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Info Paper Presented at 930513 Meeting on AP600 Piping Analysis
	ML19311B028
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Site:	05200003
Issue date:	05/13/1993
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$ l WESTINGHOUSE ELECTRIC CORPORATION i

i l PRESENTATION i

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! UNITED STATES l

NUCLEAR REGULATORY I

i COMMISSION ,

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WESTINGHOUSE ENERGY CENTER MAY 13,1993 ;

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Agenda l AP600 Piping Presentation to NRC j May 13,1993 l 8: 30 Introduction McIntyre i

8: 45 Piping analysis methodology Lindgren

. Analysis and design criteria for safety-related piping systems

. Analysis and design criteria for Class D l systems

. Leak-before-break approach for design

certification 9: 15 Analysis Status E. Johnson

. Scope of piping design included in analysis

. Level of detail in the analysis prepared

. Leak-before-break analysis status 9:45 Analysis work plan beyond Design Certification E. Johnson 10:15 Conceptual pipe layout of selected systems T. Johnson ;

l 11:30 Piping analysis ITAAC and design acceptance E. Johnson i criteria l i l

= pipe stress analyses

. high energy line break analysis

. leak-before-break analysis ;

l 12:00 Lunch 1:00 Potential piping open issues E. Johnson /

. Advance Reactor Corporation (ARC) piping Bagchi analysis issues

. RAI Responses

. SECY-93-087 Leak-Before-Break Position 2:30 Discussion All I 3:15 Action items / Comments All i

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m AP600 Piping Analysis Methodology L

Analysis and design criteria for safety-related piping systems t .

Analysis and design criteria for Class D systems L

Leak-before-break approach for design

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ik Safety-Related Piping Systems Safety-related piping is AP600 Equipment Class A, B, and C - ASME Class 1,2, and 3

. The AP600 Equipment Classification is defined in L SSAR Section 3.2.

Safety-related piping is analyzed to ASME Code, Section lli requirements.

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. Loading conditions, stress limits, and analytical i methods for the structural evaluation of the safety-

) related piping and supports for design conditions, normal conditions, anticipated transients, and postulated accident conditions are discussed in L Subsection 3.9.3.

[ . The design and evaluation requirements for the reactor coolant system are outlined in Subsection r

5.4.3.

. The reactor coolant system design transients for normal operation, anticipated transients and

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t postulated accident conditions are discussed in Subsection 3.9.1.

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F The reactor coolant loop analysis methods and re-suits are presented in Appendix 3C of the SSAR The integrated reactor coolant loop and supports f system n7odel is the basic system model used to compute leadings on components, component

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supports, and piping.

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The system model includes the stiffness and mass characteristics of the reactor coq %nt loop piping

[ and components, the stiffness of supports, and the stiffnesses of auxiliary line piping affecting the l

system.

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. This model is used to determine the static and dynamic loads on the primary loop piping and the component supports and the interfacing loads on the connecting components.

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l The AP600 reactor coolant loop (RCL) model consists

of three-dimensional finite elements such as pipes, i beams, elbows, masses, and springs of given l dimensions, sizes, and physical properties that j mathematically simulate the structural response of i the physical system.

t l . The system model contains the reactor pressure

! vessel, two steam generators, four reactor coolant-pumps, the containment interior building structure,

the reactor coolant loop piping, and the primary equipment supports.

The structural model is subjected to internal j' pressure, thermal expansion, weight and seismic i loadings with boundary conditions imposed on the l

! model. l t

. The finite element displacement method is used for the analysis.

I The stiffness matrix for each element is assembled into a system of simultaneous linear equations for l the entire structure. l

. The set of equations are solved by the wave-front technique variation of the Gaussian elimination method.

IN The reactor coolant loop piping is qualified in according to the requirements of the ASME Code, Section 111, Subsection NB.

. The 1989 Edition with 1989 Addenda is the baseline !

Code.

. The loadings for ASME Code, Section Ill, Class 1 components are defined in Subsection 3.9.3.

. The following loadings are considered in the !

reactor coolant loop piping analysis:

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. Design pressure !

j Weight j

. Thermal expansion - normal operating conditions i j Thermal expansion - other transient conditions l . Safe shutdown earthquake l j . The reactor coolant loop piping is analyzed for the l j dynamic effects of a safe shutdown earthquake. !

The damping values used for the reactor coolant j loop piping response spectra analysis are those in ASME Code Case N-411.

l When the reactor coolant loop is analyzed using the time history analysis method,4% damping is l

used.

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E j The requirements for seismic evaluation l

are outlined in SSAR Section 3.7.

! Seismic design of the AP600 seismic Category I ,

j systems and components is based on the safe j shutdown earthquake (SSE).

l . The safe shutdown earthquake is defined as the

! maximum potential vibratory ground motion at the

genenc plant site.

i, 1 j The operating basis earthquake is not used as a design requirement for the AP600.

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! . Low-level seismic effects are included in the design

! of certain equipment potentially sensitive to a l

number of such events based on a percentage of the response 3 calculated for the safe shutdown earthquake.

! Seismic analyses of the nuclear island are performed l in conformance with the guidelines provided in ASCE Standard 4-86 and the criteria within SRP 3.7.2.

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l The methods used for seismic analysis of subsys-l tems include, modal response spectrum analysis, l time- history analysis, equivalent static analysis, and

" design by rule." The method used is selected by the i designer as appropriate for a specific item.

l l The equivalent static load method involves equiva-lent horizontal and vertical static forces applied at j the center of gravity of various masses.

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. Design by rule is used for small diameter piping systems.

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ASME Code Class 1 piping equal to or less than one-inch nominal pipe size and ASME Class 2 and 3 piping with nominal pipe sizes less than or equal to two inches is analyzed using one of the following i methods.

i . The response spectra method used to evaluate )

j larger pipe systems.

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. Equivalent static analysis based on appropriate l

! load factors applied to the response spectra l acceleration values; or, l

. Seismic qualification by experienca is based on the l guidelines provided by EPRI Report NP-6628. ;

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1 l Analysis and Design Criteria i

for Clrss D Systems Class D is nonsafety-related with some additional

, requirements on procurement, inspection or !

monitoring. l l . For Class D structures, systems, and components ;

! containing radioactivity, it is demonstrated by l l conservative analysis that the potential for failure ;

j due to a design basis event will not result in ex- !

l ceeding the normal offsite doses per 10 CFR 20.

! This criterion is in conformance with the definition l of Class D in Regulatory Guide 1.26.

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. A structure, system or component is classified as

( Class D when it directly acts to prevent unnecessary actuation of the passive safety i systems. Structures, systems and components which are required to support those which directly i

{ act to prevent the actuation of passive safety i systems are also Class D.

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l Standard industrial quality assurance standards are )

] applied to Class D structures, systems, and compo- ,

j nents to provide appropriate integrity and function.

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. These standards are consistent with the guidelines :

for NRC Quality Group D.

i I The industry standards used for Class D structures, systems and components are widely used high ;

quality standards.

j . ASME Code, Section Vill for pressure vessels ,

. ANSI B 31.1, Code for Power Piping for piping.

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l The buildings containing Class D structures, sys-l tems, and components, as well as the anchorage of l the structures and components to the building, are l designed to the seismic requirements of the Uniform j Building Code.

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l . The Class D systems and components are normally j not designed for either the Uniform Building Code i or seismic Category I.

When Class D structures, systems, and l components are located near a Class A, B, or C structure, system, or component, the requirements i for seismic Category ll may apply.

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! E Piping Failure Protection i

Pipe failure protection is provided according to the 1 requirements of 10 CFR 50, Appendix A, General i Design Criterion 4. l In the event of a high- or moderate-energy pipe failure within the plant, adequate protection is :

l provided so that essential structures, systems, or i j components are not impacted by the adverse l effects of postulated piping failure. l

! . The criteria used to evaluate pipe failure protection i are generally consistent with NRC guidelines '

including those in the Standard Review Plan

! Sections 3.6.1,3.6.2 and NUREG-1061 and l applicable Branch Technical Positions. l i !

l . Dynamic effects of postulated breaks are evaluated :

l except for those analyzable sections of high-energy -

l piping systems that are eliminated by the use of ,

mechanistic pipe break methods.

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l Piping Failure Protection Requirements

. Subsection 3.6.1 provides the design bases and l criteria for the analysis required to demonstrate that essential systems are protected. l l

The high- and moderate-energy systems ;

i representing the potential source of dynamic l l effects are listed. !

j The criteria for separation and the effects of l adverse consequences are defined.

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l . Subsection 3.6.2 defines the criteria for postulated

! break location and configuration.

i High-energy pipes are evaluated for the effects of ;

l circumferential and longitudinal pipe breaks and !

through-wall cracks. l

. Moderate-energy pipes are evaluated for the I l effects of through-wall cracks. l i

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Evaluation of the dynamic effects of postulated !

breaks in selected piping systems is not required for ;

AP600 based on application of mechanistic pipe break (leak-before-break) considerations. l 1 :

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. Subsection 3.6.3 describes the application of leak- l before-break criteria to permit the elimination of i

! pipe rupture dynamic effects considerations. :

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. Design guidelines aid in the design of piping 4 systems that satisfy the requirements for mecha-

! nistic pipe break.

. The systems considered for application of leak-

! before break criteria include the reactor coolant !

j loop, reactor coolant loop branch lines, main steam j

and feedwater lines out to anchors adjacent to the !

isolation valves, and other primary and secondary ;

system piping inside containment equal to or l I

greater than the four-inch nominal pipe size. ;

. Candidate systems for application of leak-before-

. break criteria are identified in Appendix 3E. This I

appendix was provided with the response to RAI Question 210.6.

. Respmses to RAls 252.5 and 252.6 provide l clarMication for the application of leak-before-break i cr!ieria.

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( Mechanistic pipe break evaluations demonstrate that for piping lines meeting the criteria, sudden cata-

[ strophic failure of the pipe is not credible.

[ . It is demonstrated that piping that satisfies the criteria leaks at a detectable rate from postulated

[ flaws prior to growth of the flaw to a size that would fall because applied loads resulting from

[ normal conditions, anticipated transients, and a ;

postulated safe e - utdown earthquake. i

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. The use of met v iistic pipe break criteria repre-sents a higher luel of confidence of the integrity of piping systems based on additional criteria compared to the existing high level of integrity provided by the requirements of the ASME Code.

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Piping systems to which mechanistic pipe break are i

applied are high integrity systems with well understood loading combinations and conditions.

! . The piping systems to which it is applied satisfy i j the requirements of the ASME Code, Section lll for l Class 1 and 2 systems. ASME Code requirements l l also apply to the pre-service and in-service inspec- l j tion which confirm continued integrity.

4 l . The mechanistic pipe break approach is applicable to high-energy piping provided plant design, operating experience, tests, or analyses have

indicated low probability of failure from effects of i intergranular stress corrosion cracking, water l hammer, steam hammer, fatigue (thermal or j mechanical), or erosion.

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PIPING ANALYSIS STATUS

[ DESIGN CERTIFICATION

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[ 1.0 Piping Analysis Activities

[ 2.0 Piping Analysis Details 3.0 Leak-Before-Break

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! 1.0 Piping Analysis Activities l

l Preliminary pipe stress and leak-before-break j analysis for sample systems. 1 i

i i Primary loop piping ,

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! - 2 soil cases - results provided in SSAR j Appendix 3B :

) - enveloping cases - analysis in progress, !

l 12/93 schedule i :

i Pressurizer Safety and Relief Valve Piping l l

l - Scope - 14", 8", 4" - Automatic l Depressurization System (ADS)- Stage l 1,2 and 3 j -

6" Pressurizer Safety 1

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Schedule - 12/93 i

l Main steam Piping l -

Scope - 32" - inside containment, and main steam isolation valve compartment I -

Schedule - 12/93

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i i l Feedwater Piping ;

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Scope - 16" - inside containment and !

! feedwater isolation valve compartment ;

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- Schedule - 12/93 i

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! 2.0 Piping Analysis Details Preliminary Pipe Stress Analysis Methodology 3

- provided in SSAR and responses to RAl's s - followed to extent practical based on status of design data

! Preliminary Pipe Stress Analysis Approach l -

Goal - define pipe layout, pipe support steel i arrangement and module support steel l arrangement for the plant model.

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Loadings - thermal expansion, SSE and other

! limiting loadings I

l Structural Models - approximations are used l based on status of design data pipe support stiffness l

module steel stiffness l

branch line coupling L -

equipment and valve properties seismic response spectra I

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l Structural Analysis - static, SSE response l spectra, SSE time history

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Results

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ASME pipe stress evaluation equipment and valve interface loads pipe loads for leak-before-break leak-before-break screening criteria I

Analysis Iterations

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adjustments to pipe layout and support arrangement to achieve acceptable design

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! 3.0 Leak-Before-Break l i

i i Objective - provide sample analyses that cover a range of materials, fluid conditions and pipe I j sizes j Methodology provided in SSAR and responses to RAl's l l I l Primary Loop Piping 1

i - sample analysis described in SSAR i l Appendix 3B ,

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l Auxiliary Piping 1

{ - pressurizer safety and relief, mainsteam and

! feedwater l -

materials ,

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SA 376 - 316 LN l l

SA 333 - Gr 6 1

Alloy Steel :

fluid conditions i j - subcooled liquid (primary and secondary l l system) l

- saturated steam (primary system) l dry steam (secondary system) ;

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pipe sizes l l

l 4,6,8,14,16 and 32 inches schedule 12/93

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WORK PLAN BEYOND DESIGN CERTIFICATION 1.0 SCOPE FOAKE (THROUGH 9/96)

POST-FOAKE 2.0 WORKPLAN l 3.0 LEAK-BEFORE-BREAK D

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Pipe Stress Analysis Scope 1.0 Safety Related Piping ;

1.1 All ASME Class 1 piping 1.2 All ASME Class 2/3 piping inside containment (includes B31.1 extension) 1.3 ASME Class 2/3 piping outside containment that are extensions of major B31.1 piping

( > 150 F) where: Major B31.1 piping includes

! 1) all large bore

2) small bore extensions to containment isolation valves l 1.4 High energy ASME Class 2/3 piping outside containment l

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j 2.0 Non-Safety (B31.1) Piping ( > 150 F) SSE j analysis if Category Il 2.1 All large bore piping inside containment

!l l 2.2 All large bore piping outside containment l

) 2.3 Small bore piping inside containment in j compartments that have safety-related j components l 2.4 Small bore piping outside containment in compartments that have containment 3 isolation valves i

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i l Pipe Support Steel and Module Steel Stress Analysis l Scope i

. Large Bore Pipe Support Steel and j Corresponding Module Steel

- high energy piping: safety related and l j non-safety related '

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SUBJECT:

INFORMATION PRESENTED AT THE MAY 13,1993 MEETING TO DISCUSS THE AP600 PIPING ANALYSIS

Dear Mr. Borchardt:

Please find enclosed with this letter information presented by Westinghouse at a May 13,1993, Westinghouse /NRC meeting on the AP600 Piping Analysis. The purpose of the meeting was to review the piping analysis methodology, analysis status and analysis planning for upcoming analysis.

Conceptual piping layouts, piping system ITAACs, and potential piping analysis related open items.

Enclosure 1 provides a list of meeting attendees. Enclosure 2 identifies the information provided during the meeting and provided as further enclosures to this letter.

Please contact Donald A. Lindgren on (412) 374-4856 if you have any questions concerning this transmittal, N. J. Liparulo, Manager d.

Nuclear Safety and Regulatory Activities

/nja Enclosures Attachments cc: T. Kenyon, NRC (w/o Enclosures / Attachments)

R. Hasselberg, NRC B. A. McIntyre, Westinghouse (w/o Enclosures / Attachments) 170119 9MNMbb ADOCK 05200003 0}

PDR )

A PDR

ML19311B028

Text

SUBJECT:

Dear Mr. Borchardt:

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