSYS code is a general purpose finite element analysis program for linear and nonlinear, static and dynamic, structural, thermal, or other various problems, and the code has the excellent capability of pre-processing, solver, and post-processing with a comprehensive graphical user interface of the interactive access to program functions, commands, documentation, and reference material (Reference 19, 20).

ANSYS is used to model the RCL, RV and PZR to produce loads, displacements, accelerations, and stiffnesses for use in the structural analysis of those components.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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10-1

10.0 CONCLUSION

This report describes the methods used to analyze the effects of seismic and accident events on the primary components and piping, furthermore the results of analysis and lists the loads these events produce. Load values will be applied external mechanical load to stress analysis for components and piping. Stress analysis will be performed in accordance with ASME Code,Section III (Reference 2) requirements, and applied criteria of the service Level C and D service conditions. The combination of seismic load (SSE) and accident load (LOCA) will be set in conformance with SRP 3.9.3, Rev. 2 (Reference 21).

Based on these load combinations, technical reports will be shown summary of stress analysis results for primary components and piping and submit on Janualy 2011.

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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11-1

11.0 REFERENCES

1. Seismic Design Bases of the US-APWR Standard Plant. Mitsubishi Heavy Industries, Topical Report MUAP-10001 Rev.1, May 2010.

2. ASME Boiler and Pressure Vessel Code,Section III. 2001 Edition through the 2003 Addenda, American Society of Mechanical Engineers.

3. Summary of Stress Analysis Results for the US-APWR Reactor Coolant Loop Piping, Mitsubishi Heavy Industries, Technical Report MUAP-09010-P Rev.1, May 2009.

4. Summary of Stress Analysis Results for the US-APWR Reactor Coolant Loop Branch Piping, Mitsubishi Heavy Industries, Technical Report MUAP-09011-P Rev.2, December 2010.

5. Summary of Stress Analysis Results for the US-APWR Main Steam Piping inside Containment Vessel, Mitsubishi Heavy Industries, Technical Report MUAP-09013-P Rev.0, March 2009.

6. Design Bases for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture.

ANSI/ANS-58.2-1988, American National Standards Institute/American Nuclear Society.

7. Determination of Rapture Locations and Dynamic Effects Associated with The Postulated Rupture of Piping. NUREG-0800, SRP 3.6.2, Rev.2, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.

8. Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Other Dynamic Loads and Load Combinations. NUREG-1061, Section 2, Vol.4, U.S.

Nuclear Regulatory Commission, Washington, DC, December 1984.

9. Seismic System Analysis. NUREG-0800, SRP 3.7.2, Rev.3, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.

10. Seismic Subsystem Analysis. NUREG-0800, SRP 3.7.3, Rev.3, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.

11. Combining Modal Responses and Spatial Components in Seismic Response Analysis.

REGULATORY Guide 1.92, Rev.2, U.S. Nuclear Regulatory Commission, Washington, DC, July 2006.

12. Damping Values for Seismic Design of Nuclear Power Plants. Regulatory Guide 1.61, Rev.1, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.

13. Comparison of the Estimated Loads and Actual Loads for RCL Components and Piping Nozzles. Mitsubishi Heavy Industries, Technical Report MUAP-09015 Rev.1, LATER

Summary of Seismic and Accident Load Conditions for Primary Components and Piping MUAP-09002-NP (R2)

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11-2

14. Dynamic Testing and Analysis of Systems, Structures, and Components. NUREG-0800, SRP 3.9.2, Rev.3, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.

15. Comprehensive Vibration Assessment Program for US-APWR Reactor Internals.

Mitsubishi Heavy Industries, Technical Report MUAP-07027 Rev.1, May 2009, Subsection 3.2.1, Validation of Structure Models

16. Evaluation of Westinghouse Topical Reports WCAP-8708(P) and WCAP-8709(NP). Letter from John F. Stolz (NRC) to C. Elicheldinger (Westinghouse) dated JUN 17, 1977.

17. Small Break LOCA Methodology for US-APWR. Mitsubishi Heavy Industries, Topical Report MUAP-07013-P Rev.1, May 2010.

18. GOTHIC Containment Analysis Package User Manual. Version7.2a(QA), NAI 8907-02, Rev.17, Numerical Applications Inc., Richland, WA, January 2006.

19. ANSYS, Finite Element Structural Analysis Program, Release 11.0, ANSYS, Inc.,

Canonsburg, PA, 2007.

20. ANSYS, Finite Element Structural Analysis Program, Release 12.1, ANSYS, Inc.,

Canonsburg, PA, 2009.

21. ASME Code Class 1, 2 and 3 Components and Component Supports, and Core Support Structures. NUREG-0800 SRP Section 3.9.3, Rev.2, U.S. Nuclear Regulatory Commission, Washington, DC, March 2007.