ML22112A042

Abb System 80+ Design Control Document - Volume 4

ML22112A042
Person / Time
Site:	LaSalle, 05200002
Issue date:	01/31/1997
From:	ABB Combustion Engineering
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20148A597	List: LD-97-014, Transmits Final Version of Sys 80+ Design Control Document (DCD) LD-97-015, Informs That ABB-CE Has Completed Review of Effective Page Listing & Sys 80+ Design Control Document Revs Dtd Jan 1997 ML20148A821 ML22112A037 ML22112A038 ML22112A039 ML22112A042 ML22112A043 ML22112A044 ML22112A045 ML22112A046 ML22112A048 ML22112A049 ML22112A050 ML22112A051 ML22112A052 ML22112A075 ML22112A076 ML22112A077 ML22112A078 ML22112A079 ML22112A081 ML22112A085 ML22112A086 ML22112A088
References
NUDOCS 9705090171
Download: ML22112A042 (1)

Text

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V the System 80+

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J y Design ControlDocument l

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Volume 4 l

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i Combustion Engineering, Inc.

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O Copyright C 1997 Combustion Engineering, Inc.,

All Rights Reserved.

Warning, Legal Notice and Disclaimer of Liability The design, engineering and other information contained in this document have been prepared by or for Combustion Engineering, Inc. in connection with its application to the United States Nuclear Regulatory Commission (US NRC) for design certification of the System 80+ nuclear plant design pursuant to Title 10, Code of Federal Regulations Part 52. No use of any such information is authorized by Combustion Engineering, Inc.

except for use by the US NRC and its contractors in connection with review and approval of such application. Combustion Engineering, Inc. hereby disclaims all responsibility and liability in connection with unauthorized use of such informaSon.

Neither Combustion Engineering, Inc. nor any other person or entity makes any warranty or representation to any person or entity (other than the US NRC in connection svith its review of Combustion Engineering's application) conceming such information or ito use, i except to the extent an express warranty is made by Combustion Engineering, Inc. to its customer in a written contract for the sale of the goods or services described in this document. Potential users are hereby wamed that any such information may be unsuitable for use except in connection with the performance of such a written contract by Combustion Engineering, Inc.

Such information or its use are subject to copyright, patent, trademark or other rights of Combustion Engineering, Inc. or of others, and no license is granted with respect to such rights, except that the US NRC is authorized to make such copies as are necessary for the use of the US NRC and its contractors in connection with the Combustion Engineering, Inc. application for design certifidon.

Publication, distribution or sale of this document does not constitute the performance of engineering or other professional services and does not create or establish any duty of care towards any recipient (other than the US NRC in connection with its review of l

Combustion Engineenng's application) or towards any pe, son affected by this document. '

For information address: Combustion Engineering, Inc., Nuclear Systems Licensing, 2000 Day Hill Road, Windsor, Connecticut 06095

System 80+ Design ControlDocument j' Introduction v

Certified Design Material 1.0 Introduction 2.0 System and Structure ITAAC i

3.0 . Non-System ITAAC 4.0 ' Interface Requirernents 5.0 Site Parameters Approved Design Material - Design & Analysis 1.0 General Plant Description 2.0 Site Characteristics 3.0 Design of Systems, Structures & Components 4.0 Reactor -

5.0 RCS and Connected Systems 6.0 Engineered Safety Features 7.0 Instrumentation and Control 8.0 Electric Power 9.0 Auxiliary Systems 10.0 Steam and Power Conversion 11.0 Radioactive Waste Management O 12.0 Radiation Protection 13.0 Conduct of Operations l 14.0 Initial Test Program 15.0 Accident Analyses 16.0 Technical Specifications 17.0 Quality Assurance 18.0 Human Factors .;

19.0 Probabilistic Risk Assessment 20.0 Unresolved and Generic Safety Issues Approved Design Material - Emergency Operations Guidelines  !

I I

1.0 Introduction 2.0 Standard Post-Trip Actions 'l 3.0 Diagnostic Actions i l

4.0 P.cactor Trip Recovery 5.0 Loss of Coolant Accident Recovery 6.0 . Steam Generator Tube Rupture Recovery 7.0 Excess Steam Demand Event Recovery 8.0 Loss of All Feedwater Recovery  ;

9.0 Loss of Offsite Power Re:overy j 10.0 Station Blackout Recovery 11.0 . Functional Recovery Guideline

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System'80 + oesign controloocument 3.5 Missile Protection The missile protection design for Seismic Category I structures, systems and components is described in this section.

Missile protection or redundancy is provided for Seismic Category I equipment and components such that internal and external missiles will not cause the release of significant amounts of radioactivity or prevent the safe and orderly shutdown of the reactor.

The protection of essential structures, systems and components will be accomplished by one or more of the following: ,

e Reducing the potential for sources of missiles by equipment design features that prevent missile generation.

o Orientation or physical separation of potential missile sources away from safety-related equipment and components.

e Containment of potential missiles through the use of protective shields and barriers near the source. ,

o Hardening of safety-related equipment and components to withstand missile impact, where such impacts cannot be reasonably avoided by the methods above.

(( Site and plant specific missile protection information regarding potential missiles and protection thereof will be provided by the COL applicant referencing the System 80+ design. Information will include site proximity missiles, probabilistic evaluations for turbine missiles, and as-built conditions and evaluations.))

3.5.1 Missile Selection and Desedption J

Potential missiles are identified and characterized by type and source and their probability of occurrence, retention and impact. For equipment with energy sources capable of creating a missile, the selection is based on the application of a single-failure criterion to the retention features of the component. Where sufficient retention redundancy is provided in the event of a failure, no missile is postulated.

Internally generated missiles can be generated potentially from two types of equipment: rotating components and pressurized components. Rotating components include turbine wheels, fans, auxiliary pumps and their associated motors. Pressurized components include valves, heat exchangers, vessels and their associated components.

The types of missiles considered and/or not considered in the design of Seismic Category I structures, systems, and components are discussed in the following sections:

e Internally Generated Missiles (Outside Containment), described in Section 3.5.1.1.

o Containment Internal Missiles, defined in Table 3.5-1 and Section 3.5.1.2.

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l System 80+ Design ControlDocument  !

• Turbine Missiles, described in Section 3.5.1.3.

• Natural Phenomena (Tornado) Missiles, described in Section 3.5.1.4.

• Site Proximity Missiles (Except Aircraft), described in Section 3.5.1.5.

• Aircraft Hazards, described in Section 3.5.1.6.

3.5.1.1 Internally Generated Missiles (Outside Containment)

Internally generated missiles (outside containment) from rotating and pressurized components are et considered credible for the reasons discussed below.

The redundant safety systems outside of containment are physically separated such that no single gravitational or other type missile can impact both systems.

3.5.1.1.1 Auxiliary Pumps and Motors There are no postulated missiles originating from auxiliary pumps and associated motors outside containment for the following reasons:

• The pump motors are induction type which have relatively slow running speeds and are not prone to overspeed. The motors are all pretested at full running speed by the motor vendor prior to installation.

• In addition to the low likelihood of missiles due to motor overspeed, the motor stator would tend to serve as a natural container of rotor missiles.

• All pumps normally have relatively low suction pressures and, therefore, would not tend to be driven to overspeed due to a pipe break in the discharge line. In addition, the induction motor would tend to act as a brake to prevent pump overspeed.

• Industry pump designs are such (and service history records confirm) that there have been no occurrences of impeller pieces penetrating pump casings.

3.5.1.1.2 Emergency Feedwater Pump Turbines There are no postulated missiles from the Emergency Feedwater (EFW) pump turbines for the following reasons:

• Turbine overspeed protection; electrical trip at 115 % of rated speed, and mechanical trip at 125 %

of rated speed.

• Assurance of turbine disk integrity by design and inspection.

• Enclosure of the EFW pumps and turbine drivers in a reinforced concrete room.

O Approved Desiger nistories Deshers of SSC Pays .T.5 2

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I System 80+ Deslan ContralDoce_nent

'3.5.1.1.3 Vaives -

There are no missiles postulated from valves for the following reasons:

e~ All valve stems are provided with a beheat or shoulder larger than the valve bonnet opening.  ;

I e Motor operated and manual valve stems are restrained by stem threads, a

! o - Operators 'on motor, hydraulic and pneumatic operated valves prevent stem ejection.  ;

e Pneumatic operated diaphragms and safety valve stems are restrained by spring force. .;

e All valve bonnets are either pressure sealed, threaded or bolted such that there is redundant  !

retention for prevention of missile generation.  ;

l Pressure Vessels  ;

3.5.1.1.4

! All pressurized vessels outside containment are moderate energy (275 psig) or less and are designed and j constructed to the standards of the ASME Code, in addition to the ASME Code examination and testing  ;

requirements, all vessels will receive periodic in-service inspections. Where appropriate, these  !

. components are provided with pressure relief devices to ensure that no pressure buildup will exceed material design limits.  ;

i A On this basis, moderate energy pressure vessels are not considered credible missile sources.  ;

ib i 3.5.1.2 Internally Generated Missiles (Inside Cantain===t)

I Table 3.5-1 lists postulated missiles from equipment inside containment, and summarizes their t

characteristics. Included are major pretensioned studs and nuts, instruments, and the CEDM missile, 4

Other items which were considered and specifically excluded because of redundant retention features are e valve stems, valve bonnets and pressurized cover plates. ,

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. 3.5.1.3 Turbine Missiles  ;

i The probability of turbine missile generation and adverse impact effects on Seismic Category I systems and components is assured to be less than 1.0E-4 events per turbine-year by a combination of the j

! i following measures:

e Reliable turbine overspeed protection provisions (see Section 10.2.2 for details). f e' Adequate assurance of turbine disc integrity by design and inspection (see Sections 10.2.2 and 10.2.4 for details), i e- Placement and orientation of the turbine generator (described below).

e The protection provided by plant structures, not explicitly designed as barriers, that may reduce [

missile energy to less than that required to penetrate Seismic Category I structures. j O% e Adequate turbine maintenance and inspection program (see Section 10.2.4 for details). l

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System 80+

Design ControlDocument The turbine generator placement and orientation for the System 80+ Standard Design, and the corresponding low-trajectory missile strike zones, are illustrated in Figure 1.2-1. The placement and orientation of the turbine generator provides adequate protection against low trajectory turbine missiles by excluding safety-related structures, systems, and components from the low trajectory turbine missile strike zones in accordance with the guidelines of Regulatory Guide 1.115.

Critical structures (i.e., those housing safety-related equipment) and exterior equipment are located in line with, or within close proximity to, the longitudinal axis of the turbines. This makes the potential for turbine-generated missiles to strike these targets negligibly small.

The System 80+ design follows the guidelines of Regulatory Guide 1.115 by placing and orienting the turbine such that all safety-related structures, systems, and components are excluded from the low trajectory turbine missile strike zones or if site characteristics make this impossible, safety-related targets will be placed and shielded such that the combined strike and damage probability for the safety-related targets in these zones is less than 10E-3 per turbine failure.

Site-specific evaluations will verify that the turbine maintenance and inspection program will ensure that the failure and missile generation probability will be less than 1.0E-4 events per turbine-year. A summary of the turbine maintenance and inspection program and the results of the probabilistic evaluation will be submitted for review.

3.5.1.4 Missiles Generated by Natural Phenomena Tomado-generated missiles are the limiting natural hazard and, as such, are a part of the design basis for Seismic Category I structures and components. Tornado-generated missiles considered in the design are given in Table 3.5-2.

3.5.1.5 Missiles Generated by Events Near the Site Justification will be provided in the site-specific SAR.

3.5.1.6 Aircraft IIazards Justification will be provided in the site-specific SAR. Also refer to Section 2.2.1.

3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles Tornado missiles are the design basis missiles from external sources. All safety related systems, equipment and components required to safely shut the reactor down and maintain it in a safe condition are housed in Seismic Category I structures designed as tornado resistant (see Section 3.5.1.4) and as such are considered to be adequately protected.

3.5.3 Barrier Design Procedures Missile barriers, whether steel or concrete, are designed with sufficient strength and thickness to stop postulated missiles and to prevent overall damage to Seismic Category I structures. The procedures by which structures and barriers are designed to perform this function are presented in this section.

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System 80+ oestan conuot Docanent 3.5.3.1 Local Damage Prediction ]

. The prediction of local damage in the immediate vicinity of an impacted area depends on the basic l material of construction of the barrier itself(i.e. either concrete or steel). Corresponding procedures are j discussed separately below. l 3.5.3.1.1 Concrete Structures and Barriers

' Local damage prediction for concrete stmetures includes the estimation of the depth of missile penetration and an assessment of whether secondary missiles might be generated by spalling. Generally, the Modified 4

Petry Formula or the Modified NDRC Formula (References 2 and 3) is used to estimate missile j penetration with appropriate constants taken from available test data. To insure that no secondary l 4 missiles (due to spalling) are generated, a minimum barrier thickness of 3 times the penetration depth is  ;

provided. In addition, the minimum barrier thickness requirements for local damage due to tornado generated missiles shall be as indicated in Table 3.5-3. Depending on certain missile characteristics, additional penetration formulas may be employed as justified by full scale impact tests (References 3 and 4). .

t 3.5.3.1.2 Steel Structures and Barriers The Stanford equation (Reference 5) is used as the basis for the design and analysis of steel structures  ;

and barriers.

3.5.3.2 Overall Damage Prediction $

4 O The overall response of a structure or barrier to missile impact depends largely on the location ofimpact (e.g. near mid-span or near a support), the dynamic and deformation properties of the barrier and the '

• missile, and the kinetic energy of the missile itself.

Depending on the deformation characteristics of both the barrier and the missile, an impact force time '

history can be developed using either work-kinetic energy principles or conservation of momentum. The structural response to this impulse loading, in conjunction with other appropriate design loads, is evaluated by the procedures given in References 3 and 6. ,

3.5.4 General Design Bases i

Protection for all Seismic Category I structures, systems and components are provided by the following: ,

• For systems and parts of systems located inside the containment (RCS and connected systems,  !

4 Engineered Safety Feature systems), appropriate missile barrier design procedures are used to ,

^

4 ensure that the impact of credible potential missiles will not lead to a loss-of-coolant-accident or preclude the systems from carrying out their specified safety functions.

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• For systems and equipment outside containment, appropriate design procedures (e.g., proper turbine orientation, natmal separation, or missile barriers) are used to ensure that the impact of credible potential missiles does not prevent the system or equipment from carrying out its specified safety function.  :

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System 80+ Design ControlDocument

• For all systems and equipment, appropriate design procedures are used to ensure that the impact of credible potential missiles does not prevent the conduct of a safe plant shutdown, or prevent the plant from remaining in a safe shutdown condition.

• Safety-related instrumentation and control equipment are protected from potential missile sources.

The IE and associated cabling and sensing lines are also protected from potential missile sources.

Site-specific evaluations will ensure that as-built conditions provide Seismic Category I structures, systems and components protection from credible potential missiles.

References for Section 3.5

1. " Plant Design Against Missiles," ANSI /ANS-58.1. (DRAFT - Formerly ANSI N177-1974.)

2. A. Amirikan, " Design of Protective Structures," Report No. NT-3726, Bureau of Yards and Docks Dept. of the Navy, August 1950.

3. " Structural Ar.alysis and Design of Nuclear Power Plant Facilities," Manual No. 58, Chapter 6, American Society of Civil Engineers,1980.

4. Stephenson, A. E., " Full Scale Tornado-Missile Impact Tests," EPRI NP-440, July 1977, Prepared for the Electric Power Research Institute by Sandia National Laboratories.

5. Cottrell, W. B. and A. W. Savolainen, "U.S. Reactor Containment Technology," ORNL-NSIC-5, Vol.1, Chapter 6, Oak Ridge National Laboratory.

6. Williamson, R. A. and R. R. Alvy, " Impact Effects of Fragments Striking Structural Elements," l Ilolmes and Narver, Inc., Revised November,1973. l l

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l System 80+ oesign controlDocument O Table 3.5-1 Kinetic Energy of Potential Missilestil Q i Maximum Impact Kinetic Weight l Item ]t2 Energy (ft-Ib) (Ib) Impact Sectlen i l

Reactor Vessel l Closure Head Nut 1,706 100 Annular Ring. OD = 10.125" )

ID = 6.9" j I

Closure Head Nut and Stud 5,226 655 Solid Circle,6.75" Diameter Control Rod Drive Assembly 1.875" Diameter Solid Circle 57,600 1,100 within a Concentric 7" I Diameter by 0.109" Wall Shroud '

l Hil'C Assembly 57,600 500 5.0" Diameter Solid Circle j j

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Steam Generator Primary Manway i Stud and Nut 71 5.7 Solid Circle,1.625" Diameter Secondary Handhole 7 1.5 Solid Circle,1" Diameter Stud and Nut i l

3.2 Solid Circle,1.5" Diameter V Secondary Manway Stud 17 l l

Pressudzer l Safety Valve Flange Bolt 80 10 Solid Circle,2" Diameter lower Temperature Element 2,377 7.6 Edge of Solid Disk 4" Diameter and 4" Thick Manway Stud and Nut 71 6.3 Solid Circle,1.5" Diameter l Reactor Coolant Pump and Piping Temperature Nozzle 2,123 8 Edge of Solid Disk 4" with RTD Assembly Diameter and 4" Thick Surge and Spray 2,344 7 Edge of Solid Disc 4" Piping Thermowells Diameter and 4" Thick with RTD Assembly Reactor Coolant Pump Edge of Solid Disk 2.75" Thermowell with RTD 1,095 8 Diameter and 0.5" Thick i

l 181 All dimensions, weights and kinetic energies are typical values.

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Table 3.5-2 Design Basis Tornado Missiles and their Impact Velocities Design Impact I Velocity (ft/sec)

Missile Weight Impact IIorizontal Vertical Descriptions' Dimensions (Ibs) Area (in')

A Wood Plank 3.6" x 11.4" x 12' i15 41 272 191 B 6" Sch. 40 Pipe 6.6"D x 15' 287 34 171 119 C l' Steel Rod  !"D x 3' 8.8 0.79 167 117 D Utility Pole 13.5"D x 35' 1124 143 180 126 E 12 Sch. 40 Pipe 12.6"D x 15' 750 125 154 108 F Automobile 6.56' x 4.27' x 16.4' 3990 4030 194 136 Table 3.5-3 Minimum Acceptable Barrier Thickness Requirements for Local Damage Prediction Against Tornado Generated Missiles Concrete Wall Roof Strength Thickness Thickness (psi) (inches) (inches) 4000 20 16 5000 18 14 3

Missiles A, B, C, and E are to be consideird at all elevations and missiles D and F at elevations up to 30 feet above all grade levels within 1/2 mile of the structure.

Approved Design Material Design of SSC page 3.5.g

System 80+ oesign controlDocument 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping

((The COL applicant referencing the System 80+ Standard Design will provide final designs of high and moderate energy fluid systems. Information provided will include documentation of radiographic examination of welds, verification of leak-before-break (LBB) analyses, a pipe break analysis report, and a LBB evaluation report.))l High-energy piping systems not approved for LBB are subject to postulated pressure boundary failures.

The resultant consequences of these postulated breaks are assessed for their effect on maintenance of plant safe shutdown capability, containment integrity and offsite dose consequence. Consideration is given to the effects of pipe whip, jet impingement and the environmental impact of release of system contents.

Provisions of physical separation, system redundancy, component strength, and, as necessary, mitigating hardware, are utilized to protect against the effects associated with postulated pipe breaks.

Protection of vital equipment is achieved primarily by separation of redundant safe shutdown systems and by separation of high-energy pipe lines from safe shutdown systems, which are required to be functional following specific pipe rupture events. This redundancy and separation results in a design which requires very few special protective features (such as whip restraints and jet deflectors) to ensure safe shutdown '

capability following a postulated high-energy line break.

Separation is maintained by barriers such as the containment secondary shield wall, refueling cavity wall and certain Nuclear Annex walls and tunnels or by physical distance. Loadings andjet zones ofinfluence are calculated using methodology described in Section 3.6.2.

Supplemental information on design for protection against dynamic effects of postulated pipe breaks is given in Appendix 3.6A.

3.6.1 Postulated Piping Failures in Fluid Systems 3.6.1.1 Design Basis Most systems and components outside Containment required for safe plant shutdown are located in the Reactor Building Subsphere. The Reactor Building Subsphere and Nuclear Annex are divided by a structural wall which serves as a barrier between redundant trains ef safe shutdown systems and components. Each half of the Reactor Building Subsphere is compartmentalized to separate redundant safe shutdown components to the extent practical. High-energy piping ystems located in the Nuclear Annex, which are not required to be functional for safe shutdown, are rout ed primarily in designated pipe tunnels or in the Main Steam Valve Houses to provide separation from safe shutdown systems and components. The Reactor Building Subsphere and Nuclear Annex are separated by structural walls that provide physical barriers.

Systems and components inside containment, which are required to be functional for safe plant shutdown, are protected from postulated pipe failure dynamic effects primarily by separation and barriers. The secondary shield wall serves as a barrier between the reactor coolant loops and the containment liner.

The refueling cavity walls, the operating floor, and the secondary shield wall provide separation O

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System 80+ Design ControlDocument between the reactor coolant loops. The steam generators and pressurizer are enclosed in cavities which also provide separation.

Main steam, steam generator blowdown, and main feedwater (downcomer and economizer) lines outside containment are separated from essential systems and components by virtue of the plant arrangement that places these lines along the roof of the Nuclear Annex. The floors and walls adjacent to the main steam, steam generator blowdown, and main feedwater lines are Seismic Category I concrete walls. The essential portions of these systems (main steam and main feedwater isolation valves) are located in the Main Steam Valve Houses. These rooms are separated from all other essential systems and components by Seismic Category I concrete slabs and walls. (Refer to Figures 1.2-2,7,8 and 9.)

The steam supply line to the turbine driven feedwater pump is routed from upstream of the main steam isolation valves in the Main Steam Valve House through a pipe chase into the turbine driven feedwater pump room. The pipe chase and feedwater pump room walls provide protection for other safety related equipment in the event of a rupture in the steam supply line. The pipe chase is open ended in the Main Steam Valve House to provide a vent path for the turbine driven feedwater pump room.

The Chemical and Volume Control System high-energy lines are located inside the primary containment and extend through the annulus into the Naclear Annex. The high-energy lines in the Nuclear Annex are 2-inch lines. The postulated dynamic effects of the Chemical and Volume Control System high-energy lines are separated from safe shutdown systems and components by distance and configuration as much as practical. Otherwise, protection is provided for by shields and barriers. The safe shutdown components are divisionally separated and divided by .: Seismic Category I concrete divisional wall in the Nuclear Annex. In the unlikely event that a postulated dynamic event were to effect a safe shutdown component, the redundant equipment associated with the other division would still be available for safe shutdown.

Any high-energy line routed through the annulus between the primary containment and its shield building is provided with a guard pipe so that rupture of high-energy lines in the annulus need not be analyzed.

The NSSS design includes two steam generators per unit, which facilitates separation of redundant systems and components inside containment. Other than for the safety injection system components, which must circulate cooling water to the vessel, the engineered safety features are generally located outside the secondary shield wall. The safety injection system pipes and cables, which terminate inside the second f shield wall, are routed outside the secondary shield wall to the extent practical to avoid postulated I azards. Most of the main steam, steam generator blowdown, and feedwater piping inside containment is located at higher elevations, and the postulated dynamic effects are separated from safe shutdown systems and components by distance and configuration. Table 3.6-1 provides a list of plant fluid systems that contain high and moderate-energy piping in the Nuclear Annex and Reactor Building.

Table 3.6-2 provides a list f the systems that are required for safe shutdown or to support safe shutdown. High- and moderate-energy pipe failure locations are postulated as described in Section 3.6.2.

Each postulated rupture location is evaluated for its effect on safe shutdown systems and components required following the specific pipe failure event.

3.6.1.1.1 IIigh-Energy Piping Systems A high-energy pipe failure is postulated in branches or piping runs larger than one inch nominal diameter and which operate during normal plant conditions with high energy fluid.

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Included in this category are fluid systems or portions of fluid systems which are pressurized during nonnal plant conditions or are maintained pressurized under conditions where either or both of the following are met:

• Maximum operating temperature exceeds 200*F, or

• Maximum operating pressure exceeds 275 psig.

Fluid piping systems that qualify as high-energy for only short portions of their operational period are considered moderate-energy systems if the portion of their operational period within the pressure and/or temperature specified above for high energy fluid systems is less than two percent of the time period required to accomplish its system design function.

In analyzing the effects of a high-energy pipe failure, the consequences of pipe whip, water spray, jet impingement, flooding, compartment pressurization, and environmental conditions are considered.

See Appendix 3.9A, Section 1.8.1.1 for a further discussion.

3.6.1.1.2 Moderate-Energy Piping Systems A moderate-energy pipe failure is postulated in branches or piping runs larger than one inch nominal diameter and which operate during normal plant conditions with moderate-energy fluid.

/^% Included in this category are fluid systems or portions of fluid systems which are pressurized above V atmospheric pressure during normal plant conditions or are maintained pressurized under conditions where both of the following are met:  ;

• Maximum operating temperature is 200*F or less, and

• Maximum operating pressure is 275 psig or less.

In arulyzing the effects of a moderate-energy pipe failure, the consequences of water spray, jet impingement, flooding, compartment pressurization, and environmental conditions are considered.

See Appendix 3.9A, Section 1.8.1.2 for a further discussion.

4 3.6.1.2 Description A listing of the high-energy lines inside the containment is given in Table 3.6-3. A listing of high-energy lines outside the containment is given in Table 3.6-4. Since the Turbine and Radwaste Buildings contain no safety-related equipment, high-energy line breaks in those buildings are generally excluded from this table.

Essential systems are those systems that are needed to safely shut down the reactor or mitigate the consequences of a pipe break for a given postulated piping failure. However, depending upon the type and location of a postulated pipe break, certain safety equipment may not be classified as essential for that particular event.

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System 80 + Design ControlDocument The essential systems which are to be protected from the eff ects of postulated piping failures are identified below. These essential systems were selected for each postulated break to satisfy the protection criteria given in the introduction to Section 3.6.

• The following systems, or portions of these systems, are required to mitigate the consequences of postulated breaks of high-energy reactor coolant pressure boundary piping that result in a loss-of-coolant-accident (LOCA) assuming a loss of offsite power.

1. Reactor Protective System.

2. Engineered Safety Features Actuation System.

3. Safety Injection System.

4. Containment Spray System.

5. Class IE Electrical Systems, AC and DC (including switchgear, batteries, and distribution systems), IE cabling and sensing lines.

6. Diesel Generator Systems, including Diesel Generator Starting, Lubrication, and Combustion Air Intake and Exhaust Systems.

7. Diesel Fuel Oil Storage and Transfer System.

8. liydrogen Recombiner System.

9. Control Building IIVAC System.

10. Component Cooling Water System (portions required for operation of other listed systems).

11. Ultimate lleat Sink (site specific).

12. Fuel Building liVAC System.

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13. Diesel Generator Room IIVAC System.

14. Main Control Board (See Tables 7.3-2 and 7.3-14 for systems required).

15. Containment Isolation Systems:

• Penetration assemblies e isolation valves l I

• Equipment hatch

• Emergency personnel hatch

• Personnel lock 1

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System 80+ ' oeskn controlDocument e Steel containment vessel-e Test connections j e Piping between penetration assemblies and isolation valves.

16. Ex-core Neutron Monitoring System.

17. Safety-related Radiation Monitors (refer to Section 11.5).

18. Shutdown Cooling System.  ;

19. Essential Chilled Water System.

20. Safety Depressurization System.

21. Emergency Feedwater System.

. 22. Air Coolers.

23. Station Service Water System.

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24. Reactor Coolant System.

i e The following systems, or portions of these systems are required to mitigate the consequences of postulated breaks in high-energy secondary pressure boundary piping (main steam, main i

. feedwater, blowdown, or emergency feedwater) assuming a loss of offsite power.

1. Reactor Protective System. .f

2. Engineered Safety Features Actuation System.

3. Safety injection System.

4. Containment Spray System (for breaks inside the containment only).

5. Main Steam and Feedwater System (from unaffected steam generator out to the containment isolation valves, including the atmospheric steam dump, steam supply to the turbine-driven emergency feedwater pump, and the steam generator blowdown line).

6. Shutdown Cooling System.

7. Class IE Electrical Systems, AC and DC (including Switchgear, Batteries and Distribution Systems), IE cabling and sensing lines.

8. Diesel Generator System, including Diesel Generator Starting, Lubrication and Combustion Air intake and Exhaust Systems.

V 9. Diesel Fuel Oil Storage and Transfer System.

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System 80+ Design ControlDocument

10. Component Cooling Water System (portions required for operation of other listed systems).

11. Ultimate IIcat Sink (site specific).

12. Control Building IIVAC System.

13. Fuel Building IIVAC System.

14. Main Control Board (See Tables 7.3-2 and 7.3-14 for systems required).

15. Essential Chilled Water System.

16. Containment Isolation Systems:

• Penetration assemblies

• Isolation valves

• Equipment hatch

• Emergency personnel hatch

• Personnel lock

• Steel containment vessel e Test connections

• Piping between penetration assemblies and isolation valves.

17. Diesel Generator Room IIVAC System.

18. Emergency Feedwater System.

19. Reactor Coolant System.

20. Ex-core Neutron Monitoring.

21. Station Service Water System.

22. Air Coolers.

• For other postulated breaks not included in the above two categories, systems must not be affected such that any break, evaluated on a case-by-case basis, violates the following criteria:

1. The pipe break must not cause a reactor coolant, steam, or feedwater line break.

2. The function of safety systems required to perform protective actions to mitigate the consequences of the postulated break must be maintained.

A14voved Design Materia!- Design of Ssc Page 3.6-6

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. - System 80 + Deslan ControlDocument

_ 3. The ability to place the plant in a safe shutdown condition must be maintamed A systematic integrated review of safety-related and associated systems is conducted during the design

' process to verify compliance with design criteria, interface requirements, and safety design bases.

i The potential effects of flooding as a consequence of a pipe break, or leakage or through-wall cracks (as defined in Section 3.6.2.1.2) are evaluated to ensure that the operability of safety-related equipment are

. not impaired.

5 An analysis of the potential effects of missiles is discussed in Section 3.5.

i The potential environmental effects of steam on essential systems are discussed in Section'3.11. In c

general, becmte of the protective measures of redundancy and separation between systems and trains, l the conser o . s effect of the transport of steam will not be sufficient to impair the ability of the essential .

system to t, u down the plant and/or mitigate the coasequences of the given accident of interest.

L l There are no high-energy lines in the vicinity of the control room. As such, there are no effects upon j 'the habitability of the control room by pipe break either from pipe whip, jet impingement, or transport

of steam.~ Further discussion on control room habitability systems is provided in Section 6.4. -

j 3.6.1.3 Safety Evaluation l By means of design features such as separation, barriers, and pipe whip and jet impingement restraints, all of which are discussed below, the effects of pipe break will not damage essential systems to an extent g -

that would impair their design function not affect necessary component operability. .

The ability of specific safety-related systems to withstand a single active failure concurrent with a

~

postulated event is discussed in the failure modes and effects analyses provided in Sections 5.4.7, 6.2, i 6.3, 6.5, 7.2, 7.3, 8.3, 9.2 and 10.4. ,

n /

e $cparation The plant arrangement provides separation to the extent practical between redundant safety j i '

systems in order to prevent loss of safety function as a result of hazards different from those for which the system is required to function, as well as for the specific event for which the system is required to be functional. Separation between redundant safety systems with their related auxiliary supporting features is the basic protective measure.

In general, layout of the facility followed a multistep process to ensure adequate separation.

l ~ 1. Safety-related systems are located away from most high-energy piping.

2. ~ Redundant (e.g., Division 1 and 2) safety systems and subsystems are located in separate compartments.

3. As necessary, specific components are enclosed to maintain the redundancy required for those systems that must functica as a consequence of specific piping failure events. ,

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System 80 + Design controlDocument

• Barriers-Shields and Enclosures Protection requirements are met through the protection afforded by the walls, floors e:M columns in many cases. Where adequate protection does not already exist due to separation, cdditional barriers, deflectors, or shields are provided as necessary to meet the functional protection requirements. Where compacments, barriers, and structures are required to provide the necessary protection, they are designed to withstand the combined effects of the postulated failure plus normal operating loads plus canhquake loadings.

• Piping Restraint Protection Where adequate protection does not already exist due to separation, barriers, shields, or piping restraints are provided as necessary to meet the functional protection requirements. Restraints are not provided when it can be shown that the pipe break would not cause unacceptable damage to essential systems or components.

The design criteria for pipe whip restraints are given in Section 3.6.2.3.2.5.

• Facility Response Analyses An evaluation of postulated pipe break events is performed to identify those safety-related systems and components that provide protective actions required to mitigate, to acceptable limits, the consequences of the postulated pipe break event.

Whenever the separation inherent in the plant design is shown to assure the functional capability of the safety systems required following a postulated pipe break event, no additional protective measures are required for that event, and additional considerations of break type, location, orientation, restraints, and other protective measures are not require /.. When necessary, additional protective measures are incorporated into the design to assure the functional capability of safety systems required following the postulated pipe break event.

In conducting the facility response analyses, the following criteria are utilized to establish the l integrity of systems and components necessary for safe reactor shutdown and maintenance of the shutdown condition:

l

1. Offsite power is assumed to be unavailable if an automatic turbine generator trip or automatic reactor trip is a direct consequence of a postulated piping failure.

2. In addition to the postulated pipe failure and its accompanying effects, a sinF le active l l

component failure is assumed in the systems required to mitigate the consequences of the postulated piping failure. l The single active component failure is assumed, except as noted in item 4 below.

3. Each high- or moderate-energy fluid system pipe failure is considered separately as a single postulated initiating event occurring during normal plant conditions.

4. Where a postulated piping failure is assumed in one of two redundant trains of a system that is required to operate during normal plant conditions as well as to shut down the  !

reactor, single failures that prevent the functioning of the other train or trains of that Approved Design Afaterial Design of SSC Page 3.6-8

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r a-System 80+ Deslan contmloccanart i= (

system are not assumed, provided the system is designed to Seismic Category I standards,

• is powered from offsite and onsite sources, and is designed, constructed, operated, and i

i inspected to quality assurance, testing, and inservice inspection standards appropriate for nuclear safety class systems.

J J *

5. All available systems and components, including non-Seismic Category I and those '

actuated by operator actions, may be employed to mitigate the consequences of a

. postulated piping failure. In judging the availability of such systems and components, account is taken of the postulated failure and its direct consequences, such as unit trip and .; '

loss of offsite power, and of the assumed single active component failure and its direct consequences. The feasibility of carrying.out operator actions is based on a minimum of 30 minutes delay responding to alarm indication and adequate access to equipment being available for the proposed actions. (Access to the containment post-LOCA is not assumed.)

I r

6. Piping systems containing high-energy fluids are designed so that the effects of a single postulated pipe break cannot, in turn, cause failures of other pipes or components with ,

unacceptable consequences. i l 7. For a postulated pipe failure, the escape of steam, water, and heat from structures I enclosing the high-energy fluid containing piping does not preclude: l

• Accessibility to surrounding areas important to the safe control of reactor

A operations. l

] l i

• Habitability of the control room. <

• Ability of instrumentation, electric power supplies, and components and controls to initiate, actuate, and complete a safety action. (A loss of redundancy is j permissible, but not the loss of function.)

The design criteria define acceptable types of isolation for safety-related elements and for high-energy

' lines from similar elements of the redundant train. Separation is accomplished by:

1

• Routing the two groups through separate compartments, or

• Physically separating the two groups by a specified minimum distance, or

• Separating the two groups by structural barriers.

The design criteria assure that a postulated failure of a high-energy line or a safety-related element cannot take more than one safety-related train out of service. The failure of a component or subsystem of one train may cause failure of another portion of the same train; for example, a Division 2 high-energy pipe may cause failure of a Division 2 electrical tray, but not failure of any Division 1 component. The

capability to shut the plant down safely under such a failure will therefore remain intact.

' Given the separation criteria above, and the pipe break criteria in Section 3.6.2.1.2, the effects of high-

, O.. energy pipe breaks are not analyzed where it is determined that all essential systems, components, and V ' structures are sufficiently physically remote from a postulated break in that piping tun.

w ooe> nenamw Deem er ssc rose 2.s s a

System 80 + Design ControlDocument 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping l Described herein are the design bases for locating breaks and cracks in piping inside and outside containment, the procedure used to define the thrust at the break location, the jet impingement loading  ;

criteria, and the dynamic response models.

I Site-specific information will include final designs of high- and moderate-energy fluid systems. The final designs and results of high- and moderate-energy piping analyses will be documented in a pipe break analysis report. The Pipe Break Analysis Report will provide the results of the pipe break analyses.

These analyses will be based on criteria used to postulate cracks and breaks in high- and moderate-energy piping systems as defined in Section 3.6.2 and will employ the analytical methods described in Section 3.6.2 and Appendix 3.6A.

For postulated pipe breaks, the Pipe Break Analysis Report will confirm that:

1. piping stresses in the containment penetration area are within their allowable stress limits,

2. pipe whip restraints and jet shield designs are capable of mitigating pipe break loads, and

3. loads on safety-related systems, structures and components are within their design load limits.

The Pipe Break Analysis Report will also confirm that structures, systems and components required for safe shutdown can withstand the environmental effects of postulated cracks and breaks.

An inspection of the as-built high-energy piping systems will be performed. The inspection of the as-built high energy pipe break features will be performed to verify:

• The location of pipe break mitigation devices (restraints, jet shields) e Clearances / gaps between restraints and piping

• The location of nearby safety-related targets to be protected from high-energy line breaks.

l Any differences between the as-built information and the as-designed information will be reconciled and l documented in a pipe break analysis report.

3.6.2.1 Criteria Used to Define Break and Crack Locations and Configurations l

l 3.6.2.1.1 General Requirements Postulated pipe ruptures are considered in all plant piping systems and the associated potential for damage to tequired systems and components is evaluated on the basis of the energy in the system. System piping is classified as high-energy or moderate-energy, and postulated ruptures are classified as circumferential breaks, longitudinal breaks, leakage cracks, or through-wall cracks. Each postulated rupture is considered separately as a single postulated initiating event.

I For eacli postulated circumferential and longitudinal break, an evaluation is made of the effects of pipe whip, jet impingement, compartment pressurization, environmental conditions, and flooding. For piping systems where leak-before-break is approved (Sections 3.6.2.1.3 and 3.6.3), dynamic effects of pipe Approved Desiger Meterial. Design of SSC Page 3.610

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System 80+ Destas conenat Docunmut .  :

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j ' breaks are not considered. If required to demonstrate safe plant shutdown, an internal fluid system load {

evaluation is performed on the effects of fluid forces on components within or bounding the fluid system.

2-For each postulated leakage crack, an evaluation is made of the effects of compartment pressurization, ,

environmental conditions and flooding. For each postulated through-wall crack, an evaluation is made of the effects of environmental conditions and flooding. The effects of pipe ruptures and/or leakage ,

cracks are included in the environmental qualification of safety-related electrical and mechanical  ;

equipment. Environmental qualification of safety related equipment is discussed in Section 3.11. The i evaluation of the required systems and components demonstrate that the protection requirements of l Section 3.6.1 are met.  ;

1rrespective of the fact that the criteria in Section 3.6.2 may not require specific breaks, if a structure ,

outside containment separates a high-energy line from an essential component, that separating structure is designed to withstand the consequences of the pipe break in the high-energy line that produces the .  :

greatest effect on the structure. Structures inside containment which are used to separate high-energy i lines from essential components are designed to withstand the dynamic load effects of postulated pipe l breaks not eliminated by leak-before-break. In addition, these structures inside containment are  :

adequately designed to withstand the greatest effect from (1) pipe breaks not eliminated by leak-before- j break, (2) the largest through-wall leakage crack in the 1 igh-energy line (minimum 10 gpm) whether or j j

not consideration of dynamic effects is eliminated by LBB for that line. or (3) the largest leak from i another leak source, such as a valve or pump seal.

3.6.2.1.2 Postulated Ruptum Descdptions e Circumferential Break

. )

A circumferential break is assumed to result in pipe severance with full separation of the two severed pipe ends unless the extent of separation is limited by consideration of physical means.

The break plane area (A,) is assumed perpendicular to the longitudinal axis of the pipe, and is ,

' assumed to be the cross-sectional flow area of the p'pe at the break location. The break flow area j (Ar) from each of the broken pipe segments for i circumferential break, with full separation of {

the two broken pipe segments, is equal to the break plane area (A,). The break flow area, j discharge coefficient and discharge correlatio'a are substantiated analytically or experimentally. {

y

?

i e Longitudinal Break 4 i a

A longitudinal break is assumed to result in a split of the pipe wall along the pipe longitudinal j axis, but without severance. The break plane area (A,) is assumed parallel to the longitudinal l axis of the pipe and equal to the cross-sectional flow area of the pipe at the break location. The  ;

j-break flow area ( A,) is equal to the break plane area (A,). The break is assumed to be circular  !

in shape or elliptical (2D x D/2) with its long axis parallel to the axis. The discharge coefficient  !

and any other values used for the area or shape associated with a longitudinal break.are  :

substantiated analytically or experimentally.

e Leakage Crack A leakage crack is assumed to be a crack through the pipe wall where the size of the crack and  !

corresponding flow rate are determined by analysis and a leak detection system, as described in l

, .,fJ .Section 3.6.3.

l I

6 Amsmuesf Des 4pn AIsewdsf.Desden of SSC -Pope 3.5-ff l

System 80+ Design ControlDocument

• Through-Wall Crack A through-wall crack is assumed to be a circular orifice through the pipe wall of cross-sectional flow area equal to the product of one-half the pipe inside diameter and one-half the pipe wall thickness.

3.6.2.1.3 Piping Approved for Leak-Before-Break A leak-before-break evaluation is performed for the reactor coolant system (RCS) main loop piping, surge line, shutdown cooling and safety injection lines and for the main steam line inside containment which eliminates the dynamic effects of pipe break from the design basis. The evaluation meets the requirements of 10 CFR 50, Appendix A, General Design Criterion (GDC) 4. The evaluation is performed using the guidelines of NUREG-1061, Vol. 3 (Reference 1) as described in Section 3.6.3.

3.6.2.1.4 Piping Other than Piping Approved for Leak-Before-Break This section applies to all high- and moderate-energy piping other than that whose dynamic effects due to pipe breaks are eliminated from the design basis by leak-before-break evaluation, as identified in Section 3.6.2.1.3.

3.6.2.1.4.1 Postulated Rupture Locations

• Class 1 Piping f(Ruptures', as specified in below, [larepostulated to occur at thefollowing locations in each piping network designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section 111}} (Reference 2) f(for Class 1 piping:

1. The terminal ends of the pressurizedportions of the run.

2. At intermediate locations selected by either one of thefollowing methods:

• At each location ofpotential high stress andfatigue such as pipefirtings (elbows, tees, reducers, etc.), valves, flanges, and welded attachments, or
• At each location where either of thefollowing conditions is exceeded.

1) %here the marimum stress range' between any two load sets (including the zero load set) calculated by Eq. (10) in Paragraph NB-3653, ASME Code, Section Ill, exceeds 2.4 S,,, and the stress range calculated by either Eq. (12) or Eq. (13) in Paragraph NB-3653 exceeds 2.4 S,,,.

2) Where the cumulative usagefactor (U) exceeds 0.1.}}'

'        For those loads and conditions in which Level A and Level B stress limits have been specified in the design Specification (excluding earthquake loads).

2 NRC Staff approval is required prior to implementing change in this infonnation; see DCD Introduction Section 3.5. Approved Design Material- Design of SSC Pope 3.6-12

I System 80+ oenlan contrat Document  ; Where, as defined in Subarticle NB-3650: 1 S,, = allowable stress-intensity value. j U = the cumulative usage factor. As a result of piping reanalysis due to differences between the design configuration and the as-  ! built configuration, the highest stress or cumalative usage factor locations may be shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exists:

                                                                                                                                     ?

(i) The dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints and jet shields.  ; (ii) A change is required in pipe parameters such as major differences in pipe size, wall l i thickness, and routing. Through-wall and leakage crack locations for Class 1 piping are specified below.  ! i l e Class 2, Class 3, or Seismically Analyzed ANSI B31.1 Piping

                                                                                                                                     +

f(Ruptures}}2, as specttied below, ((are postulated to occur at the following locations in each piping network designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section 111}}2, (Reference 2) llfor Class 2 and Class 3 piping, or with the rules of the

    '              ASME Codefor Pressure Piping, B31, Power Piping, ANSI /ASME B31.1}} (Reference 3) \\for                           l

- seismically analyzed ANSI B31.1 piping l

1. the terminal ends of the pressurizedponion of the network, and i

2. either

                                                                                                                                      )

e intermediate locations of potential high stress orfatigue such as pipe pttings, valves, flanges and welded-on attachments, or e where thepiping contains nopttings, weld attachments, or valves, at one location at each extreme of the piping run adjacent to the protective structure, or I e intermediate locations where the stress, S, exceeds 0.8(X + Y).}}

j. where, as defined in Subarticle NC-3650:

E S = stresses under the combination ofloadings for which either Level A or Level B service limits have been specified, as calculated ! from the sum of equations (9) and (10). 1 2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction Section 3.5.  ! d

     \/    8        For those loads and conditions in which Ixvel A and level B stress limits have been specified in the I'

Design Specification (excluding canhquake loads). L 2Dee@n anenentet.Dee6n er ssc rene 2.s.12 l J

Systern 80 + Design ControlDocument X = equation (9) Service Level B allowable stress. Y = equation (10) allowable stress. As a result of piping reanalys's due to differences between the design configuration and the as-built configuration, the highest stress locations may be shifted; however, the initially determined intermediate break locations may be used unless a redesign of the piping resulting in a change in pipe parameters (diameter, wall thickness, routing) is required, or the dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints and jet shields. Through-wall and leakage crack locations for Class 2 and Class 3 piping are specified below.

• Non-Safety Related ANSI B31.1 Piping System 80+ piping is designed so as to isolate seismically analyzed piping from non-seismically analyzed piping. In cases where it is not possible or practical to isolate the seismic piping, adjacent non-seismic piping is analyzed according to Seismic Category II criteria. For non-seismic piping attached to seismic piping, the dynamic effects of the non-seismic piping are simulated in the modeling of the seismic piping. The attached non-seismic piping up to the analyzed /unanalyzed boundary is designed not to cause a failure of the seismic piping during a j

seismic event. For non-safety class piping which is not seismically analyzed, through-wall cracks are postulated at axial locations such that they produce the most severe environmental effects.

   *     \(Break Locations in Piping Runs with Multiple ASME Code Piping Classes Breaks}}' as specifled below,((arepostulated to occur at thefollowing locations:

1. The terminal ends of the pressurizedportions of the run.

2. At intermediate locations selected by either one of thefollowing methods:

1 e At each location ofpotential high stress orfatigue, such aspipefittings, valves, flanges, and welded anachments; or

• At all intermediate locations between terminal ends where the stress andfatigue limits of Section 3.6.2.1.4.1for Class 1,2,3 or seismically analyzed ANSI B31.1 piping are exceeded.))'
• Break Locations Both circumferential and longitudinal breaks are postulated to occur, but not concurrently. in all high-energy piping systems at the locations specified above, except as follows:

2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction Section 3.5. Approved Deskn hteterkal. Design of GSC Page 3.6-14

        .. , . .                                     --         -_- . _ = - - -                       .      .  - -             . - . - .
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System 80 + " Design ControlDocument

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1. Circumferential breaks are not postulated in piping runs of a nominal diameter equal to f or less than 1 inch. i i 2. Longitudinal breaks are not postulated in piping runs of a nominal diameter less than 4  :

inches.- i 3.' Longitudinal breaks are not postulated at terminal ends. ~ 4 Only one type of break is postulated at locations where, from a detailed stress analysis, such as finite-element analysis, the state of stress can be used to identify the most . 1: probable type. If the primary plus secondary stress in the axial direction is found to be  ! at least 1.5 times that in the circumferential direction for the most severe loading l combination association with Level A and Level B service limits, then only a  : circumferential break is postulated. Conversely, if the primary plus secondary stress in the circumferential direction is found to be at least 1.5 times that in the axial direction for the most severe loading combination associated with Level A and Level B service  : limits, then only a longitudinal break is postulated. l 4

5. Circumferential and longitudinal breaks are not postulated at locations where the l requirements for piping near containment isolation valves, below, are satisfied. i l

6. Circumferential and longitudinal breaks are not postulated at locations where the criterion for leakage cracks, item 2 below, is used.

l

• Crack Locations

1. Through-Wall Cracks

~ lilbrough-wall cracks are postulated in all hightnergy and moderate-energy piping systems having a nominal diameter greater than 1 inch, except that through-wall cracks j are not postulated at locations where

}

        ~

e For Class 1 piping, the calculated' value of S is less than one-half the stress or i usage limits; e For Class 2, Class 3 or seismically analyzed ANSI B31.1 piping, the calculated' mlues of S is less than one-half the stress limits;}}* \ l

• The requirements for piping near containment isolation valves, below, are.

satisfied. l

                               -*         Tne criterion in Item 2 below is used.

l i

                 '      For those loads and conditions in which level A and I.evel B stress lunits have been spccified in the -

p  : Design Specification (excluding carthquake loads). 2I NRC Staff approval is required prior to implementing change in this information; see DCD Introduction .

                      . Section 3.5.

Anwewd Deem annanter Dee6n erssc page 2.s.rs l 1

System 80+ Design ControlDocument For moderate-energy fluid systems in areas other than containment penetration, through. wall cracks are postulated at axial and circumferential locations that result in the most severe environmental consequences. Where a break in a high-energy fluid system is postulated which results in more limiting environmental conditions, the through-wall crack in the moderate-energy fluid system is not postulated. For moderate-energy fluid systems, through-wall cracks are also not postulated in those portions of piping from the containment wall to and including the inboard or outboard isolation valves provided that they meet the requirements of the ASME Code, Section III, NE-1120 and the stresses calculaed by the sum of Eq. (9) and (10) of the ASME Code, Section III, NC-3653 do not exceed 0.4 times the sum of the stress limits given in NC-3653. Through-wall cracks, instead of breaks, are postulated in the piping of fluid systems that qualify as high-energy fluid systems for short operational periods of time but that qualify as moderate-energy fluid systems for the major operational period. Where a postulated through-wall crack in a moderate-energy fluid system piping results in more limiting environmental conditions than the break in proximate high-energy fluid system piping, through-wall crack location criteria, as stated above, are used to determine crack locations.

2. Leakage Cracks A leakage crack is postulated in place of a circumferential break, or longitudinal break, or through-wall crack, ifjustified by an analysis performed on the pipeline in accordance with the requirements of Section 3.6.3.

• Piping Near Containment Isolation Valves O

Breaks and cracks are not postulated between the containment wall and the inboard or outboard isolation valves in piping which is designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code, Section III (Reference 2) and which meets the following additional requirements:

1. The fr n ving design stress and fatigue limits are not exceeded:

For ASME Code. Section III. Class 1 Pinine

• The maximum stress range between any two loads sets (including the zero load 3

set) does not exceed 2.4 S., and is calculated by Eq. (10) in NB-3653, ASME Code, Section Ill. If the calculated maximum stress range of Eq. (10) exceeds 2.4 S., the stress ranges calculated by both Eq. (12) and Eq. (13) in Paragraph NB-3653 meet the limit of 2.4 S,. 8 For those loads and conditions in which Level A and 1.cvel B stress limits have been specified in the Design Specifications (excluding canhquake loads). Approved Destyn Meterial. Design of SSC Pope 16-16

i 1 System 80+^ Desinn controlDocument - y 7 (

• The cumulative usage factor is less than 0.1.
• The maximum stress, as calculated by Eq. (9) in NB-3652 under the loadings j 1

resulting from a postulated piping failure beyond these ponions of piping does . ] ~ not exceed the lesser of 2.25 S, and 1.8 Sy except that following a failure outside ' containment, the pipe between the outboard isolation valve and the first restraint i may be permitted higher stresses provided a plastic hinge is not formed and l

<                                      operability of the valves with such stresses is assured in accordance with the                                ,

i requirements specified in SRP Section 3.9.3. Primary loads include those which

are deflection limited by whip restraints.  !

i For ASME Code. Section III. Class 2 Pioina j I

                              *      . The maximum stress as calculated by the sum of Eqs. (9) and (10) in Paragraph                                 ;
                                      'NC-3653, ASME Code, Section'ill, considering those loads and conditions                                       ;

thereof for which level A and level B stra limits have been specified in the j system's Design Specification (i.e., sustwed loads, occasional loads, and 1 thermal expansion) excluding earthquake loads does not exceed 0.8 (1.8 Sn + { Sx). The S, i and S are allowable stresses at maximum (hot) temperature and  ! A allowable stress range for thermal expansion, respectively, as defined in Anicle l NC-3600 of the ASME Code, Section III. l i- I

• The maximum stress, as calculated by Eq. (9) in NC-3653 under the loadings  !

resulting from a postulated piping failure of fluid system piping beyond these portions of piping does not exceed the lesser of 2.25 S3 and 1.8 Sy . l Primary loads include those which are deflection limited by whip restraints. The . exceptions permitted in (c) above may also be applied provided that when the  !

piping between the outboard isolation valve and the restraint is constructed

- inaccordance with the Power Piping Code ANSI B31.1 (see ASB 3-1 B.2.c.(4)), i the piping shall either be of seamless construction with full radiography of all  ! circumferential welds, or all longitudinal and circumferential welds shall be fully [ radiographed. i j 2. Welded attachments, for pipe supports or other purposes, to these portions of piping is  ! avoided except where detailed stress analyses, or tests, are performed to demonstrate l compliance with the limits of Item 1 above. , 3. The number of circumferential and longitudinal piping welds and branch connections are  ! minimized. Where guard pipes are used, the enclosed portion of fluid system piping is seamless construction and without circumferential welds unless specific access provisions i l are made to permit. inservice volumetric examination of the longitudinal and circumferential welds.

4. The length of these portions of piping is reduced to the minimum length practical.  ;

4 0 ll Ammedouewmuoner onoworssc rap 2.s.11  ; l _ . . _ - . _ . _ . _ . - .. _- _ _ . _ _ . ._ l

System 80+ Design ControlDocument The design of pipe anchors or restraints (e.g., connections to containment penetrations j 5. and pipe whip restraints) does not require welding directly to the outer surface of the l piping (e.g., flued integrally forged pipe fittings may be used) except where such welds are 100 percent volumetrically examinable in service and a detailed stress analysis is l performed to demonstrate compliance with the limits of Item 1 above.

6. Guard pipes provided for those portions of piping in the containment penetration areas are constructed in accordance with the rules of Class MC, Subsection NE of the ASME  :

Code, Section III, where the guard pipe is part of the containment boundary. In addition, the entire guard pipe assembly is designed to meet the following requirements ) and tests:

• The design pressure and temperature is not less than the maximum operating pressure and temperature of the enclosed pipe under normal plant conditions.
• The Level C stress limits in NE-3220, ASME Code, Section III, is not exceeded under the loadings associated with containment design pressure and temperature in combination with the safe shutdown earthquake.
• Guard pipe assemblies are subjected to a single pressure test at a pressure not less than its design pressure.
• Guard pipe assemblies do not prevent the access required to conduct the inservice examination specified in Item 7 below. Inspection ports, if used, are not located in that portion of the guard pipe through the annulus of dual barrier containment structures.

7. A 100% volumetric inservice examination of all pipe welds are conducted during each inspcction interval as defined in IWA-2400, ASME Code, Section XI.

8. Following a postulated pipe break of high-energy piping beyond either isolation valve, the stresses in the piping from the containment wall, to and including the length of the isolation valve, are maintained within Level C Service Limits as specified in the ASME Boiler and Pressure Vessel Code, Section III, (Reference 2).

9. The design and in-service inspection requirements, as specified in MEB 3-1 (Reference 4), are satisfied. Inservice inspection program requirements are given in Sections 5.2.4 and 6.6.

10. The containment isolation valves are appropriately qualified to assure that operability and leak tightness are maintained when subjected to any combination of loadings, which may be transmitted to the valves from postulated pipe breaks beyond the valves.

11. For moderate-energy piping, the stresses calculated by the sum of equations (9) and (10) in ASME Code, Section III, NC-3653, do not exceed 0.4 times the sum of the stress limits given in NC-3653.

O Approved Des &n Materiel Design of SSC Page 16-18

System 80 + - Deslan ControlDocument 4 3.6.2.1.4.2 Postulated Rupture Configurations e Break Configurations i Where break locations are postulated without the benefit of a stress calculation, breaks are i assumed to occur at the piping welds to each fitting, valve, or welded attachment. If detailed l stress analyses or tests are performed, break locations are selected as specified in Section -l 3.6.2.1.4.1. , Circumferential breaks are postulated in fluid system piping and branch runs as specified in .; Section 3.6.2.1.4.1. Instrument lines, one inch and less nominal pipe of tubing size are designed l to meet the provisions of Regulatory Guide 1.11. Longitudinal breaks in fluid system piping and branch runs are postulated as specified in Section  ; 3.6.2.1.4.1.

e Crack Configurations Through-wall cracks are postulated at those axial locations specified in Section 3.6.2.1.4.1.

s For high-energy piping, through-wall cracks are postulated to be in those circumferential locations - that result in the most severe environmental consequences. The flow from the crack is assumed l to wet all unprotected components within the compartment with consequent flooding in the ' compartment and communicating compartments. Flooding effects are determined on the basis i of a conservatively estimated time period required to effect corrective actions. 3.6.2.1.5 Details of Containnant Penetrations Details of containment penetrations are discussed in Sections 3.8.1 and 3.8.2. 3.6.2.2 Analytical Methods to Define Forcing Functions and Response Models 4 3.6.2.2.1 Piping Approved for Leak-Before-Break 4 There are no forcing functions or response models for the reactor coolant loop, surge line, shutdown cooling line, safety injection line and main steam line based upon elimination of dynamic effects by leak-before-break evaluation. 1

3.6.2.2.2 Analytical Methods to Define Forcing Functions and Response Models for Piping Excluding that Approved for Leak-Before-Break This section applies to all high-energy piping other than that whose dynamic effects due to pipe breaks are eliminated from the design basis by leak-before-break evaluation.

AnproweEW neenwin! Doefpn of SSC page 3.s.19 __ .. m ._ - . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1__ _ ___o

System 80 + Design ControlDocument 3.6~.2.2.2.1 Determination of Pipe Thrust and Jet Loads

• Circumferential Breaks

([Circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections, unless physically limited by piping restraints, structural members, orpiping snffness.}}2

• Dynamic Force of the Fluid Jet Discharge llThe dynamic force of the fluid Jet discharge from either a postulated circumferential or longitudinal break is based on a circular break area equal to the cross-sectionalflow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically determined thrust coefficient, as determinedfor a circumferential break at the same location.}}2 Line restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs are taken into account, as applicable, in the reduction of jet discharge.

Piping moven.ent is assumed to occur in the direction of the jet reaction, unless limited by structural memters, piping restraints, or piping stiffness.

• Pipe Blowdown Farce and Wave Force

           \\Thefluid thrustforces that resultfrom eitherpostulated circumferential or longitudinal br.n'.s, are calculated using a simphfied one stepforcingfunction methodology.}}2 This methodology is based on the simplified methods described in References 5 and 6.
           \\When the simph) led method}}2 discussed above f(leads to impracticalprotective measures, then a more detailed computer solution which more accurately reflects the postulatedpipe rupture event is used.))2 The computer solution is based on the NRC's computer program developed for calculating two-phase blowdown torces (Reference 7).

• Evaluation of Jet Impingement Effects Jet impingement force calculations are performed only if structures or components are located near postulated high energy line breaks and it cannot be demonstrated that failure of the structure or component will not adversely affect safe shutdown capability.
• Longitudinal Breaks A longitudinal break results in an axial split withou. vverance. The split is assumed to be orientated at any point about the circumference of the pipe, or alternatively at the point of highest stress as justified by detailed stress analyses. For the purpose of design, ((the longitudinal break 2

NRC Staff approval is required prior to implementing change in this information; see DCD Introduction Section 3.5. Approved Design Material Design of SSC Page 3.6-20

System 80+' oesian contrar Document

      ~'                   is assumedto be circular or elliptical (2D x 1/2D) in shape, with an area equal to the largest piping cross-sectionalflow area at the point of the break and have a discharge coeficient of                             ,

1.0.))2 Any other values used for the area, shape and discharge coefficient associated with a longitudinal break is verified by test data which defines the limiting break geometry. i 3.6.2.2.2.2' . Methods for the Dynamic Analysis of Pipe Whip I Pipe whip restraints usually provide clearance for thermal expansion during normal operation. If a break O occurs, the restraints or anchors nearest the break are designed to prevent unlimited movement'at the l - point of break (pipe whip). The dynamic nature of the piping thrust load is considered. ((In the absence of analyticaljustufication, a dynamic loadfactor of 2.0 is applied in determining restraint loading.}}2 i

Elastic-plastic pipe and whip restraint material properties may be considered as applicable. The effect 3 of rapid strain rate of material properties is considered in accordance with Reference 5. A 10 percent '

increase in yield strength is used to account for strain rate effects.

               ~ In general, the loading that may result from a break in piping is determined using either a dynamic i

'  : blowdown or a conservative static blowdown analysis. ((7he methodfor analyzing the interaction efects

               ' of a whipping pipe with a restraintis one of thefollowing: (1) the Energy Balance Method (2) Lamped                              .

Parameter Method, or (3) Equimlent Static Method.}}2 The energy balance method is based on the principle of conservation of energy. The kinetic energy of l the pipe generated during the first quarter cycle of movement is assumed to be converted into equivalent I < strain energy, which is distributed to the pipe or the whip restraint. See Appendix 3.6A for a discussion of the application of the energy balance method.

           )     The lumped parameter method is carried out by utilizing a lumped mass model. Lumped mass points are l

interconnected by springs to take into account inertia and stiffness properties of the system. A dynamic forcing function or equivalent static leads may be applied at each postulated break location with pipe whip ' interactions. A nonlinear elastic-plastic analysis of the piping-restraint system is used. The computer meth'od for this analysis is described in Appendix 3.6A. A conservative static analysis model is used for rigid rupture restraints. In order to obtain the design load

               ' for a rigid restraint, the following equation is used:

4 F = 2 x 1.1 x Fs = 2.2F3 where: I F - = the design load Fa = maximum blowdown force  ; and the dynamic load factor (DLF) is taken as 2.0 and rebound effects are accounted for by a factor of 1.1. t . (~ ( 2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction Section 3.5. ,

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System 80 + Design ControlDocument 1 3.6.2.2.2.3 Method of Dynamic Analysis of Unrestricted Pipes llThe impact velocity and kinetic energy of unrestrictedpipes is calculated on the basis of the assumption that the segments at each side of the break act as rigid-plastic cantilever beams subject to piecenise constant blowdown forces. The hinge location is fixed either at the nearest restraint or at a point determined by the requirement that the shear at an interiorplastic hinge is zero.}}2 The kinetic energy of an accelerating cantilever segment is equal to the difference between the work done by the blowdown force and that done on the plastic hinge. The impact velocity Vi is found from the expression for the kinetic energy: 2 KE = (1/2) M,,V 3 where M,, is the mass of the single degree of freedom dynamic model of the cantilever. The impacting mass is assumed equal to M,,. For a straight run of pipe rotating about a plastic hinge, the zone of influence of the whipping pipe accounts for an increasing length due to a traveling hinge point caused by strain hardening effects. The impact energy of unrestrained pipe into a barrier (e.g. the divisional wall) is governed by the vector component of its velocity at impact which is perpendicular to the barrier. Impact of small piping into building structures conservatively assumes that all of the impact energy is imparted to the barrier with no dissipation due to local crushing deformation of the pipe. Bearing area of impact on building structure is generally elliptical, but is treated as a circle of equivalent area, with dimensions based on experimental data for pipe crush behavior. As the impact load is greatest on the periphery of the ellipse, this yields a conservative force distribution into the barrier. Long term loading on the barrier subsequent to impact due to system blowdown and continued deceleration of remaining pipe (beyond the impact zone) is accounted for in addition to the initial impulsive loading. 3.6.2.3 Dynamic Analysis Methods to Verify Integrity and Operability 3.6.2.3.1 Pipe Whip Restraints and Jet Deflectors for Piping Approved for Leak-Before-Break There are no pipe whip restraints and jet deflector for the reactor coolant loop, surge line, shutdown cooling line, safety injection line and main steam line based upon elimination of dynamic effects due to pipe breaks by leak-before-break evaluation. 3.6.2.3.2 Pipe Whip Restraints and Jet Deflectors for Piping Other than that Approved for Leak-Before-Break This section applies to pipe whip restraints for all piping other than that whose dynamic effects due to pipe breaks are climinated from the design basis by leak-before-break evaluation. Supplementary information on the design and selection of pipe whip restraints is given in Appendix 3.6A. 2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction Section 3.5. Approved Design Materiel- Design of SSC Page 3.6-22

9 System 80+ Desinn controlDocanent 1 3.6.2.3.2.1 General Description of Pipe Whip Restraints  ; ( When required, pipe . whip restraints are provided to protect the plant against the effects of whipping 4 during postulated pipe break. The design of pipe whip restraints is governed not only by the pipe break blowdown thrust, but also by functional requirements, deformation limitations, properties of whipping  ;

         . pipe and the capacity of the support structure. The restraint is designed for the impact force induced by the maximum possible initial gap between the whip restraint and the process pipe..                                           ;

The impact energy is usually too high for an elastic restraint system or support structure to absorb. Therefore, energy absorbing restraints are designed utilizing the energy balance approach (impact energy

            + external work = internal energy of pipe restraint system), are provided. Energy absorbing pipe whip l           restraints consist oflaminated straps or a crush pipe or crush pad as described in Appendix 3.6A, Section 1.1.

3.6.2.3.2.2 Pipe Whip Restraint Components ,

j. Pipe whip restraints typically consist of the following components: )

e Energy Absorbing Members Members that are under the influence of impacting pipes (pipe whip) absorb energy by significant  ; plastic deformations (e.g., laminated straps, crush pipes and crushable honeycomb material). e Non-Energy Absorbing Members i ! Those components which form a direct link between the pipe and the structure (e.g., crush pipe , shim plates and components other than energy absorbing members). e Structural Attachments ! Those fasteners which provide the method of attaching connecting members to the structure (e.g., welds and bolts). o Building Structure o Steel and concrete support structures which ultimately carry the restraint load. Design criteria are specified in Sections 3.8.3 and 3.8.4. 3.6.2.3.2.3 Design Loads i Restraint design loads, the reactions, and the corresponding deflections are established using the criteria delineated in Section 3.6.2.2.2. .. I 3,6.2.3.2.4 Allowable Stasses .i l

The allowable stresses are as follows: l Allowable stresses used in the design of the pipe break restraint components are consistent with the component function. The upper design limit for pipe break restraint energy absorbing members under pipe rupture dynamic loading is 50 percent of the restraint material ultimate strain based on test of actual L =:onew nonnaw.onow orssc reen 2. m 1

i

System 80+ Design ControlDocument material used for the pipe break restraint. For steady state loading following a pipe rupture, the same strain limit applies up to either 80% of the actual material ultimate strength or 80% of the Code specified minimum ultimate strength. The allowable stresses associated with the non-energy absorbing members and structural attachments are given in Section 3.9.3.1.4. For the steel and concrete building structures, allowable stresses are specified in Sections 3.8.3 and 3.8.4. 3.6.2.3.2.5 Design Criteria The unique features in the design of pipe whip restraint components relative to the structural steel design are geared to the loads used and the allowable stresses. These are as follows:

• Energy-absorbing members are designed for the restraint reaction and the corresponding deflection established according to the pipe size and material and the blowdown force using the criteria delineated in Section 3.6.2.2.
• Non-energy-absorbing members, structural components, and their attachments to the building structure are designed to remain elastic.

All essential components are evaluated for jet impingement and pipe whip effects using a dynamic or an equivalent static analysis of testing to demonstrate either the functional capability and/or operability in addition to the structural integrity of the component. 3.6.2.3.2.6 Materials The materials used are as follows:

• For energy-absorbing members: Carbon steel such as A-106 Grade B or equivalent for crush pipe, crushable honeycomb made of stainless steel for compression, and stainless steel such as Type 304 for laminated strap restraints.
• For other components: ASTM A-588, ASTM A-572 Grade 50, and ASTM A-36.

3.6.2.3.2.7 Jet Impingement Shields Protection from jets is provided by using separation and redundancy (as described in Section 3.6.1), guard pipes, and where necessary, jet shields. 3.6.2.4 Guard Pipe Assembly Design Criteria Guard pipes to limit pressurization effects in the containment penetration area will not be used except in

" Hot Penetration" assemblies as described in Section 3.8.2.1.3.4.

3.6.2.5 Compartment Pressurization and Temperature Analysis Outside Contaimnent Energy releases into corcoartments following a postulated pipe rupture in high and moderate energy lines outside containment can et eate pressure differentials across structural walls and slabs and result in adverse environmental conditions for electrical and mechanical equipment located in the compartments. Whereas compartment pressurization analysis is performed to determine pressure loadings on building structures, environmental pressure and temperature analysis is performed to define conditions for equipment qualification. The same basic analytical methods and computer codes are used in both cases, with Approved Design Material- Design of SSC Pope 3.6-24

i

                                                                                                                                         ]

i

      ~ System 80 +                                                                                      Desinn controlDocument           l l

changes in assumptions and model where appropriate to assure conservative results. n Lo' g-term mass and i

energy releases are used to determine environmental conditions for design and evaluation of equipment and building structures.  ; The greatest loads on building structures occur shortly after the pipe rupture. The structures are i designed to maintain their integrity if such loads were to be imposed on them. The following paragraphs describe the break postulation criteria and calculational techniques used for compartment pressurization and environmental analysis outside containment. 4 3.6.2.5.1 Break Postulation Criteria Break Postulation Criteria' for high energy piping is presented in Section 3.6.2.1. For compartment analysis a minimum of one break in each compartment is postulated, and breaks are postulated so as to maximize the adverse effects from pressurization and temperature. When necessary to assure worst l conditions, the accident (e.g., the Main Steam System pipe break) is analyzed for a spectrum of pipe .

break sizes and various plant power conditions. ,

3.6.2.5.2 Dete:=i==*lon of Mass and Energy Release Rates i Piping system energy release transients for the postulated pipe rupture are determined by either a hand  ; calculation or by computer analysis. The plant operating mode which results in the greatest energy i release rate is used. For hand calculation the break mass flow rate is obtained frcm a critical flow correlation which predicts an upper bound flow rate for the rupture geometry and fluid state under . consideration. Examples are the Moody correlation (two-phase and saturated steam conditions), the Homogeneous Equilibrium Model (single phase steam), and the Henry Fauske correlation (subcooled  ; liquid). Blowdown flow rate is obtained from the following equation per ANSI /ANS-56.10: l W = CoAGc where: I W = mass flow rate  !

l. i Co = discharge coefficient 2 A = break area l j Gc = critical mass flux

                                                                                                                                        ?

The break fluid enthalpy is set equal to the stagnation enthalpy of the fluid in the ruptured pipe. A flow I discharge coefficient of 1.0 is used unless a lower value is justified as required by ANSI /ANS-56.10. . For complex systems and where less conservative release rates are needed, computer analysis is . employed. Initial conditions (e.g., fluid pressure, fluid temperature) are chosen within normal operating limits such that the set which will result in the largest release rates are used. A system model of

;        appropriate complexity is generated and computer programs of the RELAP4 type are used. To calculate                           -l the pipe break response, the fluid system is divided into discrete volumes (control volumes or nodes)                           J which are connected to other volumes by a junction. The equations of conservation of mass and energy                            j are solved in the nodes, and the one-dimensional momentum equation is solved in the flow paths. A time                           l
      ' history of system conditions is output by - the code.                   CEFLASH 4A (Section 3.9.1.2.1.21),                        ,

RELAP4/ MOD 5, and RELAPS/ MOD 3 (Reference 15) are computer codes applicable to the generation [

       . of mass and energy releases. - Also, SGNIII (Section 6.2.1.4.4) may be used in the case of main steam '
       'line breaks.                                                                                                                     ,

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System 80+ oesign controlDocument j 1 3.6.2.5.3 Compartment Pressurization Analysis and Environmental Pressure and Temperature Analysis Compartment pressurization analysis is performed to determine pressure loadings on building structures. Envirorunental pressure and temperature response analysis defines pressure and temperature conditions for qualification of mechanical and electrical equipment. Computer codes are generally used in some phase of this analysis. Typically the model includes a network of volumes and junctions. Volumes represent rooms, corridors, pipe chases, and other ponions of buildings outside Containment. When appropriate, volumes also are used to simulate the HVAC system and outside atmosphere. Junctions represent flow paths between the volumes. Multinode analysis may be required within a companment. The computer codes addressed below provide acceptable results for both compartment pressurization and environmental pressure and temperature analyses, with appropriate assumptions and models changed to obtain conservative results. The DDIFF-1 computer code (Reference 19) is used to predict subcompartment conditions following incident initiation during which the maximum pressure differentials on structures or components would occur. The transient calculations include determination of mass flow rates, mass and energy inventories, absolute and differential pressures, and temperatures in the subcompartment system. The subcompartment system is a control volume-flow path spatial network created based upon the geometry of the plant regions being analyzed. RELAP4/ MOD 5, RELAF5/ MOD 3, and COMPARE may be used for these analyses. Another computer code which may be applied here is the multicompartment containment system analysis code CONTEMPT 4/ MOD 4 (Reference 17). It is used to predict the long term thermal-hydraulic behavior of a series of standard compartments. The code calculates the time variation of compartment thermodynamic properties, temperature distributions in heat conducting structures, mass and energy inventories in compartments, and mass and energy transfer due to intercompartment junction flow by solving the mass and energy balance equations. The GOTHIC computer code (Reference 18) is a state-of-the-art program for modeling multiphase flow. It solves the conservation equations for mass, momentum and energy for multicomponent, two-phase flow. The code contains a flexible noding scheme that allows lumped parameter, one , two, or three-dimensional analysis or any combination of these to be conducted. Conservation equatio s are solved for three fields: (1) steam-gas mixture (2) continuous liquid,and (3) liquid droplet. It calculates the relative velocities between these fields, including the effect: of two-phase slip on pressure drop and heat transfer between phases and between surfaces and the fluid. 3.6.3 Leak-Before-Break Evaluation Procedure This section desews Leak 4 fore-Break (LBB) analysis for all applicable piping. LBB analysis is used to eliminate fro J sne structuu.) design bases the dynamic effects of double-ended guillotine breaks and equivalent longitudinal breaks for an applicable piping system. LBB is demonstrated for the following System 80 + piping systems:

1. Main Coolant Loop (MCL) piping, hot and cold legs

2. Surge Line (SL)  ;

3. Direct Vessel Injection (DVI) Line (main run inside containment)

Approved Design Material- Design of SSC Page .T.6-26

l . ' System'80+' Deslan coneer Document '  ; <' 4. Shutdown Cooling Line (SC) (main run inside containment) i 5. Main Steam Line (MSL) (main run inside containment) l Supplemental information on LBB methodology and on design of piping systems to LBB criteria is given ' i

            ~ in Appendix 3.9A.

3.6.3.1 Applicebility of LBB 1 Piping systems for which LBB is demonstrated are first shown to meet the applicability requirements for [ NUREG-1061, Volume 3. Specifically, the points considered for applicability for LBB are:

1. Regulatory requirements - level of susceptibility of failure from erosion, erosion / corrosion, erosion / cavitation, waterhammer, creep fatigue, corrosion resistance, indirect causes, cleavage type failure, and fatigue cracking.

            ' 2.           Technical requirements - pipe properties, normal operation, seismic load levels, and stratified         !

flow, where applicable. 3.6.3.1.1' Design Basis Loads The LBB evaluations are based on design basis loads using the design configuration. Piping analyses of final detailed designs will confirm that LBB criteria is met for each piping system listed above. 3.6.3.1.2 Susceptibility of Failure from Erosion, Erosion / Corrosion, Erosion / Cavitation j

             ' Systems susceptible to erosion / corrosion pipe wall thinning are those with wet steam, flashing liquids, or liquid flow with high localized velocities. These factors are considered along with water chemistry and usage time to determine susceptibility and appropriate preventative methods.                                    !

t i 3.6.3.1.2.1 Erosion' Corrosion Minimization ' For systems susceptible to erosion / corrosion, the following methods are used to minimize degradation: ,i e Proper material selection is essential for the prevention of excessive pipe wall thinning. Carbon steel piping is not susceptible to erosion / corrosion under dry steam conditions. Low alloy steel is significantly more resistant to wall thinning than carbon steel in wet steam or under conditions  !

of two phase flow. Stainless steel is essentially immune to erosion / corrosion and is used in the most susceptible areas such as in wet steam or flow conditions where it is difficult to maintain ,

tight control of _ water chemistry. e Additional wall thickness is sometimes specified to accommodate a limited amount of wall l thinning without violating code requirements. 1 e- The bulk fluid velocity is limited to prevent excessive erosion of the pipe wall. The following j velocity guidelines are used for carbon steel piping: l

O  !

 .                                                                                                                                 r L ..: ' Dee> neenerief Deeg of SSC                                                                   Pope 3.6-27    ,

System 80 + Design control Document l Recommended Bulk Velocity Guidelines l Service Velocity I Steam Piping 150 ft/sec Water 15 ft/sec Recirculation Lines (Infrequent Use) 20 - 25 ft/sec Velocity guidelines may be increased on a case by case basis through the utilization of engineering evaluations which address the erosion / corrosion aspects and piping material selected in the design. The engineering evaluation will be performed utilizing industry accepted tools and methods such as EPRI Checkmate.

• Pipe routing is utilized to lower susceptibility to pipe wall thinning caused by adverse hydrodynamic conditions.

3 6.3.1.2.2 Applicability to Piping for LBB Use of high quality steels, stainless steel or stainless sted imed in the MCL, SL, DVI, and SC piping prevents erosion, erosion / corrosion, and erosion / cavitation. Additionally, water chemistry for the reactor coolant system is closely controlled and monitored. There is no evidence of unusual wall thinning in these pipes due to erosion, erosion / corrosion, or erosion / cavitation in pressurized water reactor plants. Therefore, these pipes havt s very low level of susceptibility of failure from these failure mechanisms. Carbon steel is used in MS', piping. There is no evidence of wall thinning due to erosion or erosion / corrosion for MSL p' ping inside containment, because dry steam and the operating temperature prevents erosion and crosioi / corrosion degradation. Therefore, MSL piping inside containment has a very low level of susceptibil ty of failure from these failure mechanisms. 3.6.3.1.3 Susceptibility of Failure from Water IIammer 3.6.3.1.3.1 Main Coolant Loop (MCL) and Surge Line (SL) There is a very low potential for water hammer in the sub-cooled water solid portions of the reactor coolant system since these portions of the reactor coolant system are designed to preclude void formation. Safety valve discharge loads associated with the pressurizer have been specifically identified and included in the component design basis. Therefore, the MCL and SL piping have a very low level of susceptibility of failure from water hammer. 3.6.3.1.3.2 Direct Vessel Injection (DVI) Line NUREG/CR-2781 (Reference 10) idet tified four water hammer events from the NUREG/CR-2059 (Reference 20) database involving the safety inicction system. EPRI research on water hammer events included these four events and included two more related to the Safety Injection System as reported in EPRI NP-6766 (Reference 21). Five of these six events occurred in piping upstream of the injection check valves due to steam pocket collapse (3 events), filling of a voided line (1 event), and unknown (1 event). The sixth event occurred in the low-head safety injection suction pump piping due to an unknown cause. AppromiDesign Material Design of SSC Page 3.6-28

~ p - System 80+ Design ControlDocument O The most likely root cause for most of these water hammer events is the leaking of the check valves, allowing hot water to enter a low pressure region and then flash into steam bubbles. The steam pocket

 !          thus formed would permit a steam pocket collapse type water hammer to occur if it were suddenly pressurized by the addition of water to the low pressure piping. The prevention of this type of water hammer is procedurally assured for the System 80+ design during initial system fill and plant operation.

Procedures for initial fill and venting ensure that voids will not occur initially in the System 80+ DVI piping. High point vents provide for the proper venting of lines and pumps. If this piping is then pressurized (above the calculated leakage induced temperatures / saturation pressures), that pressure coupled with the generally low temperature of the DVI system ensures that the lines would remain full , and that steam bubbles would not develop near the check valves. However, further protection against this type of water hammer is provided administratively by monitoring the pressure in the injection line and flushing upon high pressure. Pressure indication and alarms are provided to alert the operator of an increase in pressure to 1000 psig (from normal of about 620 psig). This is an indication of high temperature RCS leakage past the DVI check valve. Upon alarm, the operator opens the injection line drain valve (SI-618,628,638 or 648; see Figure 6.3.2-1C). This depressurizes the injection line to the SIT pressure while replenishing the volume with subcooled water at containment ambient temperature. This replenishment is performed slowly so as not to exceed the makeup capability,therefore minimizing the potential for collapse of any steam pockets that may have formed. Normal valve operation, pump startup and pump trip will create negligible fluid transient loads for the DVI system. The " sixth event" as reported in EPRI NP-6766, is not a concern to System 80+ because the Safety Injection System does not have a low-head Safety Injection Pump. All results of all design . n bases events are mitigated with the use of four high pressure pumps. U Based on system operating procedures which require venting of DVI lines, and the low number and low i severity of events reported for safety injection type systems in PWR's, the susceptibility of water hammer induced failures in the System 80+ DVI system is very low. Thus, the DVI system meets the screening criterion for water hammer. 3.6.3.1.3.3 Shutdown Cooling (SC) Line f NUREG/CR-2781 (Reference 10) identified only one water hammer event from the NUREG/CR-2059 (Reference 20) database involving the PWR residual heat removal (RHR) system. EPRI research on water hammer events included this event and included six more related to the RHR system as reported in EPRI NP-6766 (Reference 21). These events occurred in five areas including piping adjacent to the reactor coolant isolation valves (1 event), the high point piping of the RHR heat exchanger (1 event),  : RHR pump discharge piping (2 events), branch piping to the chemical and volume control system (2 events), and piping adjacent to the reactor coolant system cold leg isolation valves (1 event). These lines  ! are only susceptible to a small number of the generic causes of water hammer - rapid valve opening or cleaning and steam bubble collapse. l There is little potential for water hammer loading due to the first cause because there are no fast-acting  ! 1 valves in the System 80+ SC system, and it is very unlikely for a steam bubble to form in the line. Under normal power operation, the valves in the line are closed and the fluid in the line is at ambient temperature. Thus, the vapor pressure is low and steam bubble formation will not occur. During .i ' shutdown cooling operation, the system is open to the RCS and will have the same vapor pressure as the tO - RCS, which will be subcooled due to the hydrostatic head formed by the water and steam in the b pressurizer. Therefore, steam bubble formation is precluded by the characteristics inherent to the system.

w Dee6a aneumin!- Deelen et ssc pose 1s.2s

i l l Systern 80 + Design ControlDocurnent l Even though there have been water hammer events reported in parts of the PWR RHR system, severe water hammer events are not expected for the shutdown cooling line. Based on the low severity for the i types of events to which the SC system is subject, the SC system meets the screening criterion for water hammer. 3.6.3.1.3.4 Main Steam Line (MSL) NUREG/CR-2781 (Reference 10) identified six water hammer events from the'NUREG/CR-2059 (Reference 20) database involving the PWR main steam system. EPRI research on water hammer events included these six events and included two more related to the main steam system as reported in EPRI NP-6766 (Reference 21). Six of these eight events occurred in piping adjacent to main steam isolation valves due to valve operation (3 events), and steam propelled water slug (3 events), one unknown event in piping downstream of the turbine bypass valve, and one unknown event in the main steam reliefline. None of the events caused damage to MSL piping. The System 80+ main steam supply system (including MSL pipe support system components) is designed to accommodate steam hanuner dynamic loads and relief valve discharge loads resulting from rapid closure of system valves and safety / relief valve operation without compromising safety functions. The number of 90-degree elbows and miters is minimized in the MSL piping layout to recuce the effects of steam and water hammer. Valves in the main steam supply system are designed to withstand loads developed from the various operating and design basis events described in Section 3.9.3. Transients due to steam-propelled water slugs are prevented by proper procedures and operation of the MSL to avoid pockets and water slugs in the piping. Based on the low severity of the water hammer events described in NUREG/CR-2781 and the design considerations of the System 80+ main steam supply system, the MSL piping has a very low level of susceptibility of failure from water hammer. 3.6.3.1.4 Susceptibility of Failure from Creep Fatigue Creep fatigue is a concern for ferritic steel piping at operating temperatures above 700*F and for austenitic stainless steel piping at operating temperatures above 800*F. Operating temperatures of the System 80+ piping systems are below these limits, and therefore not susceptible to creep fatigue failure. 3.6.3.1.5 Susceptibility of Failure from Corrosion Materials used in the MCL, SL, DVI, SC and MSL piping are highly resistant to corrosion. Material selection, fabrication controls, and water chemistry ensure resistance to corrosion. The MSL carbon steel piping material is exposed to a dry steam environment. Under these service conditions carbon steel forms a protective oxide film which inhibits further corrosion. Therefore, the general corrosion rate of carbon steel in dry steam is very low. Carbon steel is also resistant to the effects of erosion-corrosion under dry steam service conditions. To prevent intergranular stress corrosion attack of the austenitic stainless steel surge line, fabrication and operation controls are implemented. Primary ' vater chemistry is controlled to minimize contaminants, and the dissolved oxygen is at a level that would normally preclude intergranular stress corrosion cracking (IGSCC). Approved Desip Material- Desim of SSC Page 3.6-30

i i i i l System 80+ Desian controlDocument l l Therefore, through material selection, chemistry control and fabrication control, these pipes have a very ] low level of susceptibility of failure from corrosion. ) , i I See Sections 5.2.3.2,5.2.3.3 and 5.2.3.4 for water chemistry controls and fabrication of reactor coolant l

boundary components.

3.6.3.1.6 U+" "'d of Failure from Indirect Causes Pipe degradation or failure from indirect causes such as fires, missiles and component support failure is  ! prevented by designing, fabricating and inspecting to criteria that ensures low probability of the event or its impact on safety related structures. As an example, the overhead polar crane is designed to Seismic . Category II to prevent it from becoming a missile and impacting these piping systems or other safety-related equipment. Therefore, the MCL, SL, DVI, SC and MSL piping have a very low level of susceptibility to failure from indirect causes. i ~3.6.3.1.7 Cleavage Type Failure 1 Cleavage type failures are generally not a concern for the system operating temperatures and materials , used for the MCL, SL, DVI, SC and MSL piping. In addition, material tests (ASME Section III Code  ! required toughness tests and J-R tests) show the materials for these pipelines to be highly ducile and highly resistant to cleavage type failures at operating temperature. ] I 3.6.3.1.8 Susceptibility of Failure from Fatigue Cracking j 3.6.3.1.8.1 Class 1 Piping The MCL, EL, DVI, and SC piping are designed to meet the ASME Section III subsection NB fatigue i I criteria. All design basis transients identified in Section 3.9.1 are included in the detail stress analyses. 1 Therefore, those pipes have a very low susceptibility of failme from fatigue cracking. l l 3.6.3.1.8.2 Class 2 Piping I Safety Class 2 piping is designed to meet ASME Section III Subsection NC fatigue criteria. All design ' basis transients identified in Section 3.9.1 are included in establishing the allowable stress limits in accordance with NC 3611.2. That is, the allowable stress range for expansion stress is reduced for cyclic conditions based'on the number of equivalent full temperature cycles. All design basis transients are included in determining the number of equivalent full temperature cycles and the stress range reduction factors for cyclic conditions. The main steam line piping is not subjected to severe Level A or B thermal or pressure transients other

                          . than heatup and cooldown. Thermal transient heatup and cooldown rates are on the order of normal                                        .

l heatup and cooldown and temperature variations for transients other than heatup and cooldown are generally less than 50*F. Level A and B pressure variations for transients other than heatup and  ;

cooldown are generally less than 20% of the design pressure.

The impact of gross bending on the fatigue life of the piping is conservatively considered in the Class 2 design. A comparison of allowable thermal stress range from Subsection NC of the code to the allowable , 6  : alternating stress from Subsection NB demonstrates that the Class 2 allowable is less than the Class 1  :

  ?                          allowable through the range of expected equivalent full temperature cycles- in the steam line.                                         l Additionally, there are no significant terps.ture or pressure variations which would result in significant

? :Dee> nenowner- Dee> et sac rege 2.s.21 l

             . _ - _ _              a                                 .        -,%i4      -m             ., ~       rs- ...-
                                                                                                                                 +-W r- -r---"     swiwssi- ,

System 80+ Design ControlDocument local or through wall stresses. If a detailed fatigue evaluation were to be performed, a very low usage factor would be expected. Therefore it is concluded that the steam line has a very low susceptibility of failure from fatigue cracking. 3.6.3.2 Leakage Crack Location A survey of the piping is performed to determine the locations of highest stress loading and coincident poorest material properties. All base metal, weld materials, heat affected zones in the vicinity of the terminal ends, and all intermediate elbow locations are considered. 3.6.3.3 Leak Detection There are two major aspects to leak rate based on crack detection in addition to the crack opening size; leak detection capability, and flow rate correlation for leakage through a crack. 3.6.3.3.1 Leak Detection System A leak detection system is recommended by Regulatory Guide 1.45, Reference 8, capable of det.cong a leakage rate of 1.0 gpm or less to the primary reactor containment. NUREG-1061, Volume 3, recommends a safety margin of ten on the leak detection system capability. Diverse measurement means are provided for System 80+ for leakage detection, including RCS inventory monitoring, sump level and flow monitoring, and measurement of airborne radioactive particulates and gases (see Section 5.2.5). The RCS primary water inventory balance method is used to detect leakage rates of 1 gpm or less. Leak detection system requirements to support the LBB analysis for main steam line piping are met by a combination of humidity detectors. air cooler condensate flow monitors, radioactive airborne activity sensors, sump flow and level meters, and the RCS inventory balance instrumentation. Total plant leakage inside contaimnent is continuously monitored via the containment floor drain (HVT) sump, reactor cavity sump, and containment cooler condensate tank instruments described in Section 5.2.5.1. If the total unidentified plant leakage inside containment exceeds 1.0 gpm, an RCS inventory balance is performed to quantify RCS unidentified leakage. Increases in containment cooler condensate flow are attributable to containment humidity increases due to high energy leaks inside containment, and/or containment cooler tube leakage. Therefore, subtracting RCS unidentified leakage and any known high energy leaks inside containment (e.g., RCS, main steam, and feedwater leaks) from the plant leak rate determined by containment cooler condensate tank monitoring provides a conservative estimate of main steam line leakage. 3.6.3.3.2 Flow Rate Correlation The other major aspect of crack detection based on the leak rate, namely the flow rate correlation for leakage through a given crack size, cannot be predicted precisely. Variables such as surface roughness of the side walls of the crack, the nonparallel telationship of the side walls due to the elongated crack shape, and possibly zigzag tearing of the material during crack formation all introduce uncertainties in defining an exact flow rate correlation. The leakage rate required to be detectable is 1.0 gpm or less. The licensing guidelines (NUREG-1061, Volume 3) recommend a factor of 10 on that leakage rate for conservatism unless otherwise justified. The LBB evaluations of System 80+ primary side piping systems listed in Section 3.6.3 are based on Approvmf Desijn Meterial- Designs of SSC Page 3.6-32

                 -      -         -.                - ~.=        . . . -       , - - -                                     ..            -. ._       -.

l i System 80+ Deska ControlDocument i a leak detection capability of 1.0 gpm, with a safety margin of 10. The LBB evaluation of the System 80+ main steam line inside containment is based on a leak detection capability of 1.0 gpm and a safety j margin of 10.

                                                                                                                                                             ]

See Appendix 3.9A for further discussion of flow rate correlation. 3.6.3.4 Material Properties 1 For the main coolant loop, the hot and cold leg piping material is SA516 Gr70 or SA508 CLI A. All hot- ! and cold-leg pipe-to-pipe welds and the pipe-to-reactor vessel, steam generator and reactor coolant pump welds are carbon steel. All main loop component nozzles are SA508 CL 1 A,2 or 3 or SAS41 CL 1, 2 ( ' or 3. The surge line is SA312 Type 347 or Type 316 stainless steel, resulting in bimetallic safe end welds. The shutdown cooling line and the direct vessel safety injection line are Type 304 or 316 stainless steel. The main steam line is SA516 Gr70. l  ! The stainless steel piping fabricated for the surge, shutdown cooling and direct vessel injection lines are j seamless pipes. The detailed analysis of cracks in pipe welds requires consideration of the properties of the pipe and the weld materials. Previous work by C-E has shown that a conservative bounding analysis results when the material stress-strain properties of the base metal (lower yield) and the fracture properties of the weld (lower toughness) are used for the entire structure, (Reference 11). This material 4 representation is applicable to all LBB analyses discussed in Section 1.9 of Appendix 3.9A. For both  ; the final design and as-built configurations, material properties for piping systems subject to LBB which are listed in Section 3.6.3 will be reviewed. If either the base metal or the weld is found to have lower

  /~'s             fracture toughness properties than those given in Appendix 3.9A, a LBB reanalysis using the material V                with the lower fracture toughness properties as the basis for the J R curve will be performed. The tensile (stress-strain) curves and the Jo vs. Aa curves are required for each material type. Additional commitments with respect to review of final design and as-built configurations for piping systems subject                                 :

to LBB are given in Section 1.9 of Appendix 3.9A.  ! I Leakage Crack Length Detennination 3.6.3.5 J It is necessary that hypothesized through-wall cracks open significantly to allow detection by normal leakage monitoring under normal full power loadings, i The method for determining the appropriate leakage crack length is described in Section 1.9.6.2 of Appendix 3.9A. i 1 3.6.3.6 Computation of J Integral Values 3.6.3.6.1 Range of Crack Sizes The range of crack lengths are calculated using a detailed stability analysis of the through-wall cracks in the piping evaluated. The finite-element analysis is performed for the leakage crack size and twice that length. This procedure, therefore, considers the stability of a range of crack lengths for all locations selected for the analysis. l 3.6.3.6.2 ' J-Integral I i

   \

The stability of through-wall cracks is evaluated using the J-integral technique. The J-integral is j determined in the finite-element analysis for pressure, normal operation, and maximum design load, i Anorend Dee6" aseenow- Du@n er ssc rene 2.s-23 i

SyOtem 80+ Design Control Document which is the largest of the dynamic loads (due to safe shutdown earthquake, thermal stratified flow, rapid valve closure, or other load) included in the crack stability analysis. The J-integral is determined for two different crack lengths for each geometric model. For the margin on loads evaluation, the J-integral for the leakage crack size is evaluated for V2 x (Pressure + NOP + Maximum Design) loads. For the margin on crack length evaluation, the J-integral for 2 times the leakage crack size is evaluated for (Pressure + NOP + Maximum Design) loads. 3.6.3.7 Stability Evaluation The stcbility of the crac ked pipes is assessed by comparing the J-integral value due to the applied loads on the pipe to the material crack resistance. The stability criterion for ductile crack extension employed is: if J-applied < J ci material, and (dJ/ja) ,ppw < (dJ/da) w then crack stability is assured. The change in J-integral with crack length "a" is determined by analyzing several crack lengths in the region of interest. For a leakage crack of length "a", crack lengths "a", a-6, and a+6 are analyzed. Similarly, the change in J-integral with crack length in the region of length "2a" is determined by analyzing cracks with lengths 2a,2a-6, and 2a+6. This method provides the derivative information in the two regions of interest. The variation of J with crack length in the region of "a" and "2a" is plotted along with the material curve. Evaluation of the plots allows for direct verification of the stability criteria. The evaluations are performed for the locations chosen to envelop all limiting cases. The pipes with the leakage crack length subject to loads of V2 x (P + NOP + Maximum Design Load) and the pipes with crack length twice the leakage crack length with loads of (P + NOP + Maximum Design Load) are demonstrated to have significant margin between the material cutve and the loading curve, indicating that all pipe locations satisfy the LBB crack stability criteria. An acceptable alternative method for the margin on loads and margin on crack length evaluations is to combine each component of the NOP load and the Maximum Design Load absolutely. This method is referred to as "the absolute sumtnation of loads method." If this alternative method is used, the margin on load for the leakage crack size is reduced from V2 to 1. The margin on crack length (2 times the leakage crack size) remains the same. See Appendix 3.9A, Sections 1.1.9.5.4 and 1.1.9.6 for a discussion of LBB design criteria development and a further discussion of analytical methods. 3.6.3.8 Results The piping listed in Section 3.6.3 and evaluated by the methods described above are shown to meet all the ((criteriafor application ofleak-before-break))2 according to NUREG 1061, Volume 3. Specifically, these criteria lirequire thefollowing:}}2 2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction 0 Section 3.5. Approwd Design niaterial Design of SSC 111/96) Page 3.6-34

System 80+ Design ControlDocument . I O e - Cracks which are assumed to grow through the pipe wall leak significantly while remaining stable. The amount of \\ leakage is detectable with a safety margin of at least afactor of10}}2 unless otherwise justified. e flCracks of the length that leak at the rate given above can withstand normal operation plus maximum design load loads with a safetyfactor ofat least 42. Alternatively, cracks of the length that leak at the rate given above can withstand the absolute combination of normal operation load - components and maximum design load with a factor of1. e - Cracks twice as long as those addressed above will remain stable when subjected to normal operation plus maximum design load.}}2 Site specific evaluations will confirm that the bases for the LBB acceptance criteria are satisfied by the final as-built design and materials of the piping systems listed in Section 3.6.3 and will be documented in a LBB evaluation report. References for Section 3.6

1. " Evaluation of Potential for Pipe Breaks," NUREG-1061, Vol. 3.

2. ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Class 4

1, 2 or 3.

3. ASME Code for Pressure Piping, B31, Power Piping, ANSI /ASME B31.1.

  .. [}G

4. USNRC Branch Technical Position MEB 3-1 Rev. 2 - Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, attached to Standard Review Plan 3.6.2, June,1987.

5. American National Standard Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture, ANSI /ANS 58.2-1988.

6. R. T. Lahey, Jr. and F. J. Moody, " Pipe Thrust and Jet Loads," The Thermal Hydraulics of a Boiling Water Nuclear Reactor, Section 9.2.3, pp. 375-409, Published by American Nuclear i Society, Prepared by the Division of Technical Information, United States Energy Research and l' Development Administration,1977.

7. RELAP 4/ MOD 5. Computer Program User's Manual 098. 026-5.5.

8. USNRC Regulatory' Guide 1.45 " Reactor Coolant Pressure Boundary Leakage Detection Systems."  ;

9. Not used l

         -10.        NUREG/CR-2781, " Evaluation of Water Hammer Events in Light Water Reactor Plants," July 1982.

a(3 2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction

                   . Section 3.5.

4ewoner oeew neesenter ooep or ssc 11stssi reee 2.s 2s i

                                                            ~ . - - _ _        - , _ ___                              _   ,

System 80+ Design ControlDocument

11. " Analysis of Cracked Pipe Weldments," EPRI NP-5057, February 1987.

12. USNRC Regulatory Guide 1.11 (Safety Guide 11), Instrument Lines Penetrating Primary Reactor Containment: including supplement, Backfitting Considerations.

13. Not used

14. ANSI /ANS-56.10-1982, "Subcomparunent Pressure and Temperature Transient Analysis in Light Water Reactors," 1982.

15. "RELAPS/ MOD 3, Code Manual, Volume I, Code Structure, System Models, and Solution Methods (draft), EG&G Idaho, Inc., June 1990

16. "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis,"

CENPD-133P, August,1974 (Proprietary).

17. CONTEMPT 4/ MOD 4, "A Multicompartment Containment Systems Analysis Program,"

NUREG/CR-3716, EG&G Idaho, Inc.

18. " GOTHIC Containment Analysis Package User's Manual," Version 3.4, Numerical Applications, Inc., April 1991.

19. CENPD-141, Revision 2, "DDIFF-1 Code, A Description of the DDIFF-1 Digital Computer Code for Reactor Plant Subcompartment Analysis," Combustion Engineering,1nc. March,1978.

20. NUREG/CR-2059, R. L. Chapman et al, " Compilation of Data Concerning Known and Suspected Water liammer Events in Nuclear Power Plants, CY 1969-May 1981," May 1982.

21. EPRI NP-6766, D. A. Van Duyne et al, " Water Hammer Prevention, Mitigation, and Accumulation, Vohune 1: Plant Water Hammer Experience," Final Report, July 1992.

O Approved Desigra Material- Desbgra of SSC Page 3.6-36

System 80+ Design controlDocument l l n Table 3.6-1 High- and Moderate-Energy Fluid Systemst31 (v) Illgh-Energy Huld SystemsI3I Moderate-Energy Huld Systems Main Steam Chemical and Volume Control Main Feedwater Component Cooling Steam Generator Blowdown Safety Injectiont2r Auxiliary Steam Containment Air Monitoring / Sampling Reactor Coolant including Surge Line Diesel Generator Engine Fuel Oil Chemical and Volume Control Station Service Water Fire Protection Compressed Gas Pool Cooling and Purification Station Heating Contamment Sprayt2) Drain Essential / Normal Chilled Water Breathing Air Instrument Air Service Air Shutdown Coolingt21 Emergency Feedwatert2j Diesel Generator Engine Lube Oil Diesel Generator Engine Starting Air Safety Depressurization!'l [. L)

   \                                                                        Steam Generator Wet Layup Recirculationt21 Notes:

131 Systems shown are either totally or partially high-energy.

12) See Sections 3.6.1.1.1 and 3.6.1.1.2 for definitions of High and Moderate Energy Lines. These piping systems are considered Moderate Energy, since these systems operate above the temperature and/or pressure limits for only a relatively short portion (less than approximately two percent) of the time during which they perform their intended function, as prescribed by guidance of ANSI /ANS-58.2-1988. Portions of these systems are, however, interfaced directly with other liigh Energy systems or have sections which are High Energy in nature.

til Systems or port ons of systems which are not located in the Reactor Building Subsphere, Nuclear Annex, or inside Contamment were generally excluded from this table by the guidance of Section 3.6.1.2. I'l That ponion of SDS up to normally closed isolation valves; downstream portion is normally unpressurized. Or l I U Approved Design Meterial Design of SSC Page 3.6 37

System 80+ oesign controlDocument Table 3.6-2 Systems Required for Safe Shutdown and/or to Mitigate the Consequences of a Design-Basis Accident I!I Reactor Protective Reactor Coolant Safety Injection Shutdown Cooling Containment Spray Emergency Feedwater Component Cooling Water Station Service Water Auxiliary Power Area Radiation Containment Isolation Battery a"d DC Distribution Diesel Generator Diesel Fuel Oil Engineered Safety Features Hydrogen Recombiner Instrument and Control Power In< ore Thermocouple Ex-core Neutron Monitoring Main Steam Isolation Valves and Steam Dump Valves  ; Control Element Assembly Drive Control Room HVAC j Diesel Generator Room HVAC 1 ESF Switchgear Room HVAC Station Service Water Pump House HVAC Primary Containment Ventilation Safety injection System Equipment Room HVAC Essential Chilled Water l Ultimate Heat Sink (site specific) Main Control Board Safety Depressurization Remote Shutdown Panel O Ill Systems listed are either totally or partially required for safe shutdown. Approved Destyn Mornial Design of SSC Page 3.6-38

1 1 System 80+ omian controlDocumart l

Table 3.6-3 High-Energy Lines Within Coafninment .

t Operating '! i Line Operating Item Functional Pressure Temperature l No. System Description -(>275 psig) (> 200*F) Figure . Yes Yes 10.1 2,. j

 !                    !  Main' Steam            From SG No.1 Line #1 to Cont Penetration                                                     10.3.2-1                       l

i. ,

2 Main Steam From SG No.1 Line #2. Yes Yes 10.1-2, i to Cont Penetration 10.3.2-1 3 Main Steam From SG No. 2 Line #1 Yes Yes 10.1-2,  ; 1 1 to Cont Penetration 10.3.2-1 i e 4 Main Steam From SG No. 2 Line #2 Yes Yes 10.1-2,  ! to Cont Penetration 10.3.2-1 l l 5 SG SG No.1 Blowdown Yes Yes 10.1-2, , 1 Blowdown Line #1 to Common 10.4.8-1 l Blowdown Line r 6 SG SG No.1 Blowdown Yes Yes 10.1-2, , Blowdown Line F2 to Common 10.4.8-1  ; Blowdown Line 7 SG SG No. I Blowdown Yes Yes 10.4.8-1 Blowdown Common Line to Cont Penetration 8 SG SG No. 2 Blowdown Yes Yes 10.1-2, Blowdown Line #1 to Common 10.4.8-1 ) Blowdown Line  ! 9 SG SG No. 2 Blowdown Yes Yes 10.1-2, l l Blowdown Line #2 to Common 10.4.8-1 i Blowdown Line

                  -10    SG                     SG No. 2 Blowdown                Yes         Yes                         10.1-2, Blowdown               Common Line to Cont                                                      10.4.8-1 Penetration i                    11   SG Generator           SG No. I Secondary               Yes         Yes                         10.1-2 S'de Drain Line #1 to arst isolation Valve i                    12   SG Generator           SG No. I Secondary .             Yes         Yes                         10.1-2 Side Drain Line #2 to -

First ! solation Valve 1 SG Generator SG No. 2 Secondary Yes . Yes 10.1-2 Side Drain Line #1 to First Isolation Valve - 3O' 14 SG Generator - SG No. 2 Secondary Yes Yes 10.1-2' h Side Drain Line #2 to First Isolation Valve l z . :_' Doeten neuemet- Doeten er ssc . rose 2.s.2s w

System 80+ oesign controlDocument Table 3.6-3 High-Energy Lines Within Containment (Cont'd.) Line Operating Operating Item Functional Pressure Temperature No. System Description (>275 psig) (> 200*O Figure 15 S/G Wet Layup SG No.1 Wet Layup Yes Yes 10.1-2 Recirc Return Recire Nozzle to Interior Check Valve 16 S/G Wet layup SG No. 2 Wet Layup Yes Yes 10.1-2 Recite Return Recirc Nozzle to Interior Check Valve 17 Main Feedwater To SG No.1 Yes Yes 10.1-2, Economizer Nozzle #1 10.4.7-1 from Economizer Common Feedwater Line Junction 18 Main Feedwater To SG No.1 Yes Yes 10.1-2, Economizer Nozzle #2 10.4.7-1 from Economizer Common Feedwater Line Junction 19 Main Feedwater From SG No.1 Yes Yes 10.1-2, Economizer Common 10.4.7-1 Main Feedwater Line Junction to Cont. Penetration 20 Main Feedwater From SG No.1 Yes Yes 10.1-2, Downcomer Nozzle to 10.4.7-1 Cont. Penetration 21 Main Feedwater To SG No. 2 Yes Yes 10.1-2, Economizer Nozzle #1 10.4.7-1 from Economizer Common Feedwater Line Junction 22 Main Feedwater To SG No. 2 Yes Yes 10.1-2, Economizer Nozzle #2 10.4.7-1 from Economizer Common Feedwater Line Junction 23 Main Feedwater To SG No. 2 Yes Yes 10.1-2, Economizer Common 10.4.7-1 Main Feedwater Lin. Junction to Cont. Penetration Appronni Design MaterM. Design of SSC Page 3.640 l l 1

_ -.= ._. ._ _ System 80+ oesign controloccument r Table 3.6-3 High-Energy Lines Within Containment (Cont'd.) Lbe Operating Operating Item Functional Pressure Temperature No. System Description (>275 psig) (> 200*F) Figure 24 Main Feedwater From SG No. 2 Yes Yes 10.1-2, Downcomer Nozzle to 10.4.7 1 Cont. Penetration 25 Reactor Coolant SG No.1 RCS Hot Yes Yes 5.1.2-1 leg Loop 26 Reactor Coolant SG No.1 RCS Pump Yes Yes 5.1.2 1 1A Discharge Line 27 Reactor Coolant SG No.1 RCS Pump Yes Yes 5.1.2-1 IB Discharge Line 28 Reactor Coolant SG No.1 RCS Pump Yes Yes 5.1.2-1 1A Suction Line 29 Reactor Coolant SG No.1 RCS Pump Yes Yes 5.1.2-1 IB Suction Line 30 Reactor Coolant SG No. I RCS loop Yes Yes 5.1.2-1 p I A Drain Line to ( RDT (High-Energy to isolation Valve RC-334) 31 Reactor Coolant SG No.1 RCS loop Yes Yes 5.1.2-1 i IB Drain Line to l RDT (High-Energy to l 1 solation Valve  ! 1 RC-335) 32 Reactor Coolant SG No. 2 RCS Hot Yes Yes 5.1.2-1 Leg loop 33 Reactor Coolant SG No. 2 RCS Pump Yes Yes 5.1.2-1 2A Discharge Line 34 Reactor Coolant SG No. 2 RCS Pump Yes Yes 5.1.2-1 2B Discharge Line 35 Reactor Coolant SG No. 2 RCS Pump Yes Yes 5.1.2-1  ; 2A Suction Line i l 36 Reactor Coolant SG No. 2 RCS Pump Yes Yes 5.1.2-1 2B Suction Line 37 Reactor Coolant SG No. 2 RCS Imop Yes Yes 5.1.2-1 ) (~g 2A Drain Line to RDT (High-Energy to

 - (j isolation Valve RC-333)                ,                                                l
             % .a: centra neenew. Design or ssc                                                          rege zu t f

System 80+ Design Contro/ Document Table 3.6-3 High-Energy Lines Within Contalmnent (Cont'd.) Line Operating Operating Item Functional Pressure Temperature No. System Description (>275 psig) (> 200'F) Figure 38 Reactor Coolant SG No. 2 RCS 1. cop Yes Yes 5.1.2-1 2B Drain Line to RDT (High-Energy to Isolation Valve RC-332) 39 Reactor Coolant SG No. 2 RCS Hot Yes Yes 5.1.2-1 Leg Drain Line to RDT (High-Energy to Isolation Valve RC-215) 40 Not Used N/A - -- N/A l 41 Not Used N/A - - N/A l 42 Not Used N/A - - N/A l 43 Not Used N/A - - N/A l 44 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-1 & from Cold Leg Loop 5.1.2-3 1 A to I A Spray Control Valve RC-1&OE 45 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-1 & from Cold leg 1. cop 5.1.2-3 lb to IB Spray Control Valve RC-100F 46 Reacer Coolant Pressurizer Spray Line Yes Yes 5.1.2-3 from Imop 1 A Spray Control Valve RC-100E to Pressurizer Spray Common Header O Approved Design Material Design of SSC (11/96) Page 3.6-42

                                                                                      ._._._m..               .. _.    . . _

System 80+ oestan controlDocument 1 I , '\ Table 3.6-3 High-Energy Lines Within Containment (Cont'd.) l 1 Line Operating Operating j

                .Iten:                                 Functional        Pressure- Temperature                                 1 No. -         Systesa             Description       (>275 psig)  (> 200'S           Figure J

47 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-3  ! frorn loop IB Spray Control Valve j RC-100F to Pressurizer Spray 1 Conunon Header l 4 i 48 Reactor Coolant Pressurizer Spray .Yes Yes 5.1.2 -! Conunon Header to l Pressurizer i 49 Reactor Coolant Pressurizer Surge Line Yes Yes 5.1.2-1 & < 5.1.2-3 J , 50 Reactor Coolant Division 1 RCS Hot Yes Yes 5.1.2-1, -i 6.3.2-lC l i leg to Isolation Valve SI-651 (Shutdown  : - Cooling Line) 51 Reactor Coolant Division 2 RCS Hot Yes Yes 5.1.2-1, i leg to Isolation Valve 6.3.2-1C SI-652 (Shutdown Cooling Line) 52 Reactor Coolant Hot leg lajection Yes Yes 6.3.2-lC Loop #1 Check Valve SI-522 to Junction of Shutdown Cooling Line 1. cop #1 53 Reactor Coolant Hot leg Injection Yes Yes 6.3.2-lC

Loop #2 Check Valve SI-532 to Junction of Shutdown Cooling Line loop #2 54 Reactor Coolant Direct Vessel Injection Yes Yes 5.1.2-1 &

Connection #1 to SIS - 6.3.2-lC laterior Check Valve SI-247 55- Reactor Coolant Direct Vessel Injection Yes Yes 5.1.2-1 & Connection #2 to SIS 6.3.2-lC Interior Check Valve SI 227 n 56 . Reactor Coolant Direct Vessel Injection Yes Yes 5.1.21&

  -!                                          Connection #3 to SIS                                  6.3.2-lC

Interior Check Valve -

SI-237 - z ::Dee5r aseenrier. Deets or ssc - Pnne 3.6 43

System 80+ Design Control Document Table .'.6-3 Ifigh-Energy Lines Within Containment (Cont'd.)

                                !            Line           Operating    Operating Item                                Functional        Pressure   Temperature No.       System                 lhseription       (>275 ps.lg)  (> 200*F)         Figure 57   Reactor Coolant      Direct Vessel injection     Yes           Yes        5.1.2-1 &

Connection #4 to SIS 6.3.2-!C Interior Check Valvc , 3 SI-217 58 Safety Rapid Depress. Line Yes Yes 5.1.2-3 Depressurization from Pressurizer to System RC-409 59 Safety Rapid Depress. Line Yes Yes 5.1.2-3 Depressurizati m from Pressurizer to System RC-408 60 CVCS letdown Line from Yes Yes 9.3.4-1, Loop 2B to 5.1.2-1 Regenerative Hx 61 CVCS letdown Line from Yes Yes 9.3.4-1 Regenerative Hx to Letdown Hx 62 CVCS letdown IJ.ie from Yes No 9.3.4-1  ; letdown Hx to Containment Penetration 63 CVCS Charging Line from Yes No 9.3.4-1 Containment Pen to Regenerative Hx 64 CVCS Charging Line fron' Yes Yes 9.3.4-1 Regenerative lix to RCS imop 2A 65 CVCS Auxiliary Spray Line Yes Yes 9.3.4-1, to Pressurizer Spray 5.1.2-3 Common IIcader 66 CVCS SCS lix Shutdown Yes Yes 9.3.4-1 Purification Line Cont Pen Check Valve CH-304 to Letdown fix 0 Asywond Des > Atatorial Design of SSC (11/96) Page 3.6-44

                                                                                    -              =             _ -. .

System 80+ oesign contratDocument O V Table 3.6-3 High Energy Lines Within Containment (Cont'd.) Line Operating Operating item Functional Pressure Temperature , No. System Description (>275 psig) (> 200'F) Figure 67 CVCS RCP Seal Water Yes No 9.3.4-1 Header from Cont , Pent to Branch Seal Water Lines 68 SIS Safety Injection Line Yes Yes 6.3.2-1C

                                         #1 from Check Valve SI-543 to DVI Line
                                         #1 Junction 69    SIS                   Safety injection Line      Yes          Yes       6.3.2-1C 1                                         #2 from Check Valve SI-541 to DVI Line
                                         #2 Junction 70    SIS                   Safety injection Line       Yes         Yes       6.3.2-1C
                                         #3 from Check Valve SI-542 to DVI Line
                                         #3 Junction m

71 SIS Safety injection Line Yes Yes 6.3.2-lC

                                         #4 from Check Valve SI-540 to DVI Line
                                         #4 Junction 72     SIS                  Safety injection Line       Yes         No        6.3.2-IC from SIS Tank #1 to                                                            ;

Check Valve SI-247 l 1 73 SIS Safety Injection Line Yes No 6.3.2-lC { from SIS Tank #2 to Check Valve SI-227 74 SIS Safety injection Line Yes No 6.3.2 1C from SIS Tank #3 to Check Valve SI-237 75 SIS Safety injection Line Yes No 6.3.2-lC from SIS Tank #4 to Check Valve SI-217 76 SIS SIS Tank #1 Relief Yes No 6.3.2-1C Line to SIS Tank #1  ; Safety Valve SI-241 ] 77 SIS ' SIS Tank #2 Relief Yes No 6.3.2-lC Line to SIS Tank #2 gy

 ;%.))          -

Safety Valve SI-221 Anweewet Deeign neeendel Deeign of SSC Page 3.6-45

                                                                                                                         )

System 80 + Design C?ntrolDocument  ; 1 Table 3.6-3 High-Energy Lines Within Containment (Cont'd.) Line Operating Operating j Item Functional Pressure Temperature No. System Description (>275 psig) (> 200'F) Figure 78 SIS SIS Tank #3 Relief Yes No 6.3.2-IC Line to SIS Tank #3 Safety Valve SI-231 79 SIS SIS Tank #4 Relief Yes No 6.3.2-1C Line to SIS Tank #4 Safety Valve SI-241 80 SIS SIS Tank #1 Drain Yes No 6.3.2-IC Line to SI-641 81 SIS SIS Tank #2 Drain Yes No 6.3.2-IC Line to SI-621 82 SIS SIS Tank #3 Drain Yes No 6.3.2-IC Line to SI-631 83 SIS SIS Tank #4 Drain Yes No 6.3.2-IC Line to SI-611 EFW Emergeacy Feedwater Yes Yes 10.4.9-1 84 Line from Motor (Sheet 1) Driven EFW Pump #1 Cont Pen Check Valve EF-202 to SG #1 Common EFW Line 85 EFW Emergency Feedwater Yes Yes 10.4.9-1 Line from Steam (Sheet 1) l Driven EFW Pump #1 Cont Pen Check Valve E1 -200 to SG #1 Common EFW Line 86 EFW Emergency Feedwater Yes Yes 10.4.9 1 Common Line to SG (Sheet 1), l #1 Feedwater 10.4.7-1 l Downcomer 87 EFW Emergency Feedwater Yes Yes 10.4.9-1 I Line from Motor (Sheet 1) Driven EFW Pump #2 Cont Pen Check Valve l EF-203 to SG #2 Common EFW Line ( O Apyweved Deslyn Material- Design of SSC Page 3.646

9 t System 80 + ' oeshm conoorcoeumont

. l t

f" , i Table 3.6-3 High-Energy Lines Within Containment (Cont'd.)- j Line Operating Operating .} Item Functional Pressure Tessperasure  ; No. Systesa " Description (>275 peig) (> 200*F) - Hgure  ! 88 EFW Emergency Feedwater Yes Yes 10.4.9-1 i Line from Steam (Sheet 1) Driven EFW Pump #2  ; Cont Pen Check Valve i EF 201 to SG #2 Common EFW Line i 89 EFW Emergency Feedwater Yes ' Yes 10.4.9-1 Common Line to SG (Sheet 1),

                                             #2 Feedwater                                                               10.4.7-1

[ Downcomer  ; i e d G J l l Note: See Sections 3.6.1.1.1 and 3.6.1.1.2 for definitions of High and Moderate Energy lines. The following piping systems are considered Moderate Energy,'since these systems operate above the temperature and/or pressure limits of high energy status for only a relatively short portion (less than approximately two percent) of the time during which they perform their intended function, as prescribed by guidance of ANSI /ANS-58.2-1988: Safety injection System (SIS), Containment Spray System (CSS), Shutdown ~ Cooling System (SCS), and ' Emergency

                      . Feedwater System (EFW). Portions of these systems are, however, interfaced directly with
 -'(p)               : other High Energy systems or have sections which are High Energy in natme. These piping sections have been included in this table, wassen assender. semen er sec,                                                                                       p.,, 3. ga y

System 80+ Design controlDocument Table 3.6-4 High-Energy Lines Outside Containment (NoteI83) Line Operating Operating item Functional Pressure Temperature System Description (>275 psig) (> 200'F) Figure Note No. [2] MSIV Bypass Line Yes Yes 10.1-2, la Main Steam on LINE #1 10.3.2-1 (2) MSIV Bypass Line Yes Yes 10.1-2, lb Main Steam on LINE #2 10.3.2-1 I23 MSIV Bypass Line Yes Yes 10.1-2, 2a Main Steam on LINE #3 10.3.2-1 [2] MSIV Bypass Line Yes Yes 10.1-2, 2b Main Steam on LINE #4 10.3.2-1 [2] Main Steam Yes Yes 10.1-2, 3 Main Steam Atmospheric Dump 10.3.2-1 Lir on LINE #1 [2] Main Steam Yes Yes 10.1-2, 4 Main Steam Atmospheric Dump 10.3.2-1 Line on LINE #2 .. [2] Main Steam Yes Yes 10.1-2, 5 Main Steam Atmospheric Dump 10.3.2-1 Line on LINE #3 [2] Main Steam Yes Yes 10.1-2, 6 Main Steam Atmospheric Dump 10.3.2-1 Line on LINE #4 (2) 7 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #1 on LINE 10.3.2-1

                          #1

[2] 8 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #2 on LINE 10.3.2-1

                          #1

[2] 9 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #3 on LINE 10.3.2 1

                          #1 (21 10   Main Steam        Main Steam Safety      Yes         Yes       10.1-2, Valve #4 on LINE                             10.3.2-1
                          #1 (2) 11  Main Steam        Main Steam Safety      Yes         Yes       10.1-2, Valve #5 on LINE                             10.3.2-1
                          #1 O

Approwd Design Material Design of SSC Page 3.648

System 80+ Design ControlDocument Table 3.6-4 High-Energy Lines Outside Containment (Cont'd.) Line Operating Operating item Functional Pressure Temperature No. System Description (>275 psig) (> 200*F) Figure Note [2j 12 Main Steam Main Steam Safety Yes Yes 10.1-2, , Valve #1 on LINE #2 10.3.2-1 I*' 13 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #2 on LINE #2 10.3.2-1 [2j 14 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #3 on LINE #2 10.3.2-1 121 15 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #4 on LINE #2 10.3.2 1 12 16 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #5 on LINE #2 10.3.2-1 [2j ._ 17 Main Steam Main Steam Safety Yes Yes 10.1-2, d Valve #1 on LINE #3 10.3.2-1 121 18 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve h2 on LINE #3 10.3.2-1 [2j 19 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #3 on LINE #3 10.3.2-1 ('3 [2] 20 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #4 on LINE #3 10.3.2-1 121 21 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #5 on LINE #3 10.3.2-1 123 22 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #1 on LINE #4 10.3.2-1 12j 23 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #2 on LINE #4 10.3.2-1 [2j 24 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #3 on LINE #4 10.3.2-1 l' 25 Main Steam Main Steam Safety Yes Yes 10.1-2, Valve #4 on LINE #4 10.3.2-1 26 Main Steam Main Steam Safety Yes Yes 10.1-2, "' Valve #5 on LINE #4 10.3.2-1 12j 27 Main Steam Main Steam Lins Yes Yes 10.1-2, from LINE #1 to 10.3.2-1, Emergency Feed- 10.4.9-1 water Pump Turbine (Sheet 2)

                               #1 Isolation Valve EF-108 (O.
 .I   J V
          ?, a; Desye Ne95d-Deep of SSC                                                    Pope 3.6-49

System 80+ Design ControlDocument Table 3.6-4 Illgh-Energy Lines Outside Containment (Cont'd.) Line Operating Operating Item Functional Pressure Temperature System Description (>27S psig) (> 200'F) Figure Note No. I*' Main Steam Line Yes Yes 10.1-2, 28 Main Steam from LINE #4 to 10.3.2-1, Emergency Feed. 10.4.9-1 water Pump Turbine (Sheet 2)

                        #2 Isolation Valve EF-109 29    Main Steam      EFW Pump Turbine         Yes         Yes       10.4.9-1
                        #1 Drain Header to                             (Sheet 2) 4" x 1* Reducer 30    Main Steam      EFW Pump Turbine         Yes         Yes       10.4.9-1
                        #2 Drain Header to                             (Sheet 2) 4" x 1" Reducer Main Steam      Instmment Line from      Yes          Yes      10.4.9-1 31 Junction of line to                            Sheet 1 Turbine #14" x 1" Reducer to Valve EF-298 32     Main Steam     Instrument Line from     Yes          Yes      10.4.9 1 Junction of line to                            Sheet 2 Turbine #2 4" x 1" Reducer to Valve EF-299

[23 Main Steam Balance of Main Yes Yes 10.1-2, 33 Steam Piping on 10.3.2-1 LINE #1 to Turbine Bldg [2) 34 Main Steam Balance of Main Yes Yes 10.1-2, Steam Piping on 10.3.2-1 LINE #2 to Turbine Bldg (2) 35 Main Steam Balance of Main Yes Yes 10.1-2, Steam Piping on 10.3.2-1 LINE #3 to Turbine Bldg (21 36 Main Steam Balance of Main Yes Yes 10.1-2, Steam Piping on 10.3.2-1 LINE #4 to Turbine Bldg 37 SG SG #1 Common Yes Yes 10.4.8-1 Blowdown Blowdown Line from Containment Pen to Turbine Bldg Approvmf Design MotorW Design of SSC Page 3.f 50

System 80+ oesi.gn contratDocument O Table 3.6-4 High-Energy Lines Outside Containment (Cont'd.) ( ,/ Line Operating Operating Item Functional Pressure Temperature No. System Description (>27S psig) ( > 200'F) ' Figure Note 38 SG SG #2 Common Yes Yes 10.4.8-1 Blowdown Blowdown Line from Containment Pen to Turbine Bldg 39 Main SG #1 Main Yes Yes 10.1-2, Feedwater Feedwater Line 10.4.7-1 40 Main SG #1 Downcomer Yes Yes 10.1-2, Feedwater Feedwater Line 10.4.7-1 41 Main SG #2 Main Yes Yes 10.1-2, Feedwater Feedwater Line 10.4.7-1 42 Main SG #2 Downcomer Yes Yes 10.1-2, Feedwater Feedwater Line 10.4.7-1 43 CVCS Seal injection Line Yes No 9.3.4-1 from Charging Pumps (Sheets 1,4) ) to Seal Injection Hx 44 CVCS letdown Line from Yes No 9.3.4-1 (O Containment Pen to Both Letdown (Sheet 4) Pressure Reducers Downstream of . letdown Control I Valve Lines 45 CVCS Seal Injection Line Yes No 9.3.4-1 from Seal Injection (Sheet 4) Hx to Cont. Pen including both filter flow paths 46 CVCS Charging Pump #1 Yes No 9.3.4-1 l Line to Common (Sheet 1) l Charging Line j i l l l I -p l l 4 proved Desigrt heelerie!- Desigrr of SSC Page 3.6-51 l

System 80+ Design ControlDocurnent Table 3.6-4 High-Energy Lines Outside Contaimnent (Cont'd.) h Line Operating Operating item Functional Pressure Temperature System Description (>27S psig) (> 200'F) Figure Note No. 47 CVCS Charging Pump #2 Yes No 9.3.4-1 Line to Common (Sheet 1) Charging Line 48 CVCS Common Charging Yes No 9.3.4-1 Line to Containment (Sheets 1,4) Penetration including Flow Control Valve Piping Notes: . I1 See Sections 3.6.1.1.1 and 3.6.1.1.2 far .finitions of High and Moderate Energy Lines: The following piping systems are considered Mo43 * .sergy, since these systems operate above the temper' ture and/or pressure limits for only a relativeh 6 portion (less than approximately two percent) of the time during which they perform their intended function, as prescribed by guidance of ANSI /ANS-58.2-1988: Safety injection System (SIS), Containment Spray System (CSS), Shutdown Cooling System (SCS), and Emergency Feedwater System (EFW). Ponions of these systems are, however, interfaced directly with other High Energy systems or have sections which are High Energy in nature. These piping sections have been included in this table. Portions of systems which continued into the Turbine Building were generally excluded from this table by the guidance of Section 3.6.1.2. [2] The following nomenclature applies to Main Steam Lines: LINE #1 refers to Main Steam Line from Steam Generator #1, Line #1 LINE #2 refers to Main Steam Line from Steam Generator #1, Line #2 LINE #3 refers to Main Steam Line from Steam Generator #2, Line #1 LINE #4 refers to Main Steam Line from Steam Generator #2, Line #2 O Approved Design Material. Design of SSC page 3.6 52

                                                                                                          .         --         .                .          . ~ .

r- , ' System 80+ oeska controlDocument 1 l l I l POSTULATED BREAK i

                                                     +      "*---- GAP 8 G
                        *h              e,     g                     -                            -

l

                           ~7p;                    3 PROCESS PIPE l                       CRUSH PIPE                                 ,

U OFFSET O . a ' 6, _ (f -) BUILD #1G

                                                       '                                              ".  .      STRUCTURE F                        RESTRAINT STRUCTURE                        -

INITIAL PLASTIC ., HINGE LOCATION t k ) G l i l l i Typical Crmh Pipe Whip Restraint Configuration Figure 3.6-1 ( i

        - Apptosent W A0eteein!. Design of SSC                                                                              Pope 3.6-53 i

                                  . .   .- _ . ~ . - . . . . - .          .    .- .-   . .. _.       . -    - . . ~ . . , -_

] Sy~ tem 80+ i oestan contr:IDocument Effective Page Listing  ; Appendix 3.6A l l Pages Date Pages Date i, li 1/97 , iii Original

1 3.6A-1 through 3.6A-5 Original 3.6A-6, 3.6A-7 11/%

3.6A-8 through 3.6A-10 Original

                                                                                                                             -t j

J l i' 1 4 . j l I 1 i 1 Apprend one4pn neehwiel Dee@ of SSC . (USH Page I, B I i

System 70+ Desion control Document

    /3                                                       Appendix 3.6A V.

Supplemental Information on-Design and Analysis for Pipe Whip Contents Page 1.0 Selection of Restraint Type . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6A-1 1.1 Preliminary Selection of Pipe Whip Restraints . . . . .. . . . . . . . . . . . . . . . . 3 .6 A- 1 1.1.1 Energy Absorbing Restraints . . . . . . . . . . . . . . ....................3.6A-1 1.1.2 Rigid Restraints .......................... .... ... . . . . . 3.6A-2 1.2 Restraint Characterization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6A-2 2.0 Pipe Rupture Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6A-3 2.1 General Approach . . . . . 3.6A-3 (N .............. ........ . . . . . . . . . . . . Q '2.2 2.3 Procedure for Energy Balance Analysis .... Procedure for Dynamic Analysis with Simplified Models 3 . 6 A-3 3.6A 2.3.1 Approach 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6A-4 2.3.2 Approach 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 6A-4 2.3.3 WallImpact Analysis . .........................,....... . . . . 3.6A-5 2.4 Procedure for Dynamic Time-History Analysis Using Detailed Models . . . . . . . . . 3.6A-5 2.4.1 Modeling of Piping System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 6A-5

;         2.4.2   Dynamic Analysis Using Detailed Piping Model . . . . . . . . . .                        ...........                . 3.6A-5

3.0 Jet Impingement on Essential Piping and Components ........... . . . . . . . 3.6A-6 4.0 References ..............................................3.6A-6 Figures 3.6A - 1 Pipe Cmsh Bumper . .................................3.6A-8 3.6A - 2 Lamiru.ted Strap Restraint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 6 A-9 3.6A,- 3 Broken End and Unbroken End Boundary Conditions for Whipping

. Pipes .......... ................... . . . . . . . . . . . . . . 3. 6A- 10 g$ i Q /. AMwetmef Des /prP Me$sde/

• Doth of $$C Pepe &

System 80+ Design controlDocument n () 1.0 Selection of Restraint Type l l 1.1 Preliminary Selection of Pipe Whip Restraints Pipe whip restraints typically do not contact the pipe being restrained, but rather have a clearance  ! sufficiently large to envelope pipe motions under all normal and abnormal operating conditions. An energy absorbing design is generally used to minimize impactive loads and to dissipate the kinetic energy developed as the pipe whips through the clearance gap and contacts the pipe whip restraint. Elastic designs may also be used when gaps are small and pipe whip is minimized. Two types of energy absorbing restraints used are (Reference 4.1): the pipe crush bumper, a compressive member; and the laminated U-strap, a tensile member. The selection of the appropriate type is based on the direction of pipe motion following the postulated event and the direction to the nearest suitable building structure. If more than one direction of pipe motion is postulated, more than one restraint may be used near a given rupture location. An exception to the use of gapped rupture restraints occurs near containment penetration areas in which pipe stresses due to postulated ruptures beyond these areas must be minimized. For this application, the nearest rupture restraint is typically a dual-purpose pipe support, sized to accept rupture loads. These elastic restraints, when used in conjunction with energy absorbing restraints near the break, assure acceptable stresses in the exclusion area following the postulated rupture. j l 1.1.1 Energy Absorbing Restraints - Energy absorbing restraints are placed close to the postulated break and remote from plastic hinges which may form in the broken pipe. Thus, for longitudinal splits, restraints are situated at the postulated break location. For circumferential breaks, the thrust force, which is normal to the break, is applied primarily at the first change in flow direction so the restraint is placed near the first elbow on the opposite end from the break. Enough space is left between the restraint and elbow to allow in-service inspection of the pipe weld. If the first change in flow direction occurs at a large radius bend (e.g.,3D or SD), an additional restraint may be required on the bend. l i The energy to be absorbed is a function of the blowdown forces applied to the pipe and the distance traversed by the pipe as it closes the restraint gap and is subsequently brought to rest The development of the hydraulic blowdown forces is described in Section 3.6.2.2.2. The required gap is established by conservatively enveloping all pipe motions computed in the pipe stress analysis, as well as restraint  ; motions relative to the pipe due to containment LOCA expansion, seismic events, etc. It includes l clearances for pipe insulation and installation tolerances to assure that neither the pipe nor its insulation ( touches the restraint except following a rupture. However, side contact with the flexible laminated strap restraint is permitted. In lieu of detailed gap calculations, a censervatively large gap may be used. The distance the pipe moves before contacting the restraint is computed using the position of the pipe in the operating condition at which the pipe rupture is postulated to occur (Reference 4.9), i.e., hot standby or 102% power condition, whichever has the greater contained energy (Reference 4.2). The analyses for preliminary restraint sizing are described in Section 2.2 of this Appendix. A typical pipe crush bumper (Figure 3.6A-1) consists of a length of pipe oriented across the path of the whipping pipe and supported on a flat, rigid surface (Reference 4.1). It crushes locally at the point of p impact, absorbing energy. These restraints are made of seamless carbon steel pipe or tube. The force-Q deflection curve is nonlinear and begins to depart from essentially horizontal at a deflection of 85% of Awmd onian uomw on+ or ssc rare 3.sa.s

System 80+ Design controlDocument the inside diameter. In accordance with Reference 4.2, this defines the rated energy dissipating capacity. Restraints are designed to 80% of this capacity. The bumper pipe is sized and checked to assure that the crush occurs in the bumper pipe, rather than in the process pipe. The laminated strap restraint (Figure 3.6 A-2) consists of multiple layers of Type 304 stainless steel straps arranged in a "U" configuration (Reference 4.1). The broadened ends of the "U" are elastic and terminate by welding to lugs which are pinned to an anchored clevis. The elastic ends smoothly transition to a reduced cross-section which extends plastically during pipe whip, absorbing the energy. Once the clearance gap is closed and the slack is removed so the strap becomes taut, the force-deflection characteristics are proportional to the stress-strain curve of the material. The design limit is 50% of the ,miform ultimate strain of the material (Reference 4.2), the value of which is confirmed by stress-strain tests on the specific heats of material used. The design uses a low friction material on the side facing the pipe. This laminated design provides great flexibility. If the process pipe contacts the sides of the restraint during an event other than pipe rupture, negligible loads result. The thinness of the laminates also minimizes bending strains during fabrication of the "U" shape and subsequently, during impact, when the radius of the arch changes to that of the pipe. Thus, the strap acts mainly as a membrane during the postulated rupture event. 1.1.2 Rigid Restraints Restraints used to limit pipe motions, such as may be used to protect containment penetration areas, are designed as pipe supports. In addition to the requirements for pipe supports, these restraints are evaluated for rupture loads using the criteria applicable to rupture restraints. For rupture evaluation, the local stiffness of the pipe wall is considered and the contact area on the pipe is sufficiently large to minimize these local shell deformations. 1.2 Restraint Characterization Rupture restraints consist of two parts: the supporting structure and the energy absorbing component. Restraint support structures which do not also function as pipe supports are designed to the requirements of the AISC Steel Construction Manual (Reference 4.3), with the additional requirement that they be seismically rigid. In all applications involving the support of energy absorbing restraints, the support deflection is comparatively small and does not affect the impact load. Thus, in simplified analysis, the support structures are not explicitly included in the dynamic analysis. They are designed to a static load equal to the peak load in the energy absorbing member times a dynamic load factor (DLF) of 2. For detailed dynamic analysis, the support structures are modeled in sufficient detail to reflect their distributed l mass and stiffness propenies. The effective DLF varies throughout the support. The energy absorbing characteristics of a pipe crush bumper are based on the interpolation of test data (Reference 4.4) with consideration of the loading rate (Reference 4.5). While the curve may be used directly in energy balance analysis, for dynamic analyses it is represented as a multi-linear curve. In its simplest form, this may be a bilinear curve which gives a good fit to the force and energy absorption near the design point. O Approved Design Ataterial- Design of SSC Page 3.6A-2

System 80 + Deslan controlDocument \ Laminated strap restraints are modeled as an equivalent member having the'same force-deflection properties and effective length as the restraint. The procedure for establishing the properties has been checked against test results and considers the following parameters: l

• the restraint dimensions and geometry,especially the length and area of the reduced cross-section, e the diameter of the arch of the outside of the pipe, j i

e the stress-strain characteristics of the stainless steel strap; a 10% increase in force to account for strain rate effects may be used (Reference 4.2), e the initial iarance gap to the hot pipe and the additional effective gap due to slack take-up, and e the effects of off-axis impact. ' The force-deflection curve is idealized as a bilinear relation which, like the analytical model for the pipe crush bumper, accurately represents the force and dissipated energy near the design point. 2.0 Pipe Rupture Evaluation 2.1 General Approach f The approach used for evaluating ^he effects of pipe rupture is based on Reference 4.9. The specific method employed for pipe whip evJuation is generally determined by the nature of the problem and the {f size and pressure of the line being re strained:

• Energy balance analysis is the simplest form of analysis. Its use is confined to conceptual design and to the evaluation of restraints for small or relatively low pressure lines, especially the qualification of standard small line restraints. I l
• Simplified dynamic analyses are used to evaluate restraints for small and moderate size lines and to evaluate situations, such as concrete barrier impact, which are evaluated primarily by empirical relationships and which do not lend themselves to more detailed analysis.

1

                                                                                                                            )

I

• Detailed dynamic analyses are performed for all large line restraints and for the evaluation of containment penetration areas in any size line.

2.2 Procedure for Energy Balance Analysis J Energy balance analysis equates the work done by the blowdown thrust force to the energy absorbed in the restraint. This permits a designer to readily size the energy absorbing component and this approach is often used for initial restraint sizing. The work done is based on a quasi-steady-state fluid force times the distance traveled, including the deflection of the restraint. Energy absorbed by the pipe, as at a

       . plastic hinge, is conservatively ignored. The steady-state fluid forcing function is derived in accordance with Section Ill.2.c(4) of Reference 4.2. If the approach is used for final design, typically for small lines, the approach follows the requirements of Reference 4.2 and includes an amplification factor of 1.1 on the fluid forcing function to account for a possible maximum reaction beyond the first quarter cycle
     ~

of response. Anwwwmf Dwipn MewW- Du&n of SSC Page 3.6A.3 t I

System 80+ Design ControlDocument l 2.3 Procedure for Dynamic Analysis with Simplified Models h Simplified dynamic analysis models involve closed-form solutions for the pipe whip event, as detailed in Reference 4.9. Two forms of ahalysis are used, both being enhancements of the energy balance approach in which the time domain is explicitly conridered. As in energy balance analysis, an amplification factor of 1.1 is applied to the fluid forcing function. 2.3.1 Approach 1 This extension of the enegy balance evaluates the fluid thrust force as a time dependent function using the methods described in Section 3.6.2.2.2. Considering the distance from a fluid reservoir and the consequent frictional head losses in the pipe, a time-dependent thmst coefficient K, less than the theoretical steady-state maimum for frictionless flow (K = 1.26 for steam, saturated water, or steam water mixtures, or 2.0 for subcooled nonflashing water) is established. The maximum value of K, occurring at any time, is then used in an energy balance analysis identical to that described in Section 2.2. For the special case of a postulated circumferential break at the end of a ID or 1.5D elbow, the initial force pulse is ignored since the decompression wave travels around the elbow at sonic speed (almost instantaneously). During this very brief time, the pipe has a low velocity (initially at rest) and travels through a very small distance (<0.001"), such that the work done by the fluid during this time is negligible. 2.3.2 Approach 2 A further extension of this approach involves the determination of the duration of the dynamic event and the distribution of a portion of the energy to the pipe. Details of the procedure for a circumferential break, during the period prior to restraint impact, are provided in Reference 4.5. Extensions of this procedure are applicable to longitudinal breaks and to the period of restraint deflection up to the time of maximum response. The approach starts by idealizing the pipe both at the broken end and at the " fixed" end which may occur at a point of support or at an inertial constraint, either at mid-span or at an elbow (Figure 3.6A-3). Reference 4.5 describes the derivation of the equations that establish which end condition is applicable. The requisite information on the inelastic properties of elbows are established using the method of Reference 4.6. The approach has the following features:

• lumped masses (e.g., valves) are considered.
• the angular acceleration of the pipe about the plastic hinge (or hinges for a longitudinal split) is established, e the location of the plastic hinges is used to establish the relative motion of the pipe at the break compared to the smaller motion at the restraint,
• the pipe rotation at the plastic hinge and the deformation of the restraint both contribute to energy absorption, and
• the time to peak deflection is established, permitting a review of the fluid forcing function to establish an appropriate value for K.

Approved Desigrs MaterW. Desiger of SSC page16A.4

Sv~ tem 80+ oenlan control Document , } t . ,( 2.3.3 WallImpact Analysis-When separation between essential systems is achieved through the use of building structure, the concrete l wall is evaluated using the simplified analysis model fully described in Reference 4.5. The evaluation

             ' occurs in three parts:

. e The procedure of 2.3.2 is used to evaluate the whipping pipe and establishes the impact energy

                         . and velocity. The pipe is then considered as a missile impacting the wall.

l c .e- The dynamic crush stiffness of the pipe at the point of impact, typically at an elbow, and the l l impact area are established (Reference 4.5). 7 4 ! 'e The barrier is then evaluated for penetration, perforation and scabbing (References 4.5 and 4.7). Factors such as cubicle pressurization, other structural loads (e.g., deadweight of building) and . jet impingement from the other end of the ruptured line are included, as applicable. .

' 2.4 Procedure for Dynamic Time-History Analysis Using Detailed Models ,

Detailed dynamic nonlinear time-history analysis is performed using a finite element computer program )' such as ANSYS or other NRC approved computer codes. (See Section 3.9.1.2.) , l i 2.4.1' Modeling of Piping System

The piping, pipe supports and whip restraints are modeled using the same procedures used for pipe stress k

+

             . analysis. However, due to the nature of the postulated event, there are differences in several details:

o elastic-plastic properties are specified foe the piping and energy absorbing restraints, 4

e. restraint gaps are included, and t

e. except for analyses of containment penetration regions, the physical extent of the piping included ,

. in the model may be limited to the portion which actively participates in the dynamic response (Section 2.3.2). j Pipe supports not designed to carry pipe rupture loads might fail under these loadings and are selectively eliminated from the model when the loads exceed their capacity as established through engineering - i judgement and fragility data, l < 2.4.2' Dynamic Analysis Using Detailed Piping Model The nonlinear dynamic response of the piping / restraint system to time-dependent fluid forces applied at each change of flow direction or cross-section is computed by time step integration. The time step is

              . sufficiently shott to assure an accurate solution. Using program options for time step optimization, the                 l E time step may vary during the course of an analysis, being shortest near times of irapact. The analysis                   i

. ' continues past the time of peak restraint response, to a time when the kinetic energy has been essentially  ; dissipated and maxima are no longer occurring. The results of the analysis include time varying system 3 displacements, piping stresses, restraint and support loads, and restraint strains or deflections. i - g V  : The results of the dynamic analysis are evaluated to demonstrate that design allowables are met. The 1 I [  : evaluation shows that: 4preuss sentre asses, der . Osmise er ssC Pope 3.5M  !

System 80+ Design controlDocument (1) pipe stresses in containment penetration areas meet the limits specified in Reference 4.8, (2) loads and accelerations on containment isolation valves / operators are within qualified limits, (3) strains or displacements in energy absorbing restraints are within design allowables, and (4) loads on equipment are within qualified allowables. 3.0 Jet Impingement on Essential Piping and Components The criteria and procedures for evaluating jet impingement on essential piping and components follows ANSI /ANS-58.2-1988 (Reference 4.9), which includes the option to use NUREG/CR-2913 (Reference 4.10). The standard clearly defines:

• the fluid jet shape and direction, e the jet blowdown force, e the jet impingement force, including shape factor, and
• the jet impingement temperature.

The qualification of essential piping and compors includes jet impingement in Service Level D. Concurrent loads include deadweight, thermal, scwdc, pressure, and concurrent effects associated with the postulated break. The effective range for jet impingment force from piping containing steam or subcooled flashing water at pressures between 870 and 2465 psia and with no greater than 70*C subcooling may be limited to ten times the nominal pipe diameter (Reference 4.10). The suddenly applied impingement load is factored by a DLF of 2 unless lower dynamic amplifications are justified; snubbers are assumed to be locked. Seismic loads may be combined with the dynamic jet load by SRSS. The dynamic event may be followed by a long term steady-state blowdown. Snubbers are assumed inactive during this duration of jet loading. This steady-state blowdown load is combined by absolute sum with seismic loads. 4.0 References 4.1 U.S. Patents 4080998,4101117 and 4101118, Impact Energy Absorbing Pipe Restraints. 4.2 Standard Review Plan 3.6.2, " Determination of Rupture L.ocations and Dynamic Effects Associated with the Postulated Rupture of Piping," Rev.1, USNRC, July,1981. 4.3 Steel Construction Manual, 9th Edition, American Institute of Steel Construction, Chicago, Illinois. 4.4 Pecch, J. et al., " Local Crush Rigidity of Pipes and Elbows: Experiment, Analysis and Application," Paper F3/8, Fourth International Conference on Structural Mechanics in Reactor Technology, Conunission of European Communities, Brussels, Belgium,1977. Approved Design Materief - Desen of SSC (11/96) Page 3.6A-6

        - Sv-tem 80+                                                                     n--'~n cond Documart 4

4.5 Roemer, R. E., et al., " Evaluation of Pipe Whip Impact to Concrete Barriers: A Simplified Approach, Second ASCE Conference on Civil Engineering and Nuclear Power," Volume IV, pp. i 5-5-1 to 5-5-35, American Society of Civil Engineers,1980. I'

j. 4.6 East, G. H., "A Parametric Approach to Inelastic Elbow Flexibility," Paper 80-C2/PVP-128, American Society of Mechanical Engineers, New York,1980.  ;

4.7 Topical Report No. SWECO 7703, " Missile-Barrier Interaction," Stone and Webster Engineering , Corporation, September,1977. 4.8 Branch Technical Position MEB 3-1, " Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment," Rev. 2, USNRC, June,1987. 4.9 ANSI /ANS-58.2, " Design Basis for Protection of Light Water Nuclear Power Plants Against the Dynamic Effects of Postulated Pipe Rupture," American Nuclear Society, LaGrange Park, Illinois. 4.10 Weigand, G. G., et al., NUREG/CR-2913, "Two-Phase Jet Loads," Sandia National Laboratories,1983.  ; e 4 Y O l e j O V

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        ^. , -- : De*@n MenerW. Design of SSC                                                                 pp ga.y

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(C) ELBOW (OUT-OF-PLANE BENDING) (G)lMASS NEAR ELBOW l F f a v i if l (D) LONG STRAIGHT PIPE l l h (INERTIALLY CONSTRAINED) l GENERAL IDEAllZATION OF BROKEN END BOUNDARY CONDITION APPLICABLE TO (E),(F) AND (G) I  : l l l Bruken End and Unbroken End Boundary Conditions for Whipping Figure 3.6A - 3 i Pipes Approwd Design Material- Design of SSC Page 3.6A 10

f System 80+ oestan contmlDocument F 3.7 Selsinic Design j s ((The_ COL applicant referencing the System 80+ Standard Design will provide site and plant-specific i seismic design information which will include evaluation of interactions between Seismic Category I and non-Seismic Category I structures, systems, and components; procedures for pre-earthquake plannmg and i post-earthquake actions; details of inelastic piping analysis methods, if used; details of independent  : support motion response spectrum analysis methods, if used; and site-specific confirmatory analyses, as ' required.))3 3.7.1 Seismic Input i This section discusses the seismic design parameters and methodologies being used for the design of those i systems and subsystems imponant to safety and classified as Seismic Category I in Section 3.2. t 3.7.1.1- Design Response Spectra The System 80+ Standard Design as defined by this document is not based on a specific site. The design response spectra which define the free field design ground motion or control motion specified either at the site soil surface or on a hypothetical rock outcrop are shown in Figure 2.5-5. Generic site conditions i were selected to cover a range of possible conditions for the System 80+ sites. For the Nuclear Island, sets of representative cases from each of four generic site categories were evaluated. Grourri surface and foundation level spectra which correspond to the design response spectra of control motiore CMS 1, CMS 2 and CMS 3 for rock and soil cases are shown in Section 2.5. Ont of 12 soil cases analyzed in  : Section 2.5.2, ten are used in the soil structure interaction (SSI) analyses. The two cases eliminated in  ! the SSI analysis (B3 and D1) were non-governing cases whose soil response levels were enveloped by 4 other cases. See Section 2.5.2 for details of this analysis phase. Two rock cases were analyzed, one with no backfill (fixed base at bottom of basemat) and one with concrete backfill (fixed base at all subsurface elevations). The ten soil cases and the two rock cases were analyzed for all three control motions (CMS 1, CMS 2, and CMS 3). A sensitivity study was performed to demonstrate that the selected motions and the soil profiles provide a conservative design envelope for the System 80+ structures. In this study, a simplified analytical model of the 3D Ni structure was developed and analyzed with five soil profiles and the CMS 2 control motion. Three of the soil profiles were part of the 12 profiles selected for the System 80+ 3D SSI

!         analyses (B-1, B-1.5, B-2). The remaining two soil profiles were developed to serve as " test" profiles.

The two new profiles were chosen such that they have low strain soil properties that are in-between the soil properties of cases B-1, B-1.5 and B-2. Hence, they were named B-1.25 and B-1.75. Response parameters such as maximum in-structure acceleration, maximum base shear and maximum base overturning moment were used as the key parameters that determine the adequacy of the soil profile selection. The sensitivity analyses showed that structural response corresponding to the " test" soil cases B-1.25 and B-1.75 was under the envelope of structural response from the three generic cases B-1, B-1.5 ) and B-2. Therefore, it is concluded that the 12 generic soil profiles provide a conservative envelope of structural response and they cover a broad range of sites. The effect of differential seismic displacement on the equipment and supports is included in the analysis as described in Section 3.7.3.1. f'N - J. V .I I 1 COL information hem; see DCD Introduction Section 3.2.

          + , o > asenw.o > .rsse                                                                               ree s.7-r

System R0 + Design ControlDoctement 3.7.1.2 Design Time IIistory Since the System 80+ Standard Design is designed for generic site conditions, for the time history method of analysis, the generic free-field ground surface time histories are used as control motions in the analyses. In the soil-structure interaction analyses, for each generic site, the corresponding two horizontal and one vertical time histories at the free-field ground surface are used with the SSI model of that site. For the fixed-base analyses, the rock outcrop time histories are directly used as the control time histories. The response spectra at 2,5 and 7% damping of control motion CMSI, and 1,2,5 and 7% damping of control motions CMS 2 and CMS 3 and the corresponding spectral ordinates of the matching time histories are shown in Figures 3.7-1 to 3.7-12. The Power Spectral Densities of all time histories are included in Section 2.5. Each time history that is used in the SSI and rock analyses contains 20.48 seconds. For the SSI analyses, a time step of 0.005 seconds is used. For the Nuclear Island rock analyses, a time step of 0.0025 seconds is used. For Category I structures not on the Nuclear Island a time step of 0.005 seconds is used for both SSI and rock analyses. 3.7.1.3 Critical Damping Values Damping values used for various nuclear safety-related structures systems and components are based upon l Regulatory Guide 1.61 or alternative requirements for uniform envelope response spectrum piping l analysis. These values are expressed in percent of critical damping and aro given in Table 3.7-1. 3.7.1.4 Supporting Media for Seismic Category I Structures Category I structures are founded directly on rock or competent sof.. For the Nuclear Island the foundation embedment depth for System 80+ standard plant is approximately 51 feet (Reference 7). The rock properties and the layering characteristics, including shear wave velocity, shear modulus, and density, are given in Section 2.5. The System 80+ Nuclear Island is designed for the range of soil conditions discussed in Section 2.5 and shown in Appendix 3.78. 3.7.1.4.1 Soil Structure Interaction (SSI) Two different types of analysis methodologies are used for the seismic analyses for the Nuclear Island. For the fixed-base cases, modal superposition time history analyses are performed using the three control motions (CMSI, CMS 2, and CMS 3) corresponding to rock site conditions. When a structure is supported on soil, the SSI is taken into account by coupling the structural model with the soil medium. To accomplish this, the methodology of the computer program SASSI (System for Analysis of Soil Structure Interaction, Reference 6) is used. Detailed methodology and results of the SSI analysis for the Nuclear Island are presented in Appendix 3.7B. The methodology for the soil structure interaction for the non-Nuclear Island structures is presented in Appendix 3.7C. O APPronNi Design Ataterial- Design of SSC (11/96) l' age 32-2

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~
-             ; System 80+                                                                     oestan controlDocument i

3.7.2 Seismic Systeen Analysis  ; [ [/ X_ 3.7.2.1' Seismic Analysis Method  ; 3.7.2.1.1 hl==ic Category I Structures, Systeens, and Components Other 'Iban NSSS ' Seismic Category I structures, systems and components are identified in Table 3.2-1. The Nuclear Island . 4 - (NI) structures are modeled as lumped mass stick models for the _ seismic analysis. Figures 3.7-13  ! j through 3.7-17 show typical sketches of all NI structures. Figures 3.7-19 and 3.7-20 are schematic

representations of the combined structural model of the NI.' ..,

l Further details of dynamic modeling of building structures for seismic analysis are described in Section j 3.7.2.3. The horizontal model is analyzed for the plant E-W direction and N-S direction excitations and i the vertical model for vertical excitation. The results are then combined as described in Section 3.7.2.6. l The seismic analysis of the above systems is performed by one of the following methods: l t 3.7.2.1.1.1 Response Spectrum Method of Analysis } The response of a multi-degree-of-freedom system subjected to seismic excitation is represented by the l following differential equation of motion: [M] [{R} + (0,}] + [C] {x} + [K] {x} - 0 O l where. .V [M] = mass matrix (n x n)

                  . [C] = damping matrix (n x n) l

[K] '= stiffness matrix (n x n) . . l 1 {X} = column vector of relative displacements (n x 1)

                  '{X} = column vector of relative velocities (n x 1)                                                                   !

l h (R). = column vector of relative accelerations (n x 1) n = number of dynamic degrees of freedom {0,} = column vector of ground accelerations (n x 1) j l

     %/
                                                                                           .                                            l

. Anemen oneen anenenw-oneon a sac rene 2.7 2 l 1

System 80+ Design control Document in the response spectrum method of analysis, the equations of motion are decoupled using the transformation: {X} = [4] {Y} where: [4] = mode shape matrix {Y} = vector of normal, or generalized coordinates (m x 1) m = number of modes considered The decoupled equation of motion for each mode is transformed to a single degree of freedom system:

       ?) + 2 X;wj  t j + uf Yj - -T j 0, where:

Yj = generalized coordinate of j* mode A;

              =       damping ratio for the j* mode expressed as fraction of critical damping wj        =       circular frequency of j* mode of the system

{4j}T [M] {1} Tj = modal participation factor of the j* mode = {4j}r [M] {4;} The generalized maximum response of each mode is determined from: S-Yj (max) - Td "i where S,j si the spectral acceleration corresponding to frequency w). The maximum displacement of node i relative to the base due to mode j is: X g(max) = 4,3Y,(max) The modal response Xy (max) is used to 6termine other modal response quantities, such as forces. Since,

   - the maximum modal response does not occur at the same time, modal superposition is done to obtain the final response by the double sum method described in Section 3.7.2.7.

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Approved Design Metwiel Desigte of SSC Page 3.7 4

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        .3.7.2.1.1.2            Time History Method-

The solution of the differential equation of motion given in Section 3.7.2.1.1.1 can be obtained by the i method of modal superposition or by the method of direct integration. .

4 * . Modal Superposition Method The modal superposition method is used when the equations of motion can be decoupled as given { i in Section 3.7.2.1.1.1. Then the decoupled equation of motion for each mode is integrated using , a proven technique, and the total response is obtained by superposition method. e Direct Integration Method ' In this method, direct integration of the equations of motion by either implicit or explicit methods of numerical integration are used to solve the equations of motion. ,

                                                                      ~

For commonly used implicit methods, AT is not larger than 1/10 of the shortest period of interest. For explicit methods, the time step is also a function of the element size used in the model and is established on the basis of element size to ensure stability of the response. 3.7.2.1.1.3 Soil-Structure Interaction Analysis ' The soil-stmeture interaction analyses were performed using the substructure method formulated in the frequency domain using the complex response method and the finite element technique. The methodology of the computer program SASSI was used with a modified approach to compute the impedance and scattering of the soil / foundation system. Appendix 3.7B describes in detail the SSI analysis approach for the' System 80+ structures. A brief summary of the method is described below.

                                                                                                                           ~

In a sub-structuring method, the soil strata and halfspace are analyzed first in the frequency domain. From this analysis the impedances at the soil-structure interface are established, i Subsequently, the impedances are combined with a model of the superstructure, the control motion is - applied to the combined system, and the equations of motion are solved for computation of final accelerations and displacements. For the System 80+ analyses of the Nuclear Island, a modified SASSI methodology is used, which reduces the solution of the SSI problem to three steps: o Solution of the site response problem to' determine the free-field motions within the embedded part of the structure.

e. Evaluation of the foundation impedances.

e Solution of the structural problem. This involves forming the complex stiffness matrices and load vector and solving the equations of motion for the final accelerations.

  . A b      ' Figures 3.7-21 and 3.7-22 show schematic diagrams of the SSI analysis process. For the analysis using the CMS 2 and CMS 3 motions, the rock outcrop motion (R) is convolved through the soil media to j

Page 3.7 5 L 2 Dee@n etenerief- Deekn of SSC 7

Sy~ tem 80+ Design controlDocument produce the surface motion (S) and foundation level motion (F). The computed surface motion (S) is applied as the control motion in the SASSI SSI model at the free-field ground surface. For the CMS 1 analysis, the CMSI motion is applied directly at the free-field ground surface. 3.7.2.1.1.3.1 Computer Programs Used in Soil Structure Interaction Analysis 3.7.2.1.1.3.1.1 SASSI The SASSI program is used in soil-structure interaction analyses of the Nuclear Island and it is based on the Flexible Volume Sub-structuring Method. This method is a general sub-stmeturing technique, which uses the finite element method and solves the equations of motion in the frequency domain using the method of complex response. The SASSI sub-structuring scheme provides rigorous analytical solutions in each step of the SSI problem. In the Flexible Volume Method, the complete soil-structure system is divided into two substructures: the

 " foundation" and the " structure." The mass and stiffness of the " structure" is reduced by the corresponding properties of the volume of excavated soil. The mass and stiffness of the excavated soil are retained within the " foundation" model. The impedance problem is solved using the " foundation" model, and consists of a series of axisymmetric solutions of a layered site to applied point loads. In the System 80+ analysis, the SASSI standard analysis methodology is modified, as discussed in Appendix 3.7B. and the solution of the SSI problem is reduced to three steps:

• Solution of the site response problem to determine the free-field motions within the embedded part of the structure.
• Solution of the impedance and scattering problem.

O

• Solution of the structural problem. This involves forming the complex stiffness matrices and load vector and solving the equations of motion for the final accelerations.

The version of the SASSI program used in the soil-structure interaction (SSI) analysis of the System 80+ is version 4.0, dated June 1989. The SASSI program is extensively verified and validated and documented using three different methods of verification and correlation:

• Correlation to results of problems with closed form solutions, such as site response and response of simplified structural systems.
• Correlation to solutions of other well known SSI computer codes in the industry such as CLASSI and FLUSH.
• Correlation to experimental results, such as Lotung Large Scale Experiment sponsored by the Electric Power Research Institute / Nuclear Regulatory Commission / Taiwan Power Company, and others.

3.7.2.1.1.3.1.2 SIIAKE The SHAKE program is based on the one-dimensional wave propagation method, adapted for use with acceleration time histories through the Fast Fourier Transform algorithm. SHAKE incorporates nonlinear Apfcoved Design Matedel- Design of SSC Page 3.7-6 l

i i Sy ~ tem l'0 + Design controlDocument hV soil behavior and the effect of the elasticity of the base rock. The soil system consists of horizontal layers which extend to infinity in the horizontal direction. Each layer is homogeneous and isotropic and is characterized by thickness, mass density, shear modulus and damping factor. The non-linearity of the shear modulus and damping is accounted for by the use of equivalent linear soil properties using an iterative procedure to obtain values for modulus and damping compatible vith the effective strains in each layer. i 3.7.2.1.2 Seismic Analysis Method for the NSSS l l 1 3.7.2.1.2.1 lutroduction  ! The major components of the reactor coolant system are cesigned to the appropriate stress and j deformation criteria of ASME Code, Section !!!, for the set of loadings included in the component design I I specification. The adequacy of seismic loadings used for the design of the major components of the reactor coolant system are confirmed by the methods of dynamic analysis employing time history and response spectrum techniques. The major components are the reactor vessel, the steam generators, the reactor coolant pumps, the reactor coolant main loop piping, the surge line and the pressurizer. l l l Detailed dyne c models of the building structures and the NSSS are generated. Based on these detailed models, equivrJent, simplified dynamic models are developed. The simplified building and NSSS models are combined ud translated into a form suitable for input to the SSI analysis code (see Section 3.7.1.4.1). ) A number of soil cases are modeled and the time history analyses are performed. The soil cases are  ; chosen to envelope all potential building sites. The results of these analyses are contained in Appendix ( 3.711. These results, the simplified building model(s), and the detailed NSSS model are used to perform q j the analysis discussed in Section 3.7.2.1.2.3. A composite three-dimensional lumped-mass model of the reactor vessel, the two steam generators, the four reactor coolant pumps, the pressurizer, and the interconnecting main loop piping is coupled with a three-dimensional lumped-mass model of the reactor building for performing the analysis of these dynamically coupled components of the reactor coolant system. In addition, the representation of the reactor vessel as.cnnbly used in this coupled model includes sufficient detail of the reactor internals to account for possible dynamic interaction between the reactor coolant system and internals. The seismic input excitation is the translational and rotational time history motions at selected locations in the reactor building. The results of this analysis include appropriate excitation data for use in separate analyses of the surge line and of a more detailed model of the reactor internals. A model of the coupled components of the reactor coolant system is shown in Figure 3.7-23. A model of the pressurizer is shown in Figure 3.7-24. The analysis of the surge line piping employs a separate mathematical model and utilizes either response spectrum or time-history techniques as described in Section 3.7.3.1. The square root of the sum of the squares (SRSS) methods is normally used to combine the modal responses when the response spectrum modal analysis method is employed. In those cases, however, where modal frequencies are closely l I spaced, the responses of the closely spaced modes are combined by the sum of the absolute values method and, in turn, combined with the responses of the remaining significant modes by the square root of the sum of the square method, as described in Section 3.7.2.7. Contributions from all significant modes of response are retained in the analyses. The damping factors used in analysis of Seismic Category I structures, systems and equipment are

• (V] selected from Table 3.71.

Neww ony marea- on+ or ssc rag,17 7

System 80 + Design ControlDocument The damping factors given in Table 3.7-1 include those recommended in Regulatory Guide 1.61. The results of the dynamic analyses of the major components of the reactor coolant system, which are performed to confirm the adequacy of the seismic design, are contained in Appendix 3.7A. 3.7.2.1.2.2 Mathematical Models In the descriptions of the mathematical models that follow, the spatial orientations are defined by the se-of orthogonal axes for which Z represents the vertical direction and X and Y are in the horizontal plane in the directions indicated on the appropriate figure. The mathematical representation of the section properties of the structural elements employs a 12 x 12 stiffness matrix for the three-dimensional space frame models, and employs a 6 x 6 stiffness matrix for the two-dimemional plane frame model. Elbows in piping runs include the in-plane /out-of-plane bending flexibility factors as specified in the ASME Code, Section III. A schematic diagram of the composite mathematical models used in the analyses of the dynamically coupled components of the reactor coolant system is presented in Figures 3.7-23 and 3.7-24. These models include 36 mass points with a total of 96 dynamic degrees of freedom to represent the RCS, including the pressurizer. Additional mass points and dynamic degrees of freedom, not shown in the figures, are used to represent the containment and interior structures in the coupled seismic model. The surge line is very flexible relative to the rest of the structure, and is not considered in the coupled model analysis. The pressurizer is mathematically coupled to the remainder of the RCS by way of the building structure represented in the coupled seismic model. The mass points and corresponding dynamic degrees of freedom are distributed to provide appropriate representations of the dynamic characteristics of the components, as follows:

• The reactor vessel, with internals, is represented by 4 mass points with a total of 11 dynamic O

degrees of freedom.

• Each of the two steam generators is represented by 4 mass points with a total of 10 dynamic degrees of freedom, each of the four reactor coolant pumps is represented by 2 points with a total of 6 dynamic degrees of freedom.

e The pressurizer is represented by 6 mass points with a total of 13 dynamic degrees of freedom; each branch of cold leg piping is represented by a mass point with 3 dynamic degrees of freedom.

• Each branch of hot leg piping is represented by a single mass point with 2 dynamic degrees of freedom. The representation of the reactor vessel internals is formulated in conjunction with the analysis of the reactor vessel internals discussed in Section 3.7.3.14, and is designed to simulate the dynamic characteristics of the models used in that analysis.

The mathematical model provides a three-dimensional representation of the dynamic response of the wupled components to seismic excitations in both the horizontal and vertical directions. The mass is distributed at the selected mass points and corresponding translational degrees of freedom are retained to include rotary inertial effects of the components. The total mass of the entire coupled system is dynamically active in each of the three coordinate directions. Surge Line A lumped parameter, multi-mass mathematical model is employed in the analysis of the surge line. A representative model is shown schematically in Figure 3.7 25. The surge line is modeled as a three-I Approved Design Acatorie!* Design of SSC Page 3.7-8

System 80+ Design ControlDocument dimensional piping run with end points anchored at the attachments to the pressurizer and the reactor b vessel outlet piping. All suppons defined for the surge line assembly are included in the mathematical model. The total mass of the surge line is dynamically active in each of the three coordinate directions. The surge line is analyzed as uncoupled from the reactor coolant system, using the motions of the hot leg, pressurizer and suppons as input. 3.7.2.1.2.3 Analysis Modeling and analysis of the coupled components of the reactor coolant system and the pressurizer are performed using ANSYS. A description of ANSYS is given in Section 3.9.1.2.1.13. Modeling and analysis of the surge line is performed using the SUPERPIPE code, a description of which is given in Section 3.9.1.2.1.4. Time history data for all six possible components of motion are applied simultaneously to the coupled building model to analyze the coupled components of the reactor coolant system. The responses to seismic excitation for the coupled components of the reactor coolant system are computed using the transient analysis capability of ANSYS. In the analysis of the coupled components of the RCS, excitations are input at selected points in the reactor building. For the coupled components of the RCS, the relative support displacements are inherently accounted for during the coupled analysis. The building motions derived from the soil-structure interaction analysis consist of six time histories at each location per soil case, three linear and three rotational. For each soil case all six time history motions are applied at each selected point of the coupled building model to analyze the coupled a components of the RCS. The calculated motions for input to subsequent subsystem analyses therefore include the motions caused by the foundation torsion and rocking. The response of the surge line in computed using the spectrum or transient analysis capability of SUPERPIPE. The surge line is seismically analyzed as a decoupled subsystem using methods described in Section 3.7.3.1. Input excitations at the endpoints for the surge line analysis are generated by ANSYS using the results of the coupled containment building interior structure, reactor coolant system analysis. The input excitations at the intermediate supports are developed from the building analyses. b The reaction forces and moments are obtained from the dynamic seismic analysis and are summarized in tabular form in Appendix 3.7A. Since the three directions of canhquake motion are statistically independent, the maximum responses are calculated by a simultaneous application of motion resulting from all three directions of canhquake. The maximum seismic loads calculated by the response spectrum techniques are the result of combining the modal rections due to both horizontal and venical excitations. The method of modal combination is discussed it. Section 3.7.2.7. The maximum responses due to each of the three earthquake components are then corrbined by the SRSS method. Tce scbrnic loadings specified for the design of the reactor coolant system components and suppons are greater than the seismic loads calculated by the dynamic seismic analyses. 3.7.2.2 Natural Frequencies and Response Loads p V) f These data :n provided in Appendix 3.7A. Appweved Design Material- Design of SSC Page 3.7 9

System 80+ Design ControlDocument 3.7.2.3 Procedures Used for Analytical Modeling 3.7.2.3.1 Modeling of the NSSS and BOP The procedure used for modeling NSSS components and interconnecting piping is described in Section 3.7.2.1.2. The procedure used for modeling BOP components and interconnecting piping is described in the following sections. 3.7.2.3.2 Designation of Systems Versus Subsystems The calculation of the dynamic response of a nuclear power plant subject to an earthquake loading is divided into two categories. The first is the " safety-related main structural system" and the second is the

  " safety-related subsystem." The " safety-related main structural system" category refers to the analysis of major buildings and structures which house and/or support safety- related systems. The " safety-related subsystems" category refers to smaller safety-related structures, systems, and components.

The major structures which are analyzed in the main structural system analysis are:

• Steel contairunent vessel, internal structure shield building, and Nuclear Annex.
• Diesel Fuel Storage and CCW Heat Exchanger Building.

3.7.2.3.3 Decoupling Criteria for Subsystems in general, all subsystems such as equipment and piping with the exception of the reactor coolant system are deceupled from the floor which supports them. As recommended in the Standard Review Plan, Section 3.7.2, the following decoupling criteria is used in instances where a subsystem needs to be modeled:

• If R < 0.01, decoupling is done fer any R f.
• If 0.01 < Rm < 0.1, decoupling is done if Rf :s; 0.8 or Rf 2: 1.25.
• If R > 0.1, an approximate dynamic model of the subsystem is included in the main structural system.

where: R* = t tal mass of the supported subsystem total mass of the supporting system R = fundamental frequency of the supported subsystem f dominant frequency of the support motion l The masses of the decoupled subsystems are included with the structural mass of the supporting floor g , slabs in the system model. The containment internal structure model includes a simplified dynamic model W of the reactor coolant system. l Assuned Design Material Design of SSC Page 3.7-10 l.

System 80+ Design ControlDocument 3.7.2.3.4 Lumped Mass Considerations ( The safety-related structures are modeled as a multi-degree of freedom system. The major structural element systems such as floor slabs, foundation mat, roof slab, shear walls and braced vertical frames are included in the model. All subsystems such as equipment and piping are considered in accordance with the decoupling criteria described in Section 3.7.2.3.3. For all seismic analyses, dead load plus 25 % of live load to account for miscellaneous piping and equipment mass is assumed to contribute to the inertial forces. In addition, mass associated with all heavy equipment was also included in the computation of floor masses. The Shield Building and Steel Containment Vessel are each analyzed using a single model. Two independent models are used for the seismic analysis of the Internal Structure. One model is used for the horizontal excitations and the other is used for the vertical excitation. Different models are used in order to better represent the structural behavior for the different excitations. Each of the models and the method of lumping the masses are described below. The seismic analysis of other Category I structures are performed utilizing one dynamic model for all three components of input motion (i.e., the vertical and the two translational dW ctions of excitation.) The dynamic models account for any coupling between the different components of motion and include all major structural components (i.e., walls, floors, and columns.) 3.7.2.3.4.1 Model for Horizontal Excitation - The Reactor Building (RB) and Nuclear Annex (NA) for System 80+ consists of the following structures: (

• Interior Structure (IS)
• Shield Building (SB)
• Steel Containment Vessel (SCV)
• Fuel Storage Area (FS) o CVCS/ Maintenance Area (CVCS)
• Diesel Generator Areas 1 and 2 (DG-1 and DG-2)
• Control Room Areas 1 and 2 (CAA and CAB)
• Emergency Feedwater Tank Areas 1 and 2 (EFW1 and EFW2)

The modeling approach that is used for the RB and NA structural model consists of developing a 3-D finite element model (FEM) and, based on the FEM model, developing equivalent 3-D lumped parameter stick models. This approach is used for all structures except the SCV. Because of its slenderness, the Steel Containment Vessel (SCV) has significant " membrane-type" action when it vibrates, and it is explicitly modeled with shell elements. The Interior Structure stick model section properties are adjusted to dynamically tune the stick model to f _(, match the dynamic characteristics of a 3-D finite element model of the complete Interior Structure. Mode shapes, natural frequencies, and mass participation of the predominant modes of vibration are matched. Page 3.711 Amewd On> hiewM Dn> of SSC _.a

System 80+ Design contro1 Document l The areas, shear and cross sectional, and moments of inertia are adjusted until a good match is obtained in the horizontal direction and then necessary adjustments are made to the vertical areas. The concrete shield building stick model is tuned to capture the predominant frequencies, mode shapes, and mass participation determined from an axisymmetric finite element model. The stick model i properties are s.djusted on the following bases:

• The cross sectional area is reduced in the dome to account for vertical displacement caused by bending of the dome.
• The shear areas in the dome are increased to account for the fact that the cross sectional area of a plane cut through the dome is greater than the radial cross sectional area.
• The moments of inenia are decreased as the shear areas are increased.

Using the above as guidelines, the shield building stick model properties are adjusted until a good match of dynamic properties with the finite element model is reached. The Nuclear Annex consists of venical walls and slabs that are regular shaped and do not exhibit any unusual dynamic characteristics. Therefore, no dynamic tuning is done for the Nuclear Annex stick models. 3.7.2.3.4.1.1 Developenent of FEM and Stick Models of the Interior Structure The FEM of the IS is developed by defining major floor elevations and major elevations at which significant stiffness discontinuities occur across the entire area of the structure. Twelve such elevations were selected, as follows:

 +50.00 ft.       Top of Basemat
 +68.50 ft.       Second Floor (Center of Slab)
 +90.25 ft.       Third Floor (Center of Slab)
+ 104.50 ft.      Steam Generator Supports
+ 114.00 ft.      Fourth Floor (Center of Slab)
+ 120.00 ft.      Top of Reactor Vessel
+ 144.50 ft.      Operating Floor
+ 164.33 ft.      Main Steam Line Supports
+ 178.00 ft.      Top of Steam Generator Shield Walls
 + 191.33 ft.     (stiffness discontinuity)
+210.00 ft.       Top of Crane Wall The load-resisting elements of each floor consist of concrete walls. These walls are modeled with quadrilateral shell elements or solid 8-node elements depending on the thickness of the walls. Concrete slabs of significant thickness are modeled with quadrilateral shell elements.

The translational mass and mass moments of inertia are lumped at the center of mass of each floor. This is done for ease of comparison between the full 3-D FEM and the equivalent 3-D stick model. The mass of each floor includes the mass of concrete walls, concrete slabs, concrete columns, heavy steel platforms, and heavy equipment. For light equipment, secondary structural steel, piping, tanks and miscellaneous mechanical and electrical components, a cumulative uniformly distributed mass is estimated and added to each floor. Approved Desigre Material- Destgrs of SSC Page 3.712

 -_ , _ .      -. - . -      -         . . - - .. -- - --_---- - . - - ~                                                        - - .

4 < r l Syrtem 80+ oestan controloocumart i Figure 3.7-13 shows a schematic of the stick model of the IS for the horizontal analysis, t 3.7.2.3.4.1.2 Development of FEM and Stick Models of the Shield Building Since the SB is symmetric about the vertical axis of the RB, the FEM of the SB is developed using an assembly of axisymmetric shell elements. Fixed-base modal analyses are performed for the horizontal i and venical directions and, based on these analyses, mass and stiffness propenies are selected for the SB stick model. The mass of the SB is lumped at eleven nodal points along the height of the stick. 4 f

3.7.2.3.4.1.3- FEM of Steel Conemininent Vessel

          ' The SCV is modeled with shell elements as shown in Figure 3.7-18. The bottom nodes, corresponding to elevation +91 ft., are connected with rigid links to the stick model of the IS.

3.7.2.3.4.1.4 Development of FEM and Stick Models for Fuel Building, CVCS/ Maintenance Area, EFW Areas, Diesel Generator Areas and Control Room Areas l The FB, CVCS, DG, EFW and CA stick models are developed following the procedure used in the development of the IS stick model. Each floor of each area is modeled with finite elements representing the main structural load-resisting elements of that floor. Subsequently, based on these models, equivalent  : i stiffness propenies are computed for each floor which are assigned to an equivalent beam element i representing that floor in the stick model of that area. 3.7.2.3.4.1.5 Combined Model of Nuclear Island and Nuclear Annex Structures i The combined model of the RB and NA structures is generated by linking the individual stick models of all the areas in the NI and NA complex. In addition, the dynamic model of the NSSS is coupled to the j_ IS stick model at the appropriate elevations. Because of the in-plane rigidity of the slabs, all sticks are connected with rigid links at each major elevation, as shown in Figures 3.7-13 to 3.7-17. The rigid links I [ provide in-plane rigidity only. All RB and NA structures are founded on a conunon basemar, the dimensions of which are given in i j Appendix 3.7B. , i 3.7.2.3.4.2 Model for Vertical Excitation 4 The previous discussion of the models developed for horizontal excitation applies to vertical excitation 8 model development, with minor changes in the case of the IS, FB, ErW, DG, CVCS and CA models. The only difference between the horizontal and venical analysis stick models is the eccentricity of the  : i

1. center of mass to the center of rigidity at each major elevation.

3.7.2.3.5 Modeling for Three Component input Motions

~

As discussed in Section 3.7.2.3.4, two independent models of the Nuclear Island, one in the horizontal  ; i and the other in the vertical direction, are used. The horizontal and venical models are decoupled, since the response in the venical direction due to horizontal excitation will be negligible and vice versa. In the horizontal analysis of all structures, the seismic model is analyzed along both the plant E-W and N-S  ! directions.  ; 1 z_._ = canon neenerner anoon erssc page 2.712 l l l 4

                                                                         .s. o ., ., . ,    , , , , . ,    . , , ,           ,.       %, g .m

System 80+ Design contro/ Document For other Category I structures a single coupled model is developed for simultaneous horizontal and vertical excitations. 3.7.2.4 Soil Structure Interaction (SSI) The soil model and SSI analysis methodologies are described in Appendix 3.78. 3.7.2.5 Development of Floor Response Spectra The time history method of analysis is used to generate the floor response spectra. The spectra are generated according to the procedure given in Regulatory Guide 1.122. As discussed in Section 3.7.2.3.4, the horizontal and vertical models of the Nuclear Island are decoupled and the floor response in horizontal and venical directions are obtained by three separate analyses. For horizontal analysis, the response spectra are generated for each floor along the two axes of the structure. In vertical analysis, the response spectra are generated for the walls. The venical response spectra included in Appendix 3.7B do not include the effects of vertical floor flexibility. In order to account for the effects of vertical floor flexibility on the vertical floor spectra, the following is performed:

• Out-of-plane floor frequencies are calculated for various slab configurations (based on dimensions and end conditions.)
• Floors having a vertical frequency less than 40 Hz are modeled by single-degree-of-freedom oscillators in the vertical SSI models.
• Vertical spectra are generated for the oscillators, thus. , ding the effects of vertical floor flexibility. The latter vertical spectra are used in subsystem design For all other Category I structures, all three orthogonal components of motion are applied simultaneously and spectra are generated in accordance with the requirements of Regulatory Guide 1.122.

The spectra are generated for appropriate critical damping for SSE. The peaks of the response spectra are broadened as described in Section 3.7.2.9. 1 l 3.7.2.6 Three Components of Earthquake Motion 3.7.2.6.1 Seismic Category I Structures, Systems, and Components Other Than NSSS For the Nuclear Island the three statistically independent orthogonal components of earthquake motion (2 horizontal and I vertical) are applied to the stmetural models as separate loading cases. The models are analyzed using either the time-history or response spectrum method of dynamic analysis as appropriate. For time-history analysis, the total response is obtained by algebraically summing the response parameters in the time domain. For response spectrum analysis, the total response of the structure due to the three input seismic motions is obtained by combining the directional responses using the square root sum of the squares (SRSS) method. For other Category I structures all three components of input excitation are applied simultaneously. O Approved Design Material . Design of SSC Page 3.714 l l

l 3 System 80+ Deslan ControlDocument 3.7.2.6.2 Nuclear Steam Supply System. - I

      ' The procedures for considering the effects of three components of earthquake motion in determining the seismic response of NSSS systems, components and supports are in accordance with Regulatory Guide 1.92. They are discussed in Section 3.7.2.1.2.3.

I 3.7.2.7 Combination of Modal Responses j 3.7.2.7.1 Seismic Category I Structures, Systems, and Components Other Than NSSS i The total seismic response of a structure to an input response spectrum loading is obtained by combining the response of each individual mode of the ' structure in accordance with the requirements of . Regulatory Guide 1.92. If the modes are not closely spaced (i.e. no two consecutive modes have frequencies which differ from each other by 10 percent or less) then the significant modes are combined using the square root sum of the squares (SRSS) of the corresponding maximum values of the response of each element of the structure. This is expressed mathematically as: , i N 4 R = ( E ' R[)* ' l k.1 i  : l Where R is the maximum response of a given element, Rx is the peak response of the element due to the  ; K* mode, and N is the number of significant modes. If some of the modes are closely spaced the response of the individual modes is combined using the Ten Percent Method from Regulatory Guide 1.92. This can be expressed as: I N R = ( E R$ + 2 E lR i R)l )% i#j k=1 , j

                                                                                                                          )

1 Where R, Rg and N are as previously defined. The second summation is performed on all I and j modes i whose frequencies are closely spaced to one another. Alternative summation methods given in Regulatory Guide 1.92, such as the Double Sum Method, are acceptable substitutes for the method described above. i i t 3.7.2.7.2 Nuclear Steam Supply System . The SRSS method is the procedure normally used to combine the modal responses when the modal i analysis response spectrum method of analysis is employed. The procedure, in accordance with Regulatory Guide 1.92, is modified in two cases: [

• In the analysis of simple systems where three or less dynamic degrees of freedom are involved, the modal responses are combined by the summation of the absolute values method;
• In the analysis of complex systems where closely spaced modal frequencies are encountered, the ,

responses of the closely spaced modes are combined by the summation of the absolute values  ! method and, in turn,- combined with the responses of the remaining significant modes by the SRSS method. Modal frequencies are considered closely spaced when their difference is less than 10 percent of the lower frequency. , rey, s.71s Annewd ouen nemmw.onon er ssc

System 80+ Design ControlDocument 3.7.2.8 Interaction of Non-Seismic Category I Structures, Systems and Components with Seismic Category I Structures, Systems and Components The interfaces between Seismic Category I and non-Seismic Category I structures, systems and components are designed for the dynamic loads and displacements produced by both the Seismic Category I and non-Seismic Category I structures, systems and components. To ensure that the failure of a non-Seismic Category I structure, system or component under the effect of a seismic event does not impair the integrity of an adjacent Seismic Category I stmeture, system or component, the following procedures are used:

• Sufficient separation bet veen non-Seismic Category I stmetures, system and components and Seismic Category I structures, system and components is maintained, or
• The non-Seismic Category I structures, systems and components are analyzed and designed to prevent their failure under SSE conditions in a manner such that the margin of safety of these structures, systems and components is equivalent to that of Seismic Category I structures, systems and components.
• The Seismic Category I structure, system or component is designed to withstand loads due to collapse of the adjacent non-Seismic Category I structure, system or component should sufficient spatial separation not be achieved.

Plant specific information will describe the process for the design of plant specific and non-Seismic Category I structures, systems and components to reduce the potential for non-Seismic Category I to Seismic Category I (II/I) interactions and propose procedures for an evaluation of the as-built plant for II/I interactions. 3.7.2.9 Effects of Parameter Variations on Floor Response Spectra To account for the expected variation in structural properties, dampings and other parameter variations, the peaks of floor response spectmm curves are broadened by il5% and smoothed in accordance with Regulatory Guide 1.122. Soil property related spectrum peaks are further broadened, where required, to conservatively account for all potential variations of soil properties within the envelope of site conditions. 3.7.2.10 Use of Constant Vertical Static Factors A constant seismic vertical load factor is not used for the seismic design of Seisnde Category I structures, systems, components and equipment. The safety-related structures, systems, and components are analyzed in the vertical direction using the methods described in Section 3.7.2.1. Based on the vertical seismic analysis, a vertical static factor is determined to design columns and shear walls. The vertical floor flexibilities are accounted for in the response spectra at each individual floor elevation of the building structures. The floor beams are designed statically for the acceleration value obtained per Reference 1. O Approved Des > Material- Design of SSC Page 3.716

                                                                                                                  )

U

1  ; System 80+ ' Deslan ControlDocument i 3.7.2.11 Methods Used To Account for Torsional Effects The mathematical models used in analysis of Seismic Category I systems, components, and piping ~ systems include sufficient mass points and corresponding dynamic degrees-of-freedom to provide a three-dimensional representation of the dynamic characteristics of the system. The distribution of mass and- . the selected location of mass points account for torsional effects of valves and other eccentric masses. ! - The structural models used for Seismic Category I systems are constructed with elements containing 6 degrees of freedom per node, incciporating torsional effects into the models. Torsional effects are also accounted for in the building models used to generate floor response spectra. An additional eccentricity of 5 % of the maximum building dimension, which results in an accidental torque, is applied to the static finite element structural model to calculate element forces due to accidental torsion. Accidental torsion is considered in both the E-W and N-S directions. 3.7.2.12 Comparison of Responses 1 With the exception of the surge line, the time-history method is used for structural analysis of the NSSS and the associated building structures. Therefore, responses obtained from the response spectrum and

!             time-history methods are not compared.

> 3.7.2.13 Methods for hiernic Analysis of Dams I If applicable for the site, analyses of safety-related dams will be performed. The methods to be employed for seismic analysis of safety related dams will be detailed in the site specific SAR. 3.7.2.14 Deterininstion of Safety-Related Structum Overturning Moments The overturning moments and base shears due to seismic forces for Category I structures are determined l using the time history method of analysis. The seismic motion is input to the structural models in three !- independent orthogonal directions. The overtuming moments for shell structures are automatically included in the analysis of this type of structure. 3.7.2.15- Analysis Procedum for Damping For modal superposition method, composite modal damping values are used for structures with components of different damping characteristics. The composite modal damping values are based on weighting the damping factors according to the mass or the stiffness of each element. For the mass j- proportional damping, formulation is as follows: d N  !

                      - E {4 j} #i {Mi} {4 j}

#j _= N

]4 fhj {hj O howowed Denk" noneeria! Dee4" of ssc pope 2.717

Sy^ tem 80+ Design ControlDocument l where: N = total number of components, l

  #j        =      composite modal damping for mode j, i

Si

            =      critical modal damping associated with component i,                                                                               q dj        =      mode shape vector,

{M i } = subregion of mass matrix associated with component i, and [M] = the mass matrix of the system. For direct integration method, viscous damping proportional to the mass and stiffness matrix is used; thus l [C] = a[M] + #[K] where [C] is the damping matrix, [K] is the stiffness matrix and [M] is the mass matrix. The values of a and # are selected such that the damping in the range of frequency of interest is approximately equal to the damping of the structure. ((Where composite modal damping is used for piping, the input damping for piping elements is in accordance with Table 3.7-1. That is, for the Safe Shutdown Earthquake, the damping is 2.0 percent of critical dampingforpiping ofdiameter :s; 12 inches and is 3.0 percent of critical dampingforpiping of diameter > 12 inches.}}2 3.7.3 Seismic Subsystem Analysis 3.7.3.1 Seismic Analysis Methods The seismic analysis of the Seismic Category I structures, subsystems, and components other than piping is performed by either the response spectrum or time history method as described in Section 3.7.2.1.1 or an equivalent static method described in Section 3.7.3.5. When analyzed using the response spectrum method, four options are available for the choice of response spectra. These are described in Appendix 3.9A, Section 1.4.3.2.1.2. Appendix 3.7D shows sample spectra for use in the three options not related to plant specific analysis. For Seismic Category I piping, each piping system is idealized as a mathematical model consisting of lumped masses connected by elastic members. The stiffness matrix for the piping subsystem is determined using the clastic properties of the pipe. This includes the effects of torsional, bending, shear, and axial deforma' ions as well as changes in stiffness due to curved members. Generally, a response spectrum analysis is performed using the envelope of all applicable spectra to account for inertia effects. The effects of rocking and torsion are implicitly included because the spectra at the support points include motions due to rocking and torsion. The total seismic response of the piping is then calculated 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction O Section 3.5, Approved Desigro Material Design of SSC (2/95) Page 3.7-18

System 80+ Design ControlDocument O by absolute summing the results of the response spectrum analysis and a static analysis which accounts (V for the relative displacement effects between support locations. Since the displacement effects are self-limiting, it is justified to place them in the secondary stress category. As an alternative to the modal response method, a time history method of analysis may be used. This method is also used for other types of dynamic analyses such as LOCA and hydraulic transients. Either a direct integration method or a modal superposition method is used to solve the equations of motion. 3.7.3.2 Determination of Number of Earthquake Cycles The procedure used to account for the fatigue effect of cyclic motion associated with seismic excitation recognizes that the actual motion experienced during a seismic event consists of a single maximum or peak motion, and some number of cycles of lesser magnitude. The total or cumulative usage factor can also be specified in terms of a finite number of cycles of the maximum or peak motion. Based on this consideration, f(Seismic Category I subsystems, components, and equipment are designedfor a total of two SSE events with 10 maximum stress cycles per event (20 full cycles of the maximum SSE stress range). Alternatively, an equivalent number offractional vibratory cycles to that of 20 full SSE vibratory cycles may be used (but with an amplitude not less than one-third (1B) of the maximum SSE amplitude) when derived in accordance with Appendix D ofIEEE Standard 344-1987.}}2 3.7.3.3 Procedure Used for Modeling The modeling techniques incorporate either a single or multi-degree of freedom subsystem consisting of g discrete masses connected by spring elements. The associated damping coefficients are consistent with Table 3.7-1. The degree of complexity of each model is sufficient to accurately evaluate the dynamic ('y behavior of the component. For additional details on pipe modeling, see the section below. Valves (i.e., with natural frequencies greater than the frequency corresponding to the zero period acceleration (ZPA)) are included in the piping system model as lumped masses on rigid extended structures. If it is shown by test or analysis that a valve has a frequency less than a frequency corresponding to the ZPA, then a multi-mass, dynamic model of the valve, including the appropriate stiffnesses, is developed for use in the piping system model. The continuous piping system is modeled as an assemblage of beams. The mass of each beam is lumped equally at its associated end nodes, which are connected by massless elastic members, representing the physical properties of each segment. The pipe lengths between these mass points shall not be greater than the length that would produce a natural frequency equal to the cutoff frequency (ZPA) when calcuhted based on a simply supported beam. All concentrated weights on the piping system such as main vahes, relief valves, pumps, and motors are modeled as lumped masses. The torsional effects of the valve operators and other equipment with offset center of gravity with respect to centerline of the pipe are included in the analytical model. m i / G 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approved Design Material Design of SSC Page 3.719

System 80+ Decign controlDocument 3.7.3.4 Basis for Selection of Frequencies The basis for acceptability of the seismic design of equipment and subsystems is that the stresses and deformations produced by vibratory motion of the postulated seismic events, in combination with other coincident loadings, be within the established limits. The seismic design is accomplished in a manner to account for the seismic response of structures, subsystems and components in their design as wei! as to maintain the resonant frequencies outside the range that is significantly excited by the forcing frequencies. The stiffness of the restraint and supports system is designed to maintain the fundamental frequencies of equipment and subsystems sufficiently removed from the resonant range and, thereby, maintain the seismic response within the established limits. If the natural frequencies of the equipment and supporting structures are in the same range where resonance can occur, the resonance is accounted for in the analysis. 3.7.3.5 Use of Equivalent Static Load Method of Analysis The equivalent static load method involves the multiplication of the total weight of the equipment or component member by the specific seismic acceleration coefficient. The magnitude of the seismic acceleration coefficient is established on the basis of the expected dynamic response characteristics of the component. Components that can be adequately characterized as a single degree of freedom system are considered to have a modal panicipation factor of c,ne. Seismic acceleration coefficients for multi-degree of freedom systems which may be in the resonance region of the amplified response spectra curves are increased by 50% to account conservatively for the increased modal participation. If the equipment natural frequency is above the frequency corresponding to the zero period acceleration (ZPA), the seismic acceleration coefficient is equal to 1.0 times the ZPA. 3.7.3.6 Three Components of Earthquake Motion Seismic responses resulting from analysis of subsystems due to three components of earthquake motions are combined in the same manner as the seismic response resulting from the andysis of building structures (Section 3.7.2.6). The following description is applicable to safety-related components and systems: The system and equipment response is determined using three earthquake components, two horizontal and one vertical. Floor response spectra are generated for two perpendicular horizontal directions, (i.e., N-S, E-W) and the venical direction. Piping and equipment analysis is performed with these response spectra components applied in the N-S, E-W, and vertical directions. The damping values used in the analysis of equipment are those given in Table 3.7-1. 3.7.3.7 Combination of Modal Responses When a response spectrum method of analysis is used to analyze a subsystem, the maximum response (accelerations, shears, and moments) in each mode is calculated independent of time. If the frequencies of the modes are well separated, the SRSS method of mode combination gives acceptable results; however, where the structural periods are not well separated, the modes are combined in accordance with Regulatory Guide 1.92. This is automatically performed by the piping analysis computer program i SUPERPIPE. Approved Design Material Design of SSC Page 3.7-20

8 f ^ System 80+ Deslan controlDocument The effects of seismic response of supports and equipment are not directly included in the seismic analysis of piping initially as equipment and suppons are normally designed and analyzed subsequent to the piping analysis. Analytical Procedures for Piping 3.7.3.8 All Seismic Category I piping is analyzed for seismic effects as described in Section 3.7.3.1. liffinelastic methods are used in any System 80+ piping analysis, the details of the inelastic method and its acceptance criteria, as well as the scope and enent ofits application, will be provided nith site-speapc information.}}2 l 3.7.3.8.1 Dynamic Analysis Each piping system is idealized as a mathematical model consisting of lumped masses connected by clastic massless members. Appendages having significant dynamic effects on the piping system, such as motors ' j attached to motor-operated valves, are included in the model. Using the elastic properties of the pipe, J "the stiffness matrix for the piping system is determined. This includes the effects of torsional, bending, shear, and axial deformations, as well as the local flexibilities of piping curved members. Next, the frequencies and mode shapes for all the significant modes of vibrations are calculated. After the frequency is determined for each mode, the corresponding horizontal and vertical spectral accelerations with appropriate damping are read from the appropriate response spectrum curves. For each mode, the inertia response forces, moments, displacements and accelerations are determined due to excitation in the three directions simuhaneously (two horizontal and one vertical). Finally, the stresses are determined by taking the SRSS of the individual components. The relative displacement effects between piping l supports are discussed in Section 3.7.3.1. 3.7.3.8.2 Allowable Stasses Allowable stresses in the piping caused by an earthquake are in accordance with Section III of the ASME Code. Allowable stresses in the earthquake restraint components, such as snubbers, are in accordance , with any additional stress limits that may have been established by ASME Code, Section III at the time ! the restraint components were purchased. i 3.7.3.9 Multiple Supposted Equipment Components with Distinct Inputs When the equipment or component is supported at points with different elevations, either the envelope ' of these elevation response spectra or multiple support excitation is used for the seismic qualification of the equipment. For multiple support excitation, time history analysis methods are used. ((!findependent support motion (ISM) response spectrum analysis methods are used in any System 80+ 1 piping design, a detailed description of methodology, including a sample analysis and a description of  ; the computer code and its venpcation, will be provided with site-speapc information.}}2 , I 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction  ! Section 3.5. ,

        ?;.        Deeen heenenief- Deekn of SSC                                                              Page 3.7-21 1
                                                                                          -s,

Srtem 80+ Design controlDocument 3.7.3.10 Use of Constant Vertical Load Fadors In general, Seismic Category I subsystems are analyzed in the vertical direction using the methods specified in Section 3.7.3.1. No vertical static factors are used for subsystems. 3.7.3.11 Torsional Effects of Eccentric Masses Piping systems are modeled to include projecting masses such as valve motor operators. The actual stiffness of the connecting member is not expected to influence the system appreciably. However, an approximation is made by assuming a member stiffness equal to that of the piping in which the valve is installed. Torsional effects of eccentric masses are also considered in the analysis of Seismic Category I subsystems other than piping. 3.7.3.12 Piping Outside Containment Structure 3.7.3.12.1 Buried Piping Class 2 and 3 buried piping systems are designed according to the seismic analysis acceptance criteria for Seismic Category I buried piping systems. These criteria accept either dynamic analyses or equivalent static load methods with the consideration of seismic effects which are induced primarily by seismic wave passage and by differential movements between building attachment points and the ground surrounding the buried pipe. Seismic effects in buried piping are self-limiting (displacement-induced strains and associated stresses) O rather than being in equilibrium with an external load. In accordance with the ASME B&PV Code stress classification rules, the elastically calculated seismic stresses in buried piping, which are the products of strains and Young's moduli of the piping rnaterials, are secondary stresses rather than primary stresses in nature. The design and stress acceptance criteria as provided below have been modified from the Code Sub-articles NC/ND-3600. Ncmenclatural symbols are defined in the Code unless noted otherwise. 3.7.3.12.1.1 General Requirements General design requirements for ABB-CE System 80+ buried piping systems are as follows:

• Areas of direct fault displacement and unstable soil conditions, such as liquefaction, are avoided for buried pipe installation.
• llConfonnance to allowable structural and piping stresses after the line Annex is assured by the use of either expansionjoints orflexible seals.}}fen
• Minimum pipe wall thickness is per Code Equation 3 of NC/ND-3641.1 as follows:

                                                                                                                                                                   )

l 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. ) Apqwoved Design Material Design of SSC Page 3.7-22

l System 80+ Design ControlDocument

                                . P D,                                                                                                ,

t'" = +A e 2 (S + Py) ! e Allowable external pressure is per NB-3133 as follows: l 4B p'. , i 3 (D, j T) 1 1 e Since the failure of long and flexible buried piping in most cases is induced by excessive , deflection rather than by high stress level, ((the total radial deformation of buried piping is limited to within 2% ofpipe radius asfollows: D 6, s 0.02 (j)

                 ' where: 6, is the radial displacement of the pipe wall.}}2 3.7.3.12.1.2        Weight EKects (Sustained Id)

Evaluation of weight effects is performed with consideration of the following loads: e Dead loads, such as the self weight of pipe and the soil overburden pressure. e Live loads, such as weight of water content, highway / railroad loading, and construction , equipment loading. The effects of pressure, weight, and other sustained mechanical loads must meet the requirements of Code Equation 8 of NC/ND-3652 as follows: Sg- Bi P D, + B2 ZM 3 s 1.5S 3 2t, s 3.7.3.12.1.3 Seismic EKects 1 Buried piping is seismically designed to sustain soil movements during earthquake ground motions. The structural integrity of the piping is evaluated by accounting for two primary effects of earthquake ground

        - motions:

o' ~ Strains and associated stresses induced in a long buried pipe by the free-field vibration resulting from motions of the surrounding soil mass (seismic wave passage). The maximum strains associated with the free-field vibration and the soil are computed based on the guidelines of References 10 and 11. Friction between the pipe and the surrounding soil may be considered using conservative estimates of the associated frictional forces. ~ 2 i NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction

                   ' Section 3.5.

n M poedre asees, der. ow(p en pape 3.7 23 4

Systern 80+ Design Control Document

• Seismically induced differential movements of structures which the pipe enters or connects.

Equivalent static analysis is performed using the principles of beams on clastic foundations. The differential movements at the entry points are conservatively assumed as out-of-phase. Buried piping experiences both axial and bending strains and associated stresses due to seismic wave effects, seismic differential movements, as well as thermal expansion / contraction. These high axial strains and associated stresses, normally insignificant in the aboveground piping systems, are caused mostly by seismic waves in conjunction with the soil friction. Therefore in addition to bending stresses, axial stresses are included in the Code stress check. Seismic loads on buried piping are considered secondary loads, since they are generated by soil strains transmitted to the pipe and are therefore considered to be self-limiting. ((7he effects of seismic loading, for which Level D (faulted with SSE loading) service limits are designed, must meet the requirements of thefollowing equations in lieu of Code Equation 9 of NC/ND-3655: Level D (faulted) Sgt - lAISSg + _ihic+Fm + A s; 3S, _F,, j ))2 Z Z Ap p where: MssE is the range of resultant moments on cross section of pipe due to earthquake loading, including the effects of anchor displacements due to earthquake. The M/Z term can be replaced by the product of strain and Young's modulus of the piping material. Me is the range of resultant moments on cross section of pipe due to thermal expansion. Fa(ssE) is the range of axial forces on straight runs of buried pipe due to earthquake loading, including the effects of anchor displacements due to earthquake. The F,/Ap term can be replaced by the product of strain and Young's modulus of the piping material. Axial stresses due to seismic soil movements are considered significant and are therefore included for the buried pipe stress check. F, is the range of axial forces on straight runs of pipe due to thermal expansion and contraction frictional forces. A pis the cross sectional metal area of the pipe. 3.7.3.12.1.4 Thermal Expansion and Contraction Effects ([The efects of thermal expansion / contraction must meet the requirements of modified Code Equation 10 of NC/ND-3653.2(a) or modified Code Equation 11 of NCIND-3653.2(c).))2 Modifled Code Equation 11 considers the combined effects of pressure, weight, other sustained mechanical loads, and thermal expansion / contraction. Both Equations 10 and 11 are modified to include axial stresses due to thermal expansion / contraction frictional forces, which are considered to be significant for buried pipes. 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. . Approved Design Material Design of SSC Page 3.7 24

System '80 +' Oesian CorrtrolDocument b if ((Modilled Code Equation 10:  ; -Q  !

                         'Sg_= iM' + F ~ .s S, P

Meditied Code Equadon 11: M# - ', Sg ='PD" + 0.751 1 4t, . 2 (M,) A i ( Z ) 4 .F" as (SpS ) A. p where: F, is the range of axialforces on straight runs of buried pipe caused by thefrictionalforces ' between the soil and the pipe during thermal e.xpansion and contraction.}}2 [

                     ' 3.7.3.12.1.5 ' Non-Repeating Differential Settlan=d Effects l                       The effects of existing permanent, unidirectional and non-repeating differential settlement, including building settlement and the effects due to loss of support, must meet the requirements of Code Equation 10a of NC/ND-3653.2(b):

t . iM D

       ,                         g 33 c 4

t z r

      \

3.7.3.12.2 Above Ground Piping

Seismic design criteria and methods of accounting for the effects of differential movement of buildings on piping and penetrations are described in Sections 3.7.2.1.2 and 3.7.2.7.

i 3.7.3.13 Interaction of Other Piping with Category I Piping 2 The protection of Category I piping from possible tidverse effects of other piping during an earthquake L 'is accomplished by several methods. Specifically, these methods are: e f(Non-Category Ipiping systems are designed to be isolatedfrom any Category Ipiping system by either a constraint or barrier, or are remotely located with regard to the seismic Category I - l+ f piping system. Ifit is notfeasible orpractical to isolate the Category Ipiping system, adjacent non-Category ipiping is analyzed according to the same seismic criteria as applicable to the Category I piping system. For Non-Category Ipiping systems attached to Category Ipiping systems, the dynamic efects of the Non-Category I piping is simulated in the modeling of the Category I piping. The attached Non Category I piping, up to thefirst anchorbeyondthe _ f Q'.. Lj H 2 . NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction

                 $;               Section 3.5.                                                                                                                    ,
 !4                     Appmes Des (p eressenst. Des 4p er **c                                                                            psy. 3.7 2s
           ,                  J    s.             e,,      .5.,     -.                                  ,  ,,,[...  . . , ,            .,  . , d r,
                                                                                                                                                     . .-       r

System 80+ Design ControlDocument interface, is also designed in such a manner that during an earthquake of SSE intensity it will not cause afailure of the Category Ipiping.}}2 e llAll Category I boundary valves are designed to meet seismic criteria.}}2 A valve always serves as a pressure boundary and constitutes the seismic to non-seismic boundary. If failtre in the non-seismic ponion of the system could cause loss of function of the safety system, then an appropriate automatic or remote manual operator wu tid be used if the valve is open during normal reactor operation. e llThe pressure boundary valve is protected by restraining or anchoring the non-seismic ponion of the system))2 as discussed above. 3.7.3.14 Seismic Analysis of Reactor Internals, Core and CEDMs 3.7.3.14.1 Reactor Intemals and Core The seismic analyses of the reactor internals and core consist of separate analyses performed in the horizontal and venical directions. In the horizontal direction, because the relative displacements between the core and core shroud and between the core support barrel and pressure vessel snubbers are sufficiently large to close the gaps that exist between these components, a nonlinear horizontal time history analysis is performed. The horizontal nonlinear analysis is divided into two parts. In the first pan, the internals and core are analyzed to obtain the internals' response and the proper dynamic input for the reactor core model. In the second part, the core plate motion from the first part is applied to a more detailed nonlinear model of the reactor core. The input excitation to the internals model is the response time-history of the reactor vessel at the internals support determined from the RCS analysis. Coupling effects between the int,ernals and reactor vessel are accounted for by including a simplified representation of the internals with the RCS model. This is discussed in Section 3.7.2. In the vertical direction, both linear response spectrum modal analysis and non-linear time history analysis methods are used. For linear analysis cases where the response of the core is sufficiently large to cause it to lift off the core plate, non-linear methods are used. The response accelerations of the reactor vessel flange provide the input for both the linear and non-linear internals vertical analyses. In these analyses, two horizontal components and the vertical component of the seismic excitation are considered and the maximum responses for the three components are combined by the method of square root of the sum of the squares. Closely spaced modes are considered in accordance with Regulatory Guide 1.92. 3.7.3.14.1.1 Mathematical Models Equivalent multi-mass mathematical models are developed to represent the reactor internals and core. The mathematical models of the internals are constructed in terms of lumped masses and elastic-beam 1 elements. At appropriate locations within the internals and core, points (nodes) are chosen to lump the 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. j Approved Design Material . Destyn of SSC Page 3.7-26 l l

  .-     - _ ~ ~ . - - - - .                         ..   --           ---_- -._ -.                              --            . . . - - . .--

1 System 80+ Desian ContmlDocument l weights of the structure. A sketch of the internals and core showing the relative node locations for the , horizontal model is presented in Figure 3.7-26. The criterion for choosing the number and location of  ; mass concentration is to provide for accurate representation of the dynamically significant modes of i vibration of each of the internals components. Between the nodes properties are calculated for moments  ; of inertia, cross-section areas, effective shear areas, and lengths. Separate horizontal and vertical models l l of the internals and core are formulated to more efficiently account for structural differences in these directions, in the horizontal nonlinear lumped mass representation of the internals and core, shown in i Figure 3.7 27, gap and spring elements are used to represent contact between the fuel and core shroud. Lumped-mass nodes in the core are positioned to coincide with fuel-spacer grid locations. To simulate  : the nonlinear motion of the fuel, nonlinear spring couplings are used to connect corresponding nodes to i the fuel assemblies and core shroud. Incorporated into these nonlinear springs is the spacer grid impact  : stiffness derived from test results. The core is modeled by subdividing it into fuel assembly groupings and choosing stiffness values to adequately characterize its beam response and contacting under dynamic loading. - The horizontal nonlinear reactor core model consisting of one row of 17 individual fuel assemblies is depicted in Figure 3.7-28. In this model each fuel assembly is represented with mass points located at spacer grid locations. To simulate the gaps in the core, nonlinear spring couplings are used to connect  ; i corresponding nodes on adjacent fuel assemblies and core shroud. The impact stiffness and impact damping (coefficient of restitution) parameters for the gap elements are i derived from the impact tests which are described in Section 4.2. The spacer grid impact representation  ; used for the analysis is capable of representing two types of fuel assembly impact situations. In the first i

type, only one side of the spacer grid is loaded. This type of impact occurs when the peripheral fuel

's assembly hits the core shroud, or when two fuel assemblies strike one another. The second type of

. impact loading occurs typically when the fuel assemblies pile up on one side of the core. In this case, o the spacer grids are subjected to a through-grid compressive loading.

The fuel assemblies in the coupled core / internals model and the detailed core model are modeled with beam elements to represent the horizontal stiffness between mass points and rotational springs at each end ) , to simulate the end fixity existing at the top and bottom of the core. The valve used for fuel horizontal l stiffness and end fixity is based upon a parametric study in which analytic predictions are correlated with l . fuel assembly static and dynamic test data. Fuel assembly structural damping as a function of vibrational , amplitude was derived from fuel assembly forced vibration and pluck tests defined in Section 4.2. The  : i damping values used in the seismic analysis of the reactor internals are in accordance with the values in Table 3.7-1. Figure 3.7-29 shows the idealized linear vertical model. The vertical nonlinear model is

!                          shown in Figure 3.7-30.

Additional salient details of the internals and core models are discussed in the following paragraphs.

!

• Hy&v4 M Effects It has been shown both analytically and experimentally (Reference 2) that immersion of a body in a dense-fluid medium lowers its natural frequency and significantly alters its vibratory response as compared to that in air. The effect is more pronounced where the confining boundaries of the 4 fluid are in close proximity to the vibrating body as in the case for the reactor internals. The method of accounting for the effects of a surrounding fluid on a vibrating system has been to ascribe the system additional or " hydrodynamic mass." The hydrodynamic mass of an immersed V system is a function of the dimensions of the real mass and the space between the real mass and

confining boundary.

1 Nemme onon annuser.outen er ssc rene 2.127

Syntem 80+ Design ControlDocument Ilydrodynamic mass effects for moving cylinders in a water annulus are discussed in References 2 and 3. The results of these references are applied to the internals' structures to obtain the total (structural plus hydrodynamic) mass matrix that is then used in the evaluat]on of the natural frequencies and mode shapes.

• Core Support Barrel The core support barrel is modeled as a beam with shear deformation. It has been shown that the use of beam theory for cylindrical shells gives sufficiently accurate results when shear deformation is included (References 4 and 5).
• Fuel Assemblies The fuel assemblies are modeled as uniform beams with rotational springs at each end to represent the proper end condition. The member properties for the beam elements representing the fuel assemblies are derived from the results of experimental tests of fuel-assembly load deflection characteristics and fundamental natural frequency.
• Support-Barrel Flanges To obtain accurate lateral and venical stiffness of the upper and lower core-support barrel flanges and the upper guide structure support barrel upper flange, finite-element analyses of these regions are performed. As shown in Figure 3.7-31, these areas are modeled with quadrilateral and triangular ring elements. Unit deflections and rotations are applied in the lateral and axial directions, and the resulting reaction forces are calculated. These results are then used to derive the equivalent member properties for the flanges.
• Upper Guide Structure For the horizontal model, the upper guide structure including CEA shrouds, connecting plates and tie rods are modeled as cantilever beams. A separate member is modeled to account for the connection between the tie rods and the upper guide structure support plate.
• Lower Support Structure To obtain vertical stiffness for the lower support structure grid beams and cylinder, a finite element analysis is performed. Displacements due to vertical (out-of-plane) loads applied at the beam junctions are calculated through the use of the computer code ANSYS, a description of which is given in Section 3.9.1.2.2.2. Average stiffness values based on these results yield an equivalent member cross-section area for the vertical model.

3.7.3.14.1.2 Analytical Techniques

• Natural Frequencies and Mode Shapes The mass- and beam-element properties of the models are utilized in the computer code ANSYS to obtain the natural frequencies and mode shapes. The program utilizes the stiffness-matrix method of structural analysis. The natural frequencies and mode shapes are extracted from the system of equations.

Approved Design Materiel Design of SSC (11/96) Page 3.7-28

Sv~ tem 80+ oesign contrat Document i

   \              @ -W,2M) 4, = 0 where:

K = modal stiffness matri7. 1 M = modal mass matrix W, = natural circular frequency for the n* mode l 4, = normal mode shape matrix for the n* mode The mass matrix, M, includes the hydrodynamic and structural masses.

        *     ' Response Calculation Methods Response Spectra Method The response spectrum analysis is performed using the modal extraction data and the following relationships for each mode:

1. Nodal Accelerations t

N Ai , = P. A, $3 i where: R i , = absolute acceleration at node "i" for node "n" r, = modal participation factor A, = modal acceleration from response spectrum 6, i = mode shape factor at node "i" for node "n"

2. Nodal Displacement T

R Yi ,=._3 W,2 i where: Y io;= displacement at node "i" for mode "n" relative to base

                          .W. = natural circular frequency for n* mode 4prowd W AfshrW Dwen of SSC                                                                      Pape 3.7-29

l i Sy~ tem 80 + Design control Document Member Forces and Moments F, = ( nI A") -F, W ,, where: F, = actual member force for mode "n" 53, = modal member force for mode "n" The effect of the fluid environment is accounted for by defining the modal participation as follows: M E W,3 d,i p,n y y Jl E E 4in W n 4,, 61 j=l where: W,i = structural weight of node "i" Wu = structural + hydrodynamic weight terms M = number of masses The SRSS method is normally used to combine the modal responses. Where modal frequencies are closely spaced, the responses of these modes are combined by the sum of their absolute values. The modal damping factors are obtained by the method of " mass mode weighting," which gives: B" - EM i 4, where: B, = modal damping factor Mi = structural mass of mass node "i" 4,, = absolute value of the mode shape at mass node "i" B, = damping associated with mass point "i" - Approved Denips NetwW e Design of $3C 111/96) Page 3.7 30

t

        . System 80+                                                                        Deslan Control Document i

O - ""-^->>* The nonlinear seismic response and impact forces for the internals and fuel are determined using the CESHOCK computer program (see Section 3.9.1.2.2.4). The computer program provides ' the numerical solution to transient dynamic problems by step-by-step integration of the differential equations of motion. The input excitation for the model is the time-history accelogram of the reactor vessel. Input to the CESHOCK computer program consists of initial conditions, nodal lumped masses, linear-spring coefficients, mass moments of inertia, nonlinear spring curves, and the acceleration  ; ! time-histories. The output from the CESHOCK computer program consists of displacements, translational and angular accelerations, impact forces, shears, and moments. 3.7.3.14.1.3 Analysis Procedures for Damping

          . The procedures used to account for damping in the analysis of the reactor internals and core'are given in Section 3.7.3.14.1.2. Uniform modal damping factors are used in the analysis of other NSSS vendor
          ! supplied seismic subsystems.

3.7.3.14.1.4 Results The nonlinear response loads for the internals, including impacting, if any exist, are determined for the vertical and horizontal directions. Loads for the fuel are determined in a separate reactor core nonlinear analysis. The results are determined for the safe shutdown earthquake (SSE). 4 O 3.7.3.14.2 Control Element Drive Mechanisms l I The pressure-retaining components of the Control Element Drive Mechanisms (CEDM) are designed to the appropriate stress criteria of ASME Code Section III for all loadings specified. The structural integrity of the CEDM when subjected to seismic loadings is verified by combination of test and analysis. Methods of modal dynamic analysis employing response rpectrum techniques or time history analysis are supported with experimentally obtained information. l 3.7.3.14.2.1 Input Excitation Data l 1l For the dynamic analyses, a response spectra or time history definition of the excitation at the base of j ~ the CEDM nozzle is obtained from the seismic analysis of the RCS. The e:tcitation is applied i simultaneously in three mutually perpendicular directions (two horizontal and one vertical). I 3.7.3.14.2.2 Analysis

;           A dynamic analysis of the mathematical structural model is performed utilizing one or more of the computer programs discussed in Sectior: 3.9.1.2.

3.7.3.14.2.3 Functional Test

A functional test utilizing a minimum drop weight is performed to verify that drop characteristics meet the input design requirements. Resuhs from this test are compared to the calculated CEDM deflections O. under seismic loading for the individual site. Verification of the proper function is thus a,stablished based on both analytical and test results.

~ ' 4,uenc outp. neenerw. ouw or ssa rare s.7.sr  ; I

Svntem 80+ Design Control Document 3.7.3.15 Analysis Procedures for Damping 3.7.3.15.1 Proportional Damping For subsystem dynamic time history analyses using direct integration, damping is input using a proportional damping Matrix [C] that is given by: C = aM + #K where M is the subsys:em mass matrix and K is the subsystem stiffness matrix. The damping ratio, (, at any frequency, f,i is given by: ( = a/4rf + Srf3 3 Figure 3.7-33 shows a typical plot of damping ratio vs. frequency. The selection of a and # is dependent on the subsystem damping ratio, as selected from Table 3.7-1 and the frequency characteristics of the subsystem and the input forcing functions. The frequencies fi and f 2, Figure 3.7-33, are selected so that all subsystem modes with significant participation lie between f iand f .2 The a and S coefficients are then calculated so that at fi and f 2the damping ratio is equal to the damping ratio selected from Table 3.7-1. The damping ratio for all frequencies betweeni f and 2f Will be less than the selected ratio. Since all significant modes lie between if and f2, the damping ratios for these modes will also be less than the selected ratio, and are therefore conservative. 3.7.3.15.2 Subsystems Other Than NSSS The analysis procedure used to account for the damping in non-NSSS Subsystems complies with Section 3.7.2.15. 3.7.3.15.3 Nuclear Steam Supply System The procedures used to account for damping in the analysis of the reactor internals and core are given in Section 3.7.3.14. The analytical method for evaluating the faulted condition uses a linear elastic model as described in Section 3.7.3. The ASME Section III allowable stress limits are met for faulted loads, including the safe shutdown earthquake and system transient loads described in Section 3.7.1. 3.7.4 Seismic Instrumentation 3.7.4.1 Location and Description of Seismic Instrumentation State-of-the-art solid-state digital instrumentation that will enable the prompt processing of data at the plant site should be used. A tri-axial time-history accelerometer should be provided at each of the following locations:

• One at the finished grade in the free field.
• Three ia the Nuclear Island: one on the common basemat, one on the shield building wall, and one at the co.'tainment operating floor.

4prowd Design Material . Design of SSC 111/96) Page 3.7-32

System 80+ 0: sign control Document O

• Two in the Component Cooling Water Heat Exchanger Building; one on the foundation mat and one at elevation 91.75 feet.

3.7.4.2 Seismic Instrumentation Operability and Characteristics The seismic instrumentation should operate during all modes of plant operation, including periods of plant shutdown. The maintenance and repair procedures should provide for keeping the maximum number of instruments in service during plant operation and shutdown. The design should include provisions for in-service testing. The instruments should be capable of periodic channel checks during normal plant operation and the capability for in-place functional testing. The instrumentation on the foundation and at elevations within the same building or structure should be interconnected for common starting and common timing, and the instrumentation should contain provisions for an external remote alarm to indicate actuation. The pre-event memory of the instrumentaGon should be sufficient to record the onset of the earthquake. It should operate continuously during the period in which the earthquake exceeds the seismic trigger threshold and for a minimum of 5 seconds beyond the last trigger level signal. The instrument should be capable of a minimum of 25 minutes of continuous recording. The acceleration sensors should have a dynamic range of 1000:1 zero to peak (i.e., 0.00lg to 1.0g) and the frequency range should be 0.20 Hz to 50 Hz. The seismic instrumentation system is triggered by the accelerometer signals. The actuating level should be adjustable for a minimum of 0.005g to 0.02g. The trigger is actuated whenever the acceleration exceeds 0.0!g. The initial setpoint may be changed (but shall not exceed 0.02g) once sufficient plant operating data have been obtained which indicate that a different setpoint would provide better system operation. The instrumentation should be capable of on-line digital recording of all components of accelerometer signals. The digitized rate of the recorder should be at least 200 samples per second, the frequency band width should be at least from 0.20 Hz to 50 Hz, and the dynamic range should be 1000:1. The instrumentation should be capable of using the recorded signal to calculate the standardized cumulative absolute velocity (CAV) and the 5% of critically damped response spectrum. The instmments should be capable of having routine channel checks, functional tests, and calibrations. The CAV shutdown threshold of 0.16g-seconds should be calibrated with the October,1987 Whittier, California earthquake record or an equivalent calibration record provided for this purpose by the manufacturer of the instrumentation. In the event that an earthquake is recorded at the plant site, all calibrations including that of the CAV will be performed to demonstrate that the system was functioning properly at the time of the earthquake. l 3.7.4.3 Control Room Operator Notification Activation of the seismic trigger causes an audible and visual annunciation in the control room to alert the plant operator that an earthquake has occurred. 3.7.4.4 Comparison of Measured and Predicted Response Within four hours after the earthquake, the 5% damped response spectrum and the CAV for each of the three components of recorded data in the free field will be obtained and evaluated to determine if the shutdown criteria defined in EPRI reports NP-5930 (Reference 13) and NP-6695 (Reference 21) have been exceeded. The plant will be shut down when the recorded motion in the free field in any of the O three directions (two horizontal and one vertical) exceeds both the response spectrum limit and the CAV limit as follows: Approved Design Material- Design of SSC Page 3.7 33

System 80 % Design controlDocument

1. The response spectrum limit is exceeded if:

• At frequencies between 2 and 10 Hz, the recorded response spectral accelerations of 5%

damping exceed 1/3 of the corresponding SSE values or 0.2g, whichever is greater; or

• At frequencies between 1 and 2 Hz, the recorded response spectral velocity of 5%

damping exceed 1/3 of the corresponding SSE values or 6 in/sec, whichever is greater.

2. The CAV limit is exceeded if the CAV value calculated according to the procedures in EPRI report TR-100082 (Reference 22) is greater than 0.16g-sec.

3.7.4.5 In-service Surveillance Each of the seismic instruments will be demonstrated operable by the performance of the channel check, channel calibration, and channel functional test operations. The channel checks will be performed every two weeks for the first three months of service after startup. After the initial three-month period and three consecutive successful checks, the channel checks will be performed on a monthly basis. The channel calibration will be performed during each refueling. The channel functional test will be performed every six months. 3.7.4.6 Pre-Earthquake Planning and Post-Earthquake Actions Site specific information submitted to the NRC as part of the application will include procedures for pre-earthquake planning and post-earthquake actions. The procedures shall implement the seismic instrumentation program specified in Section 3.7.4 and follow the guidelines recommended in EPRI report NP-6695, with the following exceptions:

1. Section 3.1. Short-Term Actions e item 3. " Evaluation of Ground Motion Records" - There is a time limitation of four hours within which the licensee shall determine if the shutdown criterion has been exceeded. After an earthquake has been recorded at the site, the licensee shall provide a response spectrum calibration record and CAV calibration record to demonstrate that the system was functioning properly.

• Item 4. " Decision on Shutdown" - Exceedance of the EPRI criterion as amended in Subsection 3.7.4.4 of the SSAR or observed evidence of significant damage as defined by EPRI NP-6695 shall constitute a conditicq for mandatory shutdown unless conditions prevent the licensee from accomplishing an orderly shutdown without jeopardizing the health and safety of the public, e Add item 7. "Doctunentation" - The licensee shall record the chronology of events and control room problems while the earthquake evaluation is in progress.

2. Section 4.3.1. Immediate Operator Actions. Add to the checks listed in this section a prompt check of the neutron flux monitoring instruments for stability of the reactor.

3. Section 4.3.4. Pre-Shutdown Inspection. Exceeding the EPRI criterion or evidence of significant damage should constitute a condition for mandatory plant shutdown.

Page 3.7 34 4 proved Desip Afsterial* Desty of SSC

Sy: tem 80+ uesign controlDocument ( 4. Section 4.3.4.1. Safe Shutdown Equipment. In addition to the safe shutdown systems on the list, C containment integrity must be maintained following an eanhquake. Since the containment isolation valves may have malfunctioned during the earthquake, inspection of the containment isolation system is necessary to assure continued containment integrity. 3.7.5 Seismic Category I Tanks Seismic Category I atmospheric storage tanks are generally large, flat bottomed single shell, free standing cylindrical tanks anchored to reinforced concrete pads or directly on building stmeture. Basic tank dimensions range up to approximately 70 feet in diameter and a 40 foot height of contained unpressurized, non-viscous fluid (on the order of 1 million gallon maximum capacity). The tanks are typically fabricated of steel (carbon or stainless) or aluminum, and are of welded construction with a minimum wall thickness of 3/16 inches. Design procedures address the issues described in NUREG/CR-1161 (Reference 15), pages '28-30, based primarily on the methods of Haroun and Housntr (Reference 16). Due to the symmetry of these vertical tanks, only one of the two horizontal canhquake components (the larger, if they are not equal in magnitude) is combined by the SRSS method with the vertical earthquake component. The qualification of Seismic Category I storage tanks addresses four fundamental issues beyond basic presstue boundary sizing, meeting the design requirements and acceptance criteria for ASME III, Class 2/3 components. 3.7.5.1 Tank Wall Stability The assesstrent of dynamic loading on storage tanks verifies stability of the tank wall against buckling behavior, accounting for hydrodynamic loads (impulsive and convective) and shell flexibility. In the generation of dynamic loads, tanks are evaluated as filled, with consideration of convective (sloshing),impuls n (fluid-shellinteraction) and rigid mob if behavior. For the convective mode, fluid damping is taken as 1/2 % of critical damping in accordanse w.th Reference 15. For the impulsive mode, structural (tank wall) damping is taken as 4% for the SSE, in accordance with Reference 17. The effective mass, its location and natural frequency for each mode of behavior are obtained from the eqwions and graphs of Reference 16. Using the plant spectra applicable at the base of the tank (either floor response spectra or ground response spectra), spectral accelerations obtained for each mode at the appropriate damping and frequency are applied to the appropriate effective mass. Zero period acceleration (ZPA) is used for the rigid mode. Development of the input spectra account for SSI considerations on the tank foundation. The overturning moment at the base of the tank due to the combined response of the three response i modes determines the compressive loads in the tank wall. Elephant foot buckling is evaluated using the recommendations of Reference 19; the limit for conventional diamond-shaped buckling under axial load and internal pressure is determined using Reference 20. To account for the varying wall thickness along the height of the tank, an effective thickness is used based on the mean of the minimum and average wall thicknesses, in accordance with the recommendations of References 16 and 18. Sloshing heights of the (_ 5 fluid free surface are verified to be less than the available freeboard to preclude significant loadings on the tank head or roof stmeture. Agnprend Design Material Design of SSC Page 3. 7-55

System CO + Design ControlDocument ".E.'. . Anchorage Provisions Structural adequacy of the anchorage provisions for the tank, e.g. anchor bolts and embedments, is developed assuming that the overturning moment on the tank is resisted only by compression in the shell and tension in the anchor bolts. The overturning moment at the base of the tank is computed as the sum of the flexible and rigid mode responses, each of which is the product of the applicable mass, height, and spectral acceleration. The shear load at the base of the tank is based on Equation 19 of Reference 16, considering convective, flexible and rigid responses. This shear is resisted by friction over the base of the tank, with the capacity determined using a coefficient of friction of 0.55 and a normal force equal to the hydrostatic fort.e less the vertical seismic acceleration. Anchor bolts are not required to resist shear, which is reflected in bolt hole clearances. 3.7.5.3 Integral Connections The structural adequacy of anchorage connections on the tank, e.g. bolting chairs or rings, is developed using the derived anchorage loads. While the r.imber and size of anchorage provisions is determined by the magnitude of the derived anchorage loads, a minimum of eight equally spaced holddown bolts is used, independent of tank ;ha,neter. 3.7.5.4 Attached Piping Flexibility of attu:hed piping to accommodate tank displacements is verified to preclude local pressure boundary overstresses. During a seismic event, deflections due to tank flexure are imposed onto connected piping. Piping between the tank and first pipe support is designed with sufficient flexibility to ensure acceptable pipe and tank nozzle stresses. Although friction is taken as preventing slip at the base of the tank, lateral motion imposed on attached piping is conservatively assumed equal to the clearance around anchor bolts. The radial deflection due to tank flexure is based on Equation 13 of Reference 16. References for Section S.7

1. Tsai, N.C., " Spectrum-Compatible Motise for Design Purposes," Journal of Engineering Mechanics Division, ASCE, Vol. 98 EM2, April 1972.

2. Fritz, R.J., "The Effect of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry, Paper No. 71-VIB-100.

3. Mcdonald, C.K., " Seismic Analysis of Vertical PVMPs Enclosed in Liquid Filled Containers,"

ASME Paper No. 75-PVP-56.

4. . . P.J.. " Modal Response on Containment Structures," Seismic Design for Nuclear Power Plai. MIT Pros, Cambridge, Mass.

5. Forsberg, K., "Axisymmetrical and Beam-Type Vibrations of These Cylindrical Shells," AJAA Journal, Volume 7, February 1969.

O Asynrount Design Material . Design of SSC page 3.7 36

_ . _ _ . . - . . . _ __ _ _ _ . _ _ . . . . _ .. _ _ _ -, _ _ .._ .~ _ _ _ . I l System 80+ Desinn controlDocumart I i l

6. ' Lysmer, J., Tabatabaie, M., Tajirian, F., Vahdani, S., Ostadan, F., "SASSI -~A System for the _

Analysis of Soil-Structure Interaction,". Repon No. UCB/GT/81-02, Univ. of California, j

                      . Berkeley, April,1981.                                                                                           q

7. Idriss, I.M., "Eanhquake Ground Motions - Selection of Control Motion and Development of  :

Generic Soil Sites." .l

                                                                                                                                        -i

8. ' ABB Impell Repon No. 01-8503-1784, " Seismic Analysis of the Reactor Building of the System l 80+ Certified Design." l

9. .Impell Corporation, Calculation No. ALWR-2, "SSI Analysis of Case B3.5 with Common l

                      . Basemat," Job No. 8503-003-1355, Revision 6.                                                                       j

10. American Society of Civil Engineers, " Seismic Analysis of Safety Related Nuclear Structures and  ;

Commentary on Standard for Seismic Analysis of Safety Related Nuclear Structures," Publication  ; No.' ASCE 4-86, September 1986.  ! American Society of Civil Engineers, "Struenral Analysis and Design of Nuclear Plant j 11. Facilities," ASCE Manual on Engineering Practice No. 58,1980. j i

12. NUREG-1061, VOLUME 4, U.S. Nuclear Regulatory Commission, " Report of the U.S. Nuclear l Regulatory Commission Piping Review Team," April 1985. l t

l

13. EPRI Repon No. NP-5930, "A Criterion for Determining Exceedance of the OBE," July 1988.

V Newmark, N. M., and Hall, W. J. (1978), Development of Criteria for Seismic Review of i

14. i Selected Nuclear Power Plants, NUREG/CR-0098, U.S. Nuclear Regulatory Commission.

            - 15.       Coats, D.W., Rwammended Revisions to Nuclear Regulatory Commission Seismic Design Cri:eria, NUREG/CR-1161RD. Lawrence Livermore Laboratory, May,1980.                                               l

16. Haroun, M.A. and Housner, G.W., " Seismic Design of Liquid Storage Tanks," Journal of the  :

Technical Councils of ASCE, Vol.107, No. TC1, April,1981, American Society of Civil l Engineers, pp. 191-207. ,

17. Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants. ,

t

18. Veletsos, A.S. and Tang, Y., " Rocking Response of Liquid Storage Tanks," Journal of 1 Engineering Mechanics, Vol 113, No.11, November,1987 American Society of Civil  :

Engineers, pp. 1774-1792.

19. Pries 4, M.J.N, Editor and Chairman, " Seismic Design of Storage Tanks," Recommendations .

of a Study Group of the New Zealand National Society for Eanhquake Engineering, December,

                      '1986.                                                                                                              ,

i

20. - Buckling of Thin-Walled Circular Cylinders," NASA SP-8007, National Aeronautics and Space Administration, August,1%8. .

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21. EPRI Report No. NP-6695, " Guidelines for Nuclear Plant Response to an Earthquake,"

December 1989.

22. EPRI Report No. TR-100082, " Standardization of Cumulative Absolute Velocity," December 1991.

Table 3.7-1 Darnping Values

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Structure Safe Shutdown Earthquaketti (Percent of Critical) Welded steel structures 4.0 Bolted steel structures 7.0 Prestr:ssed concrete structures 5.0 Reinforced con.: rete stmetures 7.0 Equipment (steel assembly) 3.0 0 (( Piping (diameter s12 inches) 2.0}}l')

  \(Piping (diameter >12 inches)                                                    3.0}}l')

llPiping (umform envelope response spectrum method 5.Ol2)}}I') ofanalysis) I'l Soil material dampings are provided in Section 2.5.2. (21 Use of 5% damping for uniform envelope response spectrum analysis of piping is subject to the same limitations as those for Code Case N-411-1 specified in Regulatorv Ocide 1.84. l3) WRC Staff approval is required prior to implementing a change in t'ais information; see DCD Introduction L'ection 3.5. Alywond Des (gr$* ~'etenini Desigra of SSC (11/96) Page 3.7-38

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                                               . Appendix 3.7A- ,                                                               (

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_. _ ~ , fystem 80+ Deskn ControlDocument (T k - Appendix 3.7A i Coupled Reactor Coolant System Seismic Results i a e 4 Contents Page Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7A- 1 Tables 3.7A-1 Natural Frequencies and Dommant Degrees of Freedom for Significant Modes of the CoupleJ Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . 3 .7 A-2 f

  's 3.7A-2        Load Tables for Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . .                  3.7A-3 t

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System 80+ Deskn ControlDocument f Oveniew This Appendix contains the results of the coupled Reactor Coolant System (RCS) seismic analyses. Table 3.7A-1 contains the RCS coupled frequencies and modes (refer to Figures 3.7-23 and 3.7-24 for l T directions and joint locations) with rigid supports at the building RCS support interfaces. Table 3.7A-2 contains the seismic loads due to the envelope of all SSE soil cases. i i e l l

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System 80+ Design ControlDocument Table 3.7A-1 Natural Frequencies and Dominant Degrees of Freedom for Significant Modes of the Coupled Reactor Coolant System g Dominant Degrees of Freedom Mode Freq. Joint Number Direction Location (Hertz) 1 1.7 9912 X RV Internals 2 1.7 9912 Y RV Internals 3 10.9 9996 X RV 4 11.2 9996 Y RV 6 12.1 404, 3404 Y SG 9 12.5 1103, 2103, 4103, 5103 Y RCF Motor 10 12.6 1103,2103,4103,5103 X RCP Motor 11 13.4 1103,2103,4103,5103 X, Y RCP Motor 14 13.4 1103,2103,4103,5103 X, Y RCP Motor 16 16.1 409, 3409 X SG 17 17.5 9995 X RV 1S 17.6 1103,1101, 2103, 2101, etc. Z RCP 22 18.3 9912,9996 Z RV Internals, RV 24 20.5 404, 3404 Z SG 26 20.6 1101, 2101, 4101, 5101 X RCP 27 20.6 1101, 2101, 4101, 5101 Y RCP 29 21.3 9995 Y RV 31 25.6 9908 Y RV Internals 34 27.4 6120 X Pressurizer 35 27.4 6120 Y Pressurizer 38 31.5 9912 Z RV Internals 412, 3412 Z SG Internals i 47 32.4 47 40.4 6130 Z Pressurizer O Approveef Design Materief Design of SSC Page 3.7A-2 I l l

System 80+ Deslan contrat Document Table 3.7A-2 Load Tables.for Reactor Coolant System Seismic Excitation - SSE Seismic Loads, Kips and Ft-Kips > Support Location Reaction Component Calculated Maximum Steam generator upper key Fx 697 Steam generator snubber assembly Fy 1502 Steam generator senical pad Fz (1,3) 744 Fz (2,4) 2072 Steam generator lower key Fx 637 Reactor vessel horiz. column suppon Fc 2076 Reactor vessel column base Fa 18 Fb 1615 Fc 419 Ma 510 Mb 675 i Mc 183 Pump venical column Fz 169

        .eg Pump snubber                                             Fa                    508 Pump upper horizontal column                             Fa                    257 Pump lower horizontal column                             Fa                     85
,                    Pressurizer key                                          Fk                     92 Pressurizer suppon skin                                  Fv                    223                  ;

Fh 204 Mt 11 , Mb 1334 Reactor vessel inlet nozzle Fa 122 Fb 60 Fc 115 Ma 382 Mb 370 Mc 245 Reactor vessel outlet nozzle Fa 768 Fb 276 Fc 217 Ma 624 Mb 1506 Mc 2043 t w.::w unwier.w or ssc rape 2. u .2  :

System 80 + Design ControlDocument Table 3.7A-2 Load Tables for Reactor Coolant System (Cont'd) Seismic Excitation - SSE Seismic Loads, Kips and Ft-Kips Support Location Reaction Component Calculated Maximum Reactor vessel column upper flar.ge Fa 18 Fb 1615 Fc 39 Ma 507 Mb 51 Mc 167 Reactor vessel lower key Fc 417 Steam generator inlet nozzle Fa 772 Fb 228 Fc 287 Ma 1199 Mb 1238 Mc 1095 Steam generator support skirt Fx 912 Fz 4828 Mx 3963 My 3779 hiz 908 Steam generator outlet nozzle Fa 37 Fb 104 Fc 33 hia 321 hib 281 Mc 351 Pump inlet nozzle Fx 86 Fy 62 Fz 76 Mx 297 My 370 Mz 259 Pump our'.et nozzle Fa 161 Fb 56 Fc 17 Ma 213 Mb 213 Mc 733 O Approved Design Atotwin! Desigre of SSC Page 3.7A4

System 80+ Design ControlDocument O U Table 3.7A-2 Load Tables for Reactor Coolant System (Cont'd) Seismic Excitation - SSE Seismic Loads, Kips and Ft-Kips Support Location Reaction Component Calculated Maximum Pump skirt / casing interface Fx 144 Fy 87 Fz 583 Mx 405 My 287 Mz 78 - Pump motor support upper flange (lantern top) Fx 132 Fy 155 Fz 237 Mx 2605 My 2955 Mz 241 Pump motor support lower flange (lantern bottom) Fx 132 Fy 155 Fz 237 Mx 1276 My 1836 Mz 241 Piping at reactor vessel inlet nozzle . M max 542 Piping at reactor vessel outlet nozzle M max 2065 Piping at steam generator inlet nozzle M max 1633 Piping at steam generator outlet nozzle M max 470 Piping at pump inlet nozzle M max 480 i Piping at pump outlet nozzle M max 748 i 4 l NJ w eed w neererw.o wenerssc page 2.7A.5 l

System 80+ oesten contrat Document l

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(~ Effective Page Listing Appendix 3.7B - 5 Pages Date i,il 1/97 lii - viii - Original ix 11/96  ; x Original 3.7B-1 through 3.7B-138 Original 3,7B-139 11/%

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m 1 s>- - ss-Sy= tem 80+ Deslan controlDocument Appendix 3.7B ) Soil Structure Interaction (SSI) Analysis Methodology and Results Nuclear Island Structures 1 Contents Page 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 3.7B-1  : 1.1 S ASSI Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-1 1.2 Computation of Impedances . . . . . . . . . . .......... .......... 3.7B-3 1.3 Computation of Scattering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-4 1.4 Structural Analysis and Generation of Transfer Functions . . . . . . ...... 3.7B-5

1.5 SSI Analysis Cases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-6 j 1.6 Generation of Response Spectra . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-6 '

(s 1.6.1 Generation of Acceleration Time Histories .................... . 3.7B-6 1.6.2 Combination of Global Response ..................... ...... 3.7B-6 , 1.6.3 Computation of Response Spectra ... .. . . . . . . . . . . . . . . . . . . . 3 .7B-7  ; 1.7 Computation of Element Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 .7B-7 ' 1.8 Output Locations ............. 3 . 7B-8 l 1.9 Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-9 l 2.0 Parametric Studies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-9 l i 2.1 Soil Property Variation Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-9 ' 2.2 Effects of Cracked Concrete .......... . . . . . . . . . . . . . . . . . . . 3 .7B-9 2.3 Effects of Debonding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-10 2.4 Connection of Subsurface Slabs to Exterior Walls . . . . . . . . . . . . . . . . . 3.7B-10 3.0 Two-Dimensional SSI Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-11 Tables 3.7B-1 Seismic Analysis Cases ................................. 3.7B-13 3.7B-2 Soil Layers and Properties .......................... . .. 3.7B-14 3.7B-3 Soil Layers and Properties ............. ,................ 3.7B-14 3.7B-4 Soil Layers and Properties ........................... ... 3.7B-15 3.7B-5 Soil Layers and Properties .... .......................... 3.7B-16 3.7B-6 . Soil Layers and Properties ................. ............. 3.7B-17 3.7B-7 Soil Layers and Properties ............................... 3.7B-18 i v 4 proved Design AtenwW Deeign of SSC Page M

System 80+ Design ControlDocument Tables (Cont'd.) Page 3.7B-8 Soil Layers and Properties . .. ... ..... . . ...... .... 3.7B-19 3.7B-9 Soil Layers and Propenies ..... ............. ........... 3.7B-20 3.7B-10 Soil Layers and Properties ... . ..................... . 3.7B-21 3.7B-11 Soil Layers and Propenies ................ .. ... ....... 3.7B-22 Figures 3.7B 1 System 80+ Foundation Configuration ....... . . ......... 3.7B-23 3.7B-2 Location for Computation of Impedances ..... ..... . ..... 3.7B-24 3.7B-3 Typical Three-Dimensional Model for Impedance Computation .. . . 3.7B-25 3.7B-4 Computation of Scattering by Two-Dimensional Models . . . . .. ... 3.7B-26 3.7B-5 Typical Foundation Mesh for Scattering Computation . .......... . 3.7B-27 3.7B-6 Computation of Total Acceleration Response . . . . . . . . . . .. . 3.7B-28 3.7B-7 SSE Response Spectra, All Soil Cases, East-West,2% Damping, Basemat, Elevation +50' .. . . ........ . ................ ... 3.7B-29 3.7B-8 SSE Response Spectra, All Soil Cases, Nonh-South,2% Damping, Basemat, Elevation +50' . . .......... ............ . . . . 3.7B-30 3.7B-9 SSE Response Spectra, All Soil Cases, Venical,2% Damping, Basemat, Elevation +50' .... .... ......... ....... . . . 3.7B-31 3.7B-10 SSE Response Spectra, All Soil Cases, East-West, 5% Damping, Basemat, Elevation +50' . ...... .................. 3.7B-32 3.7B-Il SSE Response Spectra, All Soil Cases, Nonh-South, 5% Damping, Basemat, Elevation +50' . . .... . ........... ... .. 3.7B-33 3.7B-12 SSE Response Spectra, All Soil Cases, Vertical,5 % Damping, Basemat, Elevation +50' . .............. ...... ........ .. 3.7B-34 3.7B-13 SSE Response Spectra, All Soil Cases, East-West,2 % Damping, Interior Structure, Elevation +210' . ........... . . ... .. .... 3.7B-35 3.7B-14 SSE Response Spectra, All Soil Cases, North-South 2% Damping, Interior Structure, Elevation +210' . ...... ............. 3.7B-36 3.7B-15 SSE Response Spectra, All Soil Cases, Vertical,2% Damping, Interior Structure, Elevation +210' .... ........ ... . ........ 3.7B-37 3.7B-16 SSE Response Spectra, All Soil Cases, East West,5 % Damping, Interior Structure, Elevation +210' . ..... .............. ... .. 3.7B-38 3.7B-17 SSE Response Spectra, All Soil Cases, North-South, 5% Damping, Interior Structure, Elevation +210' .............. . .. .. 3.7B-39 3.7B-18 SSE Response Spectra, All Soil Cases, Vertical,5% Damping, Interior Structure, Elevation +210' ..... .. ........ ........ .. 3.7B-40 3.7B-19 SSE Response Spectra, All Soil Cases, East-West,2% Damping, Steel Containment Vessel, Elevation +251' . ... ..... . . . 3.7B-41 3.7B-20 SSE Response Spectra, All Soil Cases, Nonh-South,2% Damping, Steel Containment Vessel, Elevation +251' ... ........... .. ..... 3.7B-42 3.7B-21 SSE Response Spectra, All Soil Cases, Vertical, 2% Damping, Steel Containment Vessel, Elevation +251' ........... .... . ... 3.7B-43 3.7B-22 SSE Response Spectra, All Soil Cases, East-West,5% Damping, Steel Containment Vessel, Elevation +251' ............ ... .. .... 3.7B-44 l 3.7B-23 SSE Response Spectra, All Soil Cases, Nonh-South,5 % Damping, Steel Containment Vessel, Elevation +251' ............ .... ......  ?.73-45 Appetved Design Material- Design of SSC Pagetv

I i System 80+ Desen Con 001 Document. Figures (Cont'd.) Page , 1

           ' 3.7B       SSE Response Spectra, All Soil Cases, Vertical, 5% Damping, Steel                                             i Containment Vessel, Elevation +251' . . . . . . . . . . . . . . . . . . . . . . .                   3.7B-46   i 3.7B-25         SSE Response Spectra, All Soil Cases, East-West,2% Damping, Shield                                            !

Building, Elevation +263.5' . . . . . . . . . .. . . , . . . . . . . . . . . . . . . . . 3.7B-47 i 3.7B-26 SSE Response Spectra, All Soil Cases, North-South,-2% Damping, i Shield Building, Elevation +263.5' ......................... 3.7B-48 l 3.7B-27 SSE Response Spectra, All Soil Cases, Venical,2% Damping, Shield Building, Elevation +263.5' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-49 l

           ' 3.7B-28         SSE Response Spectra, All Soil Cases, East-West,5% Damping, Shield                                            l Building, Elevation + 263.5 ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-50 3.78-29       . SSE Response Spectra, All Soil Cases, Nonh-South, 5% Damping,                                                 ;

Shield Building, Elevation +263.5' ......................... 3.7B-51  ! 3.7B-30 SSE Response Spectra, All Soil Cases, Venical,5% Damping, Shield 1 Building, Elevation + 263.5' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-52 l 3.7B-31 SSE Response Spectra, All Soil Cases, East-West, 2% Damping,- Fuel j Building, Elevation +170' ................. .......... .. 3.7B-53 l 3.7B-32 SSE Response Spectra, All Soil Cases, Nonh-South,2% Damping, Fuel

                           ' Building, Elevation +170' ...............................                                          3.78-54 3.7B-33          SSE Response Spectra, All Soil Cases, Venical, 2% Damping, Fuel                                               ;

Building, Elevation +170' ....... ................. .... 3.7B-55 3.7B-34 SSE Response Spectra, Ali Soil Cases, East-West,5% Damping, Fuel Building, Elevation +170' ............................... 3.7B-56 3.7B-35 SSE Response Spectra, All Soil Cases, North-South,5% Damping, Fuel j Building, Elevation +170' .............................. 3.7B-57 } 3.7B-36 SSE Response Spectra, All Soil Cases, Venical, 5% Damping, Fuel l Building, Elevation +170' ............................... 3.7B-58 l 3.7B-37 SSE Response Spectra, All Soil Cases, East-West,2% Damping, Control  ! Area 1, Elevation +130' ................................ 3.7B-59 l 3.7B-38 SSE Response Spectra, All Soil Cases, Nonh-South, 2% Damping, Control Area 1, Elevation + 130' . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-60 i 3.7B-39 SSE Response Spectra, All Soil Cases, Venical,2% Damping, Control  ! Area 1, Elevation +130' ................................ 3.7B-61 3.7B-40 SSE Response Spectra, All Soil Cases, East-West,5 % Damping, Control , Area 1, Elevation +130' .............................. . 3.7B-62 l 3.7B-41 SSE Response Spectra, All Soil Cases, Nonh-South, 5% Damping, Control Area 1, Elevation + 130' . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7B-63 i 3.7B-42 SSE Response Spectra, All Soil Cases, Venical,5% Damping, Control Area 1, Elevation +130' ..................... .......... 3.7B-64 3.7B-43 Comparison of Base Case B3.5 vs. Upper Bound B3.5, East-West, 5% Damping, Steel Containment Vessel, Elevation +251' ............ 3.7B-65 3.7B-44 Comparison of Base Case B3.5 vs. Upper Bound B3.5, East-West, 5% Damping, Interior Structure, Elevation +210' ................ 3.7B-66 3.7B-45 Uncracked vs. Cracked Concrete, Case B3.5, East-West,5 % Damping, Interior Structure, Elevation +210' ... ..................... 3.7B-67 3.7B-46 Uncracked vs. Cracked Concrete, Envelope of All cases with Uncracked p Concrete, Cracked Concrete -with Case A-1 and Fixed-Base (with .1 Concrete Backfill) 5% Damping, Interior Structure Elevation +91.75', East-West .... ..................................... 3.7B-68 w o ,aan.,w.o e,,erssc rue. v t

               .~            . . . . .     .       .                             _                                -

System 80+ Design ControlDocument i Figures (Cont'd.) Page 3.7B-47 Uncrac8.ed vs. Cracked Concrete, Envelope of All Cases with Uncracked  ; Concrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete Backfill) 5% Damping, Interior Structure Elevation +91.75', , North-South ............... ................. . ... 3.7B-69 l 3.7B-48 Uncracked vs. Cracked Concrete, Envelope of All Cases with Uncracked Concrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete Backfill) 5% Damping, Interior Structure Elevation +210', East-West .............. .......... ............. . 3.7B-70 3.7B-49 Uncracked vs. Cracked Concrete, Envelope of All Cases with Uncracked Concrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete Backfill) 5% Damping, Interior Structure Elevation +210', North-South .......... . ...... ........ .... ..... 3.7B-71 3.7B-50 Uncracked vs. Cracked Concrete, Envelope of All Cases with Uncracked Concrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete 'drekfill) 5 % Damping, Control Area A Elevation + 115', East-West .................... ....................... 3.7B-72 3.7B-51 Uncrr.:ked vs. Cracked Concrete, Envelope of All Cases with Uncracked Concrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete Backfill) 5% Damping, Control Area A Elevation +115', North-South . ........... .......................... 3.7B-73 3.7B-52 Uncracked vs. Cracked Concrete, Envelope of All cases with Uncracked C(ncrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete Backfill) 5% Damping, Fuel Building Elevation +170', East-West ........ . ....... ....... .... ...... . 3.7B-74 3.7B-53 Uncracked vs. Cracked Concrete, Envelope of All Cases with Uncracked Concrete, Cracked Concrete with Case A-1 and Fixed-Base (with Concrete Backfill) 5% Damping, Fuel Building Elevation +170', North-South . .. .. ..... ... .. ... ... .. .. 3.7B-75 3.7B-54 Bonded vs. Partially Debonded Foundation, Case B3.5, E-2, 5% Damping, Steel Containment Vessel, Elevation +251' .......... 3.7B-76 3.7B-55 Bonded vs. Partially Debonded Foundation, Case B3.5, E-W, 5% Damping, Interior Structure, Elevation +210' . . ...... 3.7B-77 3.7B-56 Base Case vs. Slabs-to-Exterior Walls, Soil Case B3.5, E-W, 5% Damping, Steel Containment Vessel, Elevation +251' ... . . 3.7B-78 3.7B-57 Base Case vs. Slabs-to-Exterior Walls, Soil Case B3.5, E-W, 5% Damping, Interior Structure, Elevation +210' .. . .. ...... 3.7B-79 3.7B-58 Maximum Floor Accelerations, All Soil Cases, CMS 1, Fuel Building, East-West . . ........ ...... ..... .. ..... . 3.7B-80 3.7B-59 Maximum Floor Accelerations, All Soil Cases, CMS 1, Fuel Building, North-South ........... .. . .. .. ... . ..... ... 3.7B-81 3.7B-60 Maximum Floor Accelerations All Soil Cases, CMSI, Fuel Building, Vertical . ... ...... .......... .............. .. 3.7B-82 3.7B-61 Maximum Floor Acceleraticas, All Soil Cases, CMS 2, Fuel Building, East-West .. . ....... ......... ........ ... . .. 3.7B-83 3.7B-62 Maximum Floor Accelerations, All Soil Cases, CMS 2, Fuel Building, North-South ........... ........ .... .. ...... ... 3.7B-84 3.7B-63 Maximum Floor Accelerations, All Soil Cases, CMS 2, Fuel Building, Vertical . . . ................ ........ .. . ... 3.7B-85 Approvuut Design Motorial Design of SSC Page vi i

l S3 tem 80+ Design ControlDocument ( Figures (Cont'd.) Page  ; C l 3.7B-64 Maximum Floor Accelerations, All Soil Cases, CMS 3, Fuel Buildmg,  ! East-West ......................................... 3.7B-86 3.7B-65 Maximum Floor Accelerations, All Soil Cases, CMS 3, Fuel Building, Nonh-South . ................................... . 3.7B-87  ! 3.7B-66 Maximum Floor Accelerations, All Soil Cases, CMS 3, Fuel Building, Vert ical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... .. .... 3.7B-88 3.7B-67 Maximum Floor Accelerations, All Soil Cases, CMS 1, Interior Structure, i l East-West . .......................... ........... 3.7B-89 3.7B-68 Maximum Floor Accelerations, All Soil Cases, CMS 1, Interior Structure, Nonh-South .......... ........ ... ... ...... ... . 3.7B-90 1 3.7B-69 Maximum Floor Accelerations, All Soil Cases, CMS 1, Interior Stmeture, Venical . . . . . . . . . . . . ................ . ........... 3.7B-91 3.7B-70 Maximum Floor Accelerations, All Soil Cases, CMS 2, Interior Structure, i East-West .................... .. ........... ..... 3.7B-92 l 3.7B-71 Maximum Floor Accelerations, All Soil Cases, CMS 2, Interior Structure, l Nonh-South ... ..... ........ ....... .. .......... 3.7B-93 3.7B-72 Maximum Floor Accelerations, All Soil Cases, CMS 2, Interior Structure, Vertical . . . ..... ........... .......... .. .. . 3.7B-94 3.7B-73 Maximum Floor Accelerations, All Soil Cases, CMS 3, Interior Structure, East-West .......... .. ....... .. .... .......... 3.7B-95 3.7B-74 Maximum Floor Accelerations, All Soil Cases, CMS 3, Interior Structure,  ! g Nonh-South ................ . . ... .. ..... ...... 3.7B-96 i 3.7B-75 Maximum Floor Accelerations, All Soil Cases, CMS 3, Interior Struare, i (V Venical ... .. . ......... . ... . . .... .... 3.7B-97 l 3.7B-76 Maximum Floor Accelerations, All Soil Cases, CMS 1, Shield Building, l East-West ............... .. ........ ..... ..... 3.7B-98  ! 3.7B-77 Maximum Floor Accelerations, All Soil Cases, CMS 1, Shield Buildmg, J Nonh-South .... ... .... ..... ... .. ......... . 3.7B-99 3.7B-78 Maximum Floor Accelerations, All Soil Cases, CMS 1, Shield Building, Venical . . . . . . . . ..... .. .. .. . .. .. . .... 3.7B-100 3.7B-79 Maximum Floor Accelerations, All Soil Cases, CMS 2, Shield Building, East-West ......... . .... .. ..... ...... .... 3.7B-101 3.7B-80 Maximum Floor Accelerations, All Soil Cases, CMS 2, Shield Building, Nonh-South ...... . .... .. . ... .... .. ... .. 3.7B-102 3.7B-81 Maximum Floor Accelerations, All Soil Cases, CMS 2, Shield Building, Venical . . . . . . . . . . . ... ... .......... ....... . 3.7B-103 3.7B-82 Maximum Floor Accelerations, All Soil Cases, CMS 3, Shield Building, East-West . .............. .... .. ......... ... 3.7B-104 1 3.7B-83 Maximum Floor Accelerations, All Soil Cases, CMS 3, Shield Building, Nonh-South ........ . ... .... .. ... . .. ... 3.7B-105 3.7B-84 Maximum Floor Accelerations, All Soil Cases, CMS 3, Shield Building, Venical . . . . . . . . . . .. ............ .............. 3.7B-106 3.7B-85 Cumulative Stick Shears, CMS 1 Motion, East-West Direction . ... 3.7B-107 3.7B-86 Cumulative Stick Shears, CMS 1 Motion, Nonh-South Direction ... .. 3.7B-108 3.78-87 Cumulative Stick Moments About East-West Axis, CMSI Motion . . 3.7B-109 3.7B-88 Cumulative Stick Moments About North-South Axis, CMS 1 Motion 3.7B-110 p) (~ - 3.7B-89 Cumulative Stick Axial Forces, CMS 1 Motion, Venical Direction . . 3.7B-111 3.78-90 Cumulative Stick Torques, CMS 1 Motion ..... . ...... ... 3.7B-112 Appmved Design Materiel Design of SSC Page vi

System 80+ Design ControlDocument l Figures (Cont'd.) Page , 3.7B-91 Direction Cumulative Stick Shears, CMS 2 Motion, East-West ... .... 3.7B-113 3.7B-9~2 Cumulative Stick Shears, CMS 2 Motions, North-South Direction ...... 3.7B-114 3.7B-93 Cumulative Stick Moments About East-West Axis, CMS 2 Motion ..... 3.7B-115 3.7B-94 Cumulative Stick Moments About North-South Axis, CMS 2 Motion .... 3.7B.116 3.7B-95 Cumulative Stick Axial Forces, CMS 2 Motion, Venical Direction .. . 3.7B-117 3.7B-96 Cumulative Stick Torques, CMS 2 Motion ................. .. 3.7B-118 3.7B-97 Cumulative Stick Shears, CMS 3 Motion, East-West Direction . . ..... 3.7B-119 3.7B-98 Cumulative Stick Shears, CMS 3 Motion, North-South Direction ...... 3.7B-120 3.7B-99 Cumulative Stick Moments About East-West Axis, CMS 3 Motion ... . 3.7B-121 3.7B-100 Cumulative Stick Moments About North-South Axis, CMS 3 Motion .... 3.7B-122 3.7B-101 Cumulative Stick Axial Forces, CMS 3 Motions, Venical Direction . . . . . 3.7B-123 3.7B-102 Cumulative Stick Torques, CMS 3 Motion .... .............. 3.78-124 3.7B-103 2-D SSI Model of Nuclear Island, Nonh-South Direction . . . . . . . . . . . 3.7B-125 3.7B-104 2-D SSI Model of Nuclear Island, East-West Direction ......... .. 3.7B-126 3.7B-105 Representation of Turbine Building on East Side of Nuclear Island (NI) . ............................. ...... 3.7B-127 3.7B-106 Maximum Soit Dynamic Pressure Distributions Along the Basemat/ Soil Interfu, All Soil Cases, Nonh-South Earthquake, CMS 1 Motion, North-South Model ... .. .... .. ........... ... 3.7B-128 3.7B-107 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, Nonh-South Eanhquake, CMS 2 Motion, Nonh-South Model ......... ..... ..... .... . .. 3.7B-129 3.7B-108 Maximum Soit Dynamic Pressure Distribution Along The Basemat/ Soil Interface, All Soil Cases, Nonh-South Eanhquake, CMS 3 Motion, Nonh-South Model . ... ............ ...... .. ..... 3.7B-130 3.7B-109 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, Venical Earthquake, CMS 1 Motion, Nonh-South Model .... ................ ...... ... 3.73-131 3.7B-110 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, Venical Eanhquake, CMS 2 Motion, North-South Model ......... ..... .... .......... 3.7B-132 3.7B-111 Maximum Soit Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, Venical Eanhquake, CMS 3 Motion, Nonh-South Model .... ... . .. ....... ... ...... 3.7B-133 3.7B 112 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, East-West Earthquake, CMS 1 Motion, East-West Model .. ... ... ... . . ............ ... 3.7B-134 3.7B-113 Maximum Soil Dynamic Pressure Distributions Along the Basemat/ Soil Interface, All Soil Cases, East-West Eanhquake, CMS 2 Motion, East-West Model ........... . .. ..... .. ... .... 3.7B-135 3.7B-114 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, East-West Earthquake, CMS 3 Motion, East-West Model ...... . ............. ... ..... . 3.7B-136 3.7B-115 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, Vertical Earthquake, CMSI Motion, East-West Model . . . . . . . . . . .......... . . ............ . 3.7B-137 Approved Design Material Design of SSC Pope vm

I System 80+ Design ControlDocument . Figures (Cont'd.) Page 3.7B-116 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil Interface, All Soil Cases, Venical Eanhquake, CMS 2 Metion, East-West Model . . . . . . . ................................. . 3.7B-138 3.7B-117 Maximum Soil Dynamic Pressure Distribution Along the Basemat/ Soil  ! Interface, All Soil Cases, Vertical Earthquake, CMS 3 Motion, l East West Model .................. ................. 3.7B-139 3.7B-118 Maximum Lateral Dynamic Pressure Distributaon Along the Embedded Wall / Soil Interface, All Soil Cases, N-S Eanhqmke, CMS 1 Motion, > Nonh-South Model ................. ...... ......... 3.7B-140 i 3.7B-119 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, All Soil Cases, Venical Earthquake, CMS Motion, Nonh-South Model .................................. 3.7B-141 3.7B-120 Maximum Lateral Dynamic Pressure Distribution Along the Ebedded Wall / Soil Interface, All Soil Cases, North-South Eanhquake, CMS 2 Motion, Nonh-South Model ....................... ..... 3.7B-142 3.7B-121 Maximum Lateral Dynamic Pressure Distribution Along the Embended Wall / Soil Interface, All Soil Cases, Venical Eanhquake, CMS 2 Motion, Nonh-South Model .................................. 3.7B-143  ; 3.78-122 Maximum Lateral Dynamic Pressure Distribution Along the Embedded j Wall / Soil Interface, All Soil Cases, N-S Eanhquake, CMS 3 Motion, ' Nonh-South Model ......... ........................ 3.7B-144 3.7B-123 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall /SoilInterface, All Soil Cases, Venical Earthquake, CMS 3 Motion, Nonh-South Model ...,........ ..................... 3.7B-145 3.7B-124 Maximum Lateral Dynamic Pressure Distribution Along the Embeded Wall / Soil Interface, East Wall (Adjacent to Turbine Building), All Soil Cases, East-West Eanhquake, CMSI Motion, East-West Model . . . . . . . 3.7B-146 3.7B-125 Maximum 1.ateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, West Wall (Adjacent to Radwaste Building), All Soil Cases, East-West Eanhquake, CMS Motion, East-West Model ....... 3.7B-147 3.7B-126 Maximum Lateral Dynamic Pressure Distribution Along the Embedded l Wall / Soil Interface, East Wall (Adjacent to Turbine Building), All Soil Cases, East-West Eanhquake, CMS 2 Motion, East-West Model . . . . . . 3.7B-148 3.7B-127 Maximum Lateral Dynamic Pressure Distribution Along the Embedded l Wall / Soil Interface, West Wall (Adjacent to Radwaste Building), All Soil Cases, East-West Eanhquake, CMS 2 Motion, East-West Model . . . . . . . 3.7B-149 i 3.7B-128 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, East Wall Adjacent to Turbine Building), All Soil Cases, East-West Eanhquake, CMS 3 Motion East-West Model . . . . . . . 3.7B-150 3.78-129 Maximum Lateral Dynamic Pressure Disttibution Along the Embedded Wall /SoilInterface, West Wall (Adjacent to Radwaste Building), All Soil Cases, East-West Eanhquake, CMS 3 Motion, East-West Model . . . . . . . 3.7B-151 3.7B-130 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, East Wall (Adjacent to Turbine Building), All Soil Cases, Venical Eanhquake, CMS 1 Motion, East-West Model . . . . . . . . 3.7B-152 3.7B-131 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, West Wall (Adjacent to Radwaste Building), All Soil Cases, Vertical Earthquake, CMS 1 Motion, East-West Model ........ 3.7B-153 Anno, e onw unema. cosw a ssc nrm rope u

System 80+ Design ControlDocument Figures (Cont'd.) Page 3.7B-132 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, East Wall (Adjacent to Turbine Building), All Soil Cases, Venical Earthquake, CMS 2 Motion, East-West Model ........ 3.7B-154 3.7B-133 Maximum Lateral Dynamic Pressure Distribution Along the Embedded q Wall / Soil Interface, West Wall (Adjacent to Radwaste Building), All Soil Cases, Venical Eanhquake, CMS 2 Motion, East-West Model ...... . 3.7B-155 3.78-134 Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, East Wall (Adjacent to Turbine Building), All Soil Cases, Venical Eanhquake, CMS 3 Motion, East-West Model . .. ,.

                                                                                                                  . 3.7B-156 3.7B-135         Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, West Wall (Adjacent to Radwaste Building), All Soil Cases, Venical Eanhquake, CMS 3 Motion, East-West Model                            ....        . 3.7B-157 3.7B-136         SRSS Combination of Maximum Soil Dynamic Pressure Along the Basement / Soil Interface Along the Diagonal of the Basemat, All Soil Cases, CMSI Motion .......... ............ ...                                              ...      3.7B-158 3.7B-137         SRSS Combination of Maximum Soil Dynamic Pressure Along the Basemat/ Soil Interface Along the Diagonal of the Basemat, All Soil Cases, CMS 2 Motion                .       . ...     ......... .. . . ....                           3.7B-159 3.7B-138        SRSS Combination of Maximum Soil Dynamic Pressure Along the Basemat/ Soil Interface Along the Diagonal of the Basemat, All Soil Cases, CMS 3 Motion           ........... ... ....                         .         ......          3.7B-160 3.78-139         Extent of Basemat Uplift Under Postulated Seismic Loads When Combined with Maximum Buoyancy Effects                   .      ....          . ....           . 3.7B-161 3.78-140        Envelope of Maximum lateral Dynamic Pressure Distribution Along the Embedded Wall /SoilInterface, All Soil Cases, All Motions, North-South M od el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . .       ..               . 3.7B 162 3.7B-141        Envelope of Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, West Wall (Adjacent to Radwaste Building), All Soil Cases, All Motions, East-West Model                       .........              3.7B-163 3.7B-142        Envelope of Maximum Lateral Dynamic Pressure Distribution Along the Embedded Wall / Soil Interface, West Wall (Adjacent to Turbine Building), All Soil Cases, All Motions, East-West Model                    . .........               3.7B-164 O

Approved Design Materini Design of SSC Pagex

System 80+ Design ControlDocument f3 y! Overview This Appendix describes the SSI methodology and presents analysis results used to establish seismic design loads for the Nuclear Island (NI) structures and Reactor Coolant System (RCS) of the System 80 + Standard Design. Three dimensional SSI analyses were performed based on a Safe ShtMown Earthquake (SSE) excitation of 0.30g horizontal peak ground acceleration. Three different control motbns (CMS 1, CMS 2, CMS 3) were used as the input excitation. The spectral characteristics of the motions are described in Section 2.5. A set of ten soil profiles developed in Section 2.5 to represent generic site conditions were used as the soil medium in the SST analysis. ~he SSI analysis results are provided in the form of in-structure response spectra corresponding to major elevations, and internal resisting forces at each floor of the NI structures. Two sets of fixed-base analyses with no SSI effects were also performed using the three control motions (CMS 1, CMS 2, and CMS 3) the rock outcrop motion as direct input excitation to the NI foundatien. For the first set of fixed-base analyses, the superstructure is fixed at the basemat elevation only. For the second set of fixed-base analyses, concrete backfill is assumed, which also fixes the sidewalls in addition to the basemat. The SSI analyses for the NI are performed with a common basemat that founds all Reactor Building (RB) and Nuclear Annex (NA) structures. A number of parametric studies were also prepared using the 3-D SSI model. These studies and conclusions are also described in this appendix. In addition to the 3-D SSI analyses,2-D SSI analyses were performed for the NI to determine the effects Q C/ of: 1) Structure-Soil-Structure Interaction effects between the NI and adjacent structures; 2) the maximum dynamic lateral pressures on the embedded walls of the N1; and 3) the bearing pressure distribution along the soil / mat interface. These 2-D SSI analyses and results are also discussed in this appendix. 1.0 Introduction This Appendix describes the Soil-Structure Interaction (SSI) analyses that are performed for the System 80+ Nuclear Island (NI) structures. The SSI analyses utilize the control motions and soil profiles described in Section 2.5 and the structural models described in Section 3.7. The description of the SSI methodology, the analysis process and the results follow below. 1.1 SASSI Methodology For the SSI analyses of the NI structures, the methodology of the computer program SASSI is used. SASSI (System for Analysis of Soil-Structure Interaction) uses a general substructuring method, which is formulated in the frequency domain using the complex response method and the finte element technique. In a substructuring method, the soil strata and halfspace are analyzed first in the frequency domain. From this analysis the impedances at the soil-structure interface are established. Subsequently, these properties are used as boundary conditions in a dynamic analysis of the structure with a loading that depends on the free-field motions. For the System 80+ analyses, a modified SASSI methodology is used, which reduces the solution of the SSI problem to three steps: y~ ) Solution of the s!te response problem to determine the free-field motions within the embedded part of the structure. Atywovest Design Afsterist Onsign of SSC Page 3.781

I l System 80+ Deshn ControlDocument

• Evaluation of the foundation impedances. h
• Solution of the structural problem. This involves forming the complex stiffness matrices and load vector and solving the equations of motion for the final displacements.

SASSI is structured in a modular form. The code is segmented into independent subprograms (modules) which are executed sequentially. Each of these modules performs one of the tasks required in the sequence of the analysis of SSI. It is not necessary to execute all modules for a given SSI analysis. The modules that are executed depend on the type of problem (seismic analysis or forced vibration analysis), the type of model (2-D, 3-D or axisymmetric) and the kind of results desired (transfer functions, accelerations, velocities, displacements). For this analysis the following modules are utilized:

• SITE
• POINT
• IIOUSE
• ANALYS
• RIMP
• COMBIN
• MOTION In the general SASSI approach, the most time-consuming step is the inversion of the impedance matrix of the foundation-soil interface points. The impedance matrix is computed from the flexibility matrix as:

[X]=[F]-3 where [X] and [F] denote the impedance and the flexibility matrices of the foundation / soil interface points, respectively. This method involve.s the complete inversion of the flexibility matrix [F]. Since the soil cases considered cover a wide variation of soil stiffnesses, the vertical mesh discretization, especially for soft soils, has to be sufficiently fine to capture the response at the frequency range of interest. This, coupled with the large size of the System 80+ foundation (380' x 322' x 51.75', shown in Figure 3.7B-1) results in a foundation model which is very large. The finite element discretization of such a model, vertically and ! horizontally, in order to capture the embedded soil profile and the frequency range of interest, would ! require such a large number of interaction nodes that it would be impractical to solve the problem by the general SASSI methodology. An alternate method described below is used. By taking advantage of the foundation rigidity, the method of the SASSI module RIMP can be used. This method greatly reduces the computational effort required. In RIMP, the foundation is assumed to be rigid. Therefore, the complete flexibility matrix [F] of all interaction nodes can be transformed into the flexibility matrix of a single point, by rigid body transformation. The transformation is performed based on the geometry of the foundation. The foundation stiffness is not required because of the inherent rigid body assumption. To compute the impedance matrix, the inversion of the flexibility matrix (size 6x6) AMweved Des &n Matenet Des gn of SSC page 3.78 2

h i Design ControlDocanent Qtem 80+ i j of only a single point, (say, the center of the foundation) is required, as shown in Figure 3.7B-2. The SASSI module RIMP is used in all the analyses by choosing the impedance to be computed at a single  ;

point located at the center of the foundation.

In all subsequent sections, X direction corresponds to the plant EW direction, Y direction corresponds to the plant NS direction and Z direction corresponds to the plant Venical direction.

,                                                                                                                                             i U,       Computation of Impedances Frequency-dependent inW=m, at specified frequencies, are computed by the module RIMP of SASSI.                              a RIMP has the capability to compute impedance matrices of a single rigid embedded foundation of arbitrary shape. The advantage of using RIMP is that the inversion of the complete flexibility matrix of                     ,

the foundation is by-passed. In this analysis, the foundation is modeled as rigid. Based.on the NI layout,  : it is judged that modeling the foundation as rigid is an appropriate representation for the purpose of  ! generating in-structure spectra. This is because there are numerous shear walls in both horizontal  ! directions at all subsurface elevations, which provide significant out-of-plane stiffness to the basemat and  ! sidewalls. For a rigid foundation, the impedance at only one point is necessary. This point corresponds  ; to the location where the superstructure is attached to the foundation. Prior to using RIMP, several steps l 3 are followed, as described below: j

• The module SITE is run to solve the site response problem. The site propenies, wave composition, direction of wave travel and location at which the control motion is applied are ,

;                         defined in this step. For horizontal motion, venically propagating S-waves are considered while                      J O                      for venical motion, venically propagating P-waves are considered. The control motion is specified at the surface (Elev. +91.75 ft.) in all soil profiles. Module SITE is executed for a set of selected frequencies and for each of the soil profiles.

• The HOUSE module is executed for the foundation. The rigid foundation is modeled in 3-D in 3

order to define the foundation shape. The venical discretization is controlled by the thicknesses of the soil layers up to the depth of the embedment. The HOUSE module is executed on this model in order to save the nodal coordinates of the foundation for later use. The mass and stiffness matrices are not required due to the assumption of a rigid foundation. As such, this

HOUSE model contains only nodal point coordinates. The nodal points define the 3-D geometry 4

of the foundation. A typical 3-D foundation mesh is shown schematically in Figure 3.7B-3. The mesh size varies for different soil profiles. t

• POINT 3 (for 3-D analysis) is executed for computation of the flexibility matrix of the interaction nodes. The tributary region of each node is defined by the mesh size of the 3-dimensional HOUSE model of the foundation (step B). This step uses as input the results from the SITE analysis in step A above.
• The module ANALYS is executed to generate the flexibility matrix at all the foundation nodes at the specified frequencies. - This step uses results from POINT 3 and HOUSE above. Because of the large memory requirements for this step, impedances for all the frequencies cannot be computed in a single run. -Hence, this step is repeated for various sets of frequencies to cover the frequency range of interest and the results are combined, as discussed later.

O

                 %. 'De& aineerint- One&o of SSC                                                                             Page 3.76-3

Syrtem 80+ _ Design ControlDocument

• The HOUSE module is then executed once for the horizontal and vertical superstructure models.

The mass and stiffness matrices and the blocking information of these two matrices are saved. For the impedance computation, RIMP requires only the blocking information, which is passed on later to ANALYS.

• The module RIMP is executed for computation of a 6x6 unpedance matrix for the rigid foundation using the results from the previous e ps. This is performed for each speci5ed frequency. RIMP is executed twice, for the horiental model and vertical model separately so that the output files it creates are compatible with the corresponding superstmeture from HOUSE.

The impedance matrices at the center of the foundation (Elev. +40 ft.) are computed and saved for the horizental and vertical models. 1.3 Computation of Scattering While RIMP computes the impedance matrix, it does not compute tce scattering matrices required to compute the foundation response due to input motions at the ground surface. For significant embedmem, such as that for the System 80+ structures, which have embedments as deep as 50.75 ft., the surface and foundation motions are not similar. The motion at the foundation level is in general less than that at the surface. To account for the proper foundation response, the scattering matrices for this analysis are computed externally in a separate step. These scattering matrices, together with the impedances computed above, are then read by the SASSI module ANALYS for further processing to determine structure responses. To compute the scattering matrices, advantage is taken of the following characteristics of the foundation:

• The foundation is symmetrical (for input in a particular direction, some cross-terms in the scattering matrix are zero, e.g. translation and rocking in the perpendicular direction, and torsion).
• The foundation is rectangular (two perpendicular cross-sections are sufficient to define the geometry).

Both the above characteristics are used to simplify the scattering computation. Referring to Figure 3.7B-2, the scattering matrix has the following characteristics: For X-direction input: ui # 0, u #s 0, u = u3 = u,= u = 6 0 (due to synunetry) For Y-direction input: u2 # 0, u4# 0, ui = q = u3 = u6 = 0 (due to synunetry) For Z-direction input: u3 # 0, u3 = u2 = u3 = u4 = u3 = u6 = 0 (due to symmetry) From the above derivations, the complete 3-D foundation characteristics can be approximated adequately by two-dimensional models representing the foundation geometry in two perpendicular cross-sections. For the System 80+ standard plant, since the foundation is rectangular with a uniform cross section, the Approved Design Materieh Design e? SSC Page 3.78 4

System 80+ Design controlDocument 3-D response can be approximated adequately by two 2D models in the XZ and YZ planes respectively, as illustrated in Figure 3.7B-4. The two-dimensional model in the X-Z plane is used to compute the scattering matrix for the X-direction and Z-direction input motions. The two-dimensional model in the Y-Z plane is used to compute the scattering matrix for input motion in the Y-direction. The vertical response is obtained by applying vertical input motion to the XZ two-dimensional model. Using the Beneral SASSI approach, the scattering matrices related to the foundation response due to a control motion applied at the soil surface are computed by using the two 2-D models of the foundation. The results from the two-dimensional models are later combined to derive a 6x3 scattering matrix, which is frequency-dependent, at the point where the superstructure is attached to the foundation, for each discrete frequency of interest, for the 3-D foundation. i The foundation is modeled as rigid. Plane-strain elements with rigid properties and rigid beam elements are used to model the side walls and the foundation base in order to provide translational and rotational degrees of freedom at the walls and the foundation base for subsequent computation of scattering matrices in six degrees of freedom. The model discretization varies with the soil profile used and is controlled by the thickresses of the soil layers. Typical schematics of the two-dimensional meshes of the foundation model are shown in Figure 3.7B-5. 1.4 Structural Analysis and Generation of Transfer Functions Generation of transfer functions at key locations of the structures is the next step after the impedance and 1 scattering matrix computations. This is accomplished through the following process. l The stiffness and mass information of the general 3-D superstructures (NI stmetures) are generated using the module HOUSE. The solution of the combined system (soil, foundation and superstructures) is carried out in the module ANALYS. ANALYS (Mode 6) is executed to generate the transfer functions at all superstructure nodes for control motions applied in the X, Y and Z directions using the impedance matrices from RIMP and the superstructure data from liOUSE (superstructure). The appropriate data corresponding to the superstructure models are usui for each motion direction. The transfer functions are computed at the specified frequencies for each af the directions X, Y and Z. The frequencies need not be identical in the three directions. l i Because the analyses are performed in the frequency domain, the transfer functions are generated up to a maximum " cutoff" frequency. Cutoff frequencies are the maximum frequencies that the soil media can transmit without loss of accuracy in the solution. In the present analyses, cutoff frequencies are computed based on the dimensions of the soil discretization. The maximum frequency that a soil layer can transma corresponds to a wavelength equal to h/5, where, h is the layer thickness. If the maximum frequency is found to be higher than 40 Hz, a cutoff frequency of 40 Hz is selected. This is consistent with the fact that the rock outcrop input motion has no frequency content beyond 40 Hz. Cutoff frequencies vary according to the soil profiles used in the analyses. Table 3.7B-1 summarizes the cutoff frequencies for all analysis cases. The same frequencies as in the impedance computation are analyzed. NNeeved Design nietodeh Design of SSC Page 3.78-5

System 30+ Design ControlDocument 1.5 SSI Analysis Cases A summary of all SSI analysis cases is presented in Table 3.7B-1. Ten SSI analyses are performed using all generic soil profiles described in Section 2.5 of the Approved Design Material. All analyses are three-dimensional with input excitation provided in three directions simultaneously. The generic soil sites differ from each other with respect to soil properties and depth of soil over bedrock. As described in Section 2.5, there is one case with embedment depth to bedrock, five cases with depth of soil to bedrock of 100 ft. and four cases with soil depth to bedrock of 200 ft. The embedment depth of the RB is the same, approximately 51 feet, in all cases. The soil layers used in the SSI models and their associated properties are shown in Tables 3.7B-2 to 3.7B-11 for all soil cases respectively. 1.6 Generation of Response Spectra Ger.eration of response spectra first involves the computation of acceleration time histories at selected in-structure locations, combination of these time histories from the three directions of excitations and computation of response spectra at specified frequency points and damping values, as described in the following subsections. 1.6.1 Generation of Acceleration Time Histories Acceleration time histories at key locations from the horizontal and vertical models due to the three control motions are computed. The time histories are then added to obtain one final time history at each location. The locations selected are summarized in Section 1.8. Time histories are computed for X, Y and Z translations, and, for certain locations in the Interior Structure, rotational time histories are also computed for input in subsystem analysis.

• The module MOTION is first used to generate transfer functions in the frequency domain at key locations. MOTION uses the uninterpolated transfer functions at the computed frequencies from ANALYS and performs the interpolation for intermediate frequencies. The transfer functions are computed at 4096 frequency poir . based on the duration of the control motion, which is 20.48 seconds at a time step of 0.005 seconds. At this step, the control motions are not yet applied.
• Once the transfer functions are finalized, MOTION is again executed for the same set of frequencies in step (a), this time multiplying the control motions with the transfer functions in the frequency domain and then obtaining acceleration time histories in the time domain through a Inverse Fourier Transform technique. In this step, MOTION convolutes the control motions and outputs results in the time domain. Three sets of MOTION runs are made, for X, Y and Z directions, using the appropriate transfer function file in each direction with output at identical locations. This step is repeated three times, for control motions in X, Y, and Z direction. All control motions are for a duration of 20.48 seconds at a time step of 0.005 seconds.

1.6.2 Combination of Global Response At each location, the output time histories from the horizontal motions are added algebraically to the corresponding time histories from the vertical motion, in order to obtain the resultant time histories due to the combined effect of motions in the X,Y and Z directions. Appromt Design Afstorial Design of SSC Page 3.78-6

Systom 80L Design ControlDocument p) For any location, at any time "t", the total acceleration response is given by (refer to Figure 3.7B-6): (J w X(total) = X(x). + X(y) + X(z) Y(total) = Y(x) + Y(y) + Y(z) Z(total) = Z(x) + Z(y) + Z(z) where, . X(x) is response in X direction due to carthquake excitation in the X direction, X(y) is response in X direction due to earthquake excitation in the Y direction, etc. The above summation applies to both translational and rotational directions. For reference, EW, NS and Vert. directions correspond to global X, Y and Z respectively.

      . Figures 3.7B-58 to -84 show the variation of maximum acceleration with height in each of three typical building areas: Interior structure, Shield Building and Fuel Building. The results from each control motion (CMS 1, or CMS 2, or CMS 3) are plotted separately for each structure.

1.6.3 Computation of Response Spectra

p. The time histories at each location from step 2 above are then used as input to the program RESPEC for

( computation of response spectra for 2% and 5% damping. The response spectra are computed at 107 frequencies, shown below (units of Hertz). These frequencies cover the frequency range of interest and exceed the guidelines of the Standard Review Plan (NUREG-0800). 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.15 3.3 3.45 3.6 3.8 4 4.2 4.4 4.6 4.8 5 5.25 5.5 5.6 5.75 6 6.1 6.15 6.25 6.5 6.75 7 7.25 7.5 7.75 8 8.5 8.75 9 9.5 9.75 10 10.5 10.75 11 11.5 11.75 12.0 12.5 12.75 13.0 13.5 14 14.5 15 15.2 15.5 16 16.5 17. 17.5 18 18.75 19.0 19.5 20 20.5 20.75 21 22 22.5 23 24 25 27.5 30 31 33 35 37 40 50 60 70 80 90 100 1.7 Computation of Element Forces Maximum forces (axial, shear, moments and torsion), at all the sticks are computed in SASSI using the STRESS module. The maximum forces are obtained separately for the horizontal and the vertical models .[V j when subjected to the three input motions CMS 1, CMS 2 and CMS 3 described in Section 2.5.

       .^_ .. : Design nieterlat- Design of SSC                                                          Page 3.7B-7

=

System 80+ Design ControlDocumejn The maximum forces of the sticks across each major elevation are subsequently summed and plotted in Figures 3.7B-85 to 3.7B-90 as a function of elevation. 1.8 Output Locations Output acceleration time histories are obtained at the following building elevations. Results are obtained at the mass points nearest to the building elevation: Building Elevation (ft) Nearest Mass Point Elev. of Mass Point (ft) Interior Structure 50.0 501 50.0 91.75 506 88.09 210.0 529 208.122 Shield Building 263.5 644 263.5 Steel Cont. Vessel 250.97 460 250.97 Fuel building 70.0 3 69.527 104.0 9 103.688 146.0 18 145.386  ! 170.0 21 169.092 CVCS/Maint. Area 70.0 103 68.067 115.5 112 113.762 170.0 121 168.613 Diesel Gen. Area 1 91.75 205 91.696 Diesel Gen. Area 2 91.75 305 91.6 % EFWl 70.0 703 69.773 130.5 715(H-Model) 131.561 713(V-Model) 156.0 721(H-Model) 154.255 719(V-Model) EFW2 70.0 803 69.773 130.5 815(H-Model) 131.561 813(V-Model) 156.0 821(H-Model) 154.255 819(V-Model) Control Roorn Area 1 70.0 903 69.574 115.5 909 112.028 130.5 912 129.595 Control Roorn Area 2 70.0 1003 69.574 115.5 1009 112.028 130.5 1012 129.595 Appreewd Design Material Desiger of SSC Page 3.78-8

p i Svetem 80+ Desian contrat occument s 1.9 ' Analysis Res,alts i j Representative plots of envelope SSE response spectra are presented in Figures 3.7B-7 to 3.7B-42. The  :

!-                    corresponding locations are at the center of the basemat (Elevation +50'), the top of the IS (Elevation
.                     +210'), the top of the SCV (Elevation +251'), the top of the SB (Elevation +263'), the top story of the   ,

FB (Elevation +170') and the top of the CAA (Elevation +130'). The spectra are raw spectra, i.e. l without broadening and smoothening. Damping ratios are 2 and 5% of criticat Responses from all motions (CMS 1, CMS 2 and CMS 3) are superimposed on the same plot for each location and each-i direction (X, Y, Z). The fixed-base spectra are also superimposed on those plots for completeness of the j . envelope. At the foundation, the horizontal and vertical spectra show a wide scatter of the maximum peaks which l' are mostly accounted by the variations in the free-field motion at the foundation level. At the top of the j SCV, most horizontal spectra have maximum amplitudes in the area near 5 Hz, while for the vertical the spectra have maximum amplitudes in the 10-20 Hz range. However, in the IS horizontal spectra, most. 4 of the peaks are concentrated in the 3-6 Hz and the 9-20 Hz ranges reflecting amplification due to rocking

                  - of the structure (at the low frequency range) and amplification near the main horizontal natural            ,
                  ' frequencies of the IS (at the high frequency range). Similar behavior to that of the IS are also observed
                  . for the top of the other areas such as the FB and CAA.                                                         j 1

The range of site parameters used in the SSI analyses cover a broad range of site conditions. Soil amplification occurs at frequencies in the range of dominant structural modal frequencies. Therefore, resonance effects between the soil and the structures are captured in the SSI analyses and they are reflected in the results. As such, the combined SSI results ensure that adequate seismic loads for the i _ System 80+ NI structures have been generated for sites that are compatible with the generic sites used in these analyses. l 2.0 Parametric Studies j 2.1 Soil Property Variation Study l- A sensitivity study on the effect of soil property variation is performed by changing the low-strain shear modulus by +100% (i.e. 2 times) for the upper bound (UB) case. The base case is arbitrarily taken as soil case B3.5. . Acceleration response spectra are computed at selected points. These are then compared - with corresponding spectra from the base case. , l Figures 3.1B-43 and 3.7B-44 show the response spectra at the top of the SCV and the top of the IS. It l is shown that if the Upper Bound case is compared to the envelope of all the soil cases (Figures 3.7B-22 and 3.7B-16), it is completely covered by other soil cases. The effects of soil variation are accounted for by considering the wide range of soil profiles that were used in the System 80+ analyses. 2.2 Effects of Cracked Concrete L Since the System 80+ structures are subjected to a 0.3g SSE, the dynamic response of the superstructure could be impacted by (potential) concrete cracking due to the imposed seismic and other loads. To evaluate concrete cracking effects, three analyses of the superstructure are performed considering cracked concrete properties. The selected cases are A-1, B-3.5 and the fixed base with concrete backfill, with ,s the CMS 2 motions. The fixed base case with concrete backfill and the A-1 case result in the highest spectral peaks at the top of the IS with uncracked concrete. Therefore, they are critical cases for i AmwomiDesgn AfseenW Dosyv er SSC page 178-9

System 80+ Design ControlDocument maximum response on concrete structures and, as such, they are analyzed with cracked concrete properties. The Nuclear Island structures consist of areas with short and thick walls, which are typically shear-resisting elements. It P assumed that concrete cracking results in a 30% reduction in stiffness. The superstructure stick models are modified by reducing the Modulus of Elasticity (E) by 30% for all concrete elements. The results of these analyses are shown in Figures 3.7B-45 to 3.7B-53. Figure 3.7B-45 shows the shift that cracked concrete properties produces to the dominant spectral peak with soil case B-3.5. This shift is accompanied by a small amplitude increase. The spectral amplitude with cracked concrete properties is Sg at approximately llhz, which is well below the envelope of all soil cases at that location (refer to Figure 3.7B-48 for the envelope). Therefore, case B-3.5 is not critical. Figures 3.7B-46 through 3.7B-53 present the envelopes of all soil and fixed-base cases with uncracked t concrete properties at four locations. Superimposed on these envelopes are the cracked concrete cases i A-1 and fixed base with concrete backfill (shown with thick solid lines). For the four location shown in Figures 3.7B-46 to 3.7B-53, the spectra with cracked concrete do not affect the envelopes. Based on the above, it is concluded that the effects of concrete cracking on the design forces and response spectra from the SSI analyses is enveloped by the design parameters. 2.3 Effects of Debonding In all the SSI analyses, the foundation is assumed to be fully bonded with the side soil, i.e. no separation or gaps exist between the rigid subsurface cavity and the site soil. To evaluate the effects of potential debonding of the foundation with the soil, a parametric SSI analysis with SASSI was performed using soil case B3.5 with control motion CMS 2. In this parametric analysis, the foundation cavity is assumed to have debonded from the side soil at the top 20 feet of the embedment area. The results of this analysis are shown in Figures 3.7B-54 and 3.7B-55. Figure 3.7B-54 shows that, at the top of the SCV, the response spectra are similar to the bonded case, with the exception of a slight shift in the fundamental peak. This shift will be adequately covered when il5% broadening will be applied to the raw spectra for design purposes. At the top of the IS, the fundamental peak is not shifted and it is reduced in amplitude. The increase in amplitude in the low frequency range is adequately covered by other soil cases. Therefore, since a wide range of soil cases are considered in the original analyses, debonding does not affect the envelope of the seismic forces and spectra computed with the fully bonded case. 2.4 Connection of Subsurface Slabs to Exterior Walls In the SSI analyses with SASSI, the impedances are computed based on the assumption of a rigid foundation. Therefore, the impedances are computed at a single point, which is selected at the center of the basemat. The superstructure model is then connected at this point to obtain the complete SSI system. In the superstructure model, the exterior walls are modeled with actual concrete properties. This is generally considered conservative, since the exterior walls, as part of the superstructure model, may deform more without the additional local restraint of the side soil. To address NRC's concern of the potential of a dynamic effect on the in-structure response spectra due to the local connection of the (axially rigid) subsurface slabs with the exterior walls and the soil, a parametric study is performed considering this local connection. Apsproved Design Material. Design of SSC Page 3.78-10

l l Syztem 80+ Design ControlDocument 7 (") Soil case 83.5 with the CMS 2 motion is selected as the case study, since it is one of the controlling cases at various in-stmeture locations. The superstructure model is modified, so that the subsurface portions of the sticks are laterally rigid. Therefore, the entire subsurface cavity together with the superstructure model is laterally rigid. The results are shown in Figures 3.7B-56 and 3.7B-57. The in-structure response spectra at the top of the IS and the SCV using the rigid subsurface model are completely enveloped by the original spectra. Therefore, it is concluded that the modeling of the superstructure connection with the foundation is conservative. The original spectra will still be used in the seismic design of the System 80+. 3.0 Two-Dimensional SSI Analyses Two-dimensional SSI analyses of the System 80+ Nuclear Island (NI) were performed to determine the following:

1. Dynamic lateral pressures on the embedded walls of the NI,

2. The effect of Structure-Soil-Structure Intermion (SSSI) between the NI and other Category I structures that do not share a common foundation with the NI, and

3. The bearing pressure distribution resulting from earthquake loading at the soil / mat interface.

Separate 2-D SSI models were created for the N-S and E-W directions. Figures 3.7B-103 and 3.7B-104 O V show the N-S and E-W models, respectively. In these models the NI superstructure was modeled with lumped mass and stiffness properties, representing a one foot slice through the center of the NI. The stiffness properties of the 2-D NI have been adjusted so that the (fixed-base) response matches the 3-D (fixed-base) response. The foundation cavity, which was modeled as rigid in the 3-D analysis, was modeled in the 2-D analysis with rigid beam elements. The soil adjacent to the foundation cavity was modeled with plane strain soil elements, from which mid-element soil pressures were obtained. A separate model was created for assessing the sidewall pressures in the E-W direction. That model includes the Turbine Building immediately to the East of the N1, as shown in Figure 3.7B-105. The Turbine Building was represented in hunped parameter fashion with separate sticks and foundations for the pedestal and for the surrounding building. The Radwaste Building, immediately to the West of the NI, was not included in this 2-D E-W model because its foundation mat is at approximately the same embedment depth as the NI's. The evaluation of dynamic pressure under the Basemat considered 6 soil cases: A-1., B-1, B-1.5, C-1, C-2, and Rock with concrete backfill. Based on a review of the SSI response profiles from the 3-D SSI analyses, these soil cases weie chosen as representative cases that would provide an envelope of the desired response parameters. All analyses were performed subject to the three control motions CMSI, CMS 2, and CMS 3. Figures 3.7B-106 through 3.7B-108 show the maximum bearing pressure distribution for the soil / mat interface under the basemat due to the horizontal excitation for each of the 3 control motions for the N-S model. Figures 3.7B-109 through 3.7B-llI show corresponding pressure plots due to vertical excitation of the N-S model for each of the three control motions. Figures 3.7B-112 through , 3.7B-114 show the maximum pressure distributions under the mat due to horizontal excitation of the E-W model, and Figures 3.7B-115 through 3.7B-ll7 show the maximum pressure distributions for the E-W model due to vertical excitation. (nV) Approved Design A4aterial- Design of SSC Page 3.78-11

System 80+ Design controlDocument _ For the evaluation of dynamic sidewall pressures,7 soil cases were considered: A-1, B-1, B-1.5, B-3.5, C-1. C-2, and Rock with concrete backfill. The soil cases are the same ones considered for the basemat evaluations, with the addition of soil B-3.5. This case was included in order to represent the enveloping site condition with impedance mismatch. Figures 3.7B-118 through 3.7B-123 show the maximum dynamic pressure on the embedded NI walls for the N-S model. The lateral pressures due to horizontal input and due to vertical input motions are shown separately for each of the three control motions. The pressure distributions are an envelope at each elevation of the pressures on the north and south walls. Figures 3.7B-124 through 3.7B-135 show the maximum dynamic pressure distributions for the E-W model. The east and west sides are shown separately d e to the potential surcharge effects resulting from the presence of the Turbine Building on the east side. The basemat resultant maximum pressures due to SRSS combination of horizontal and vertical input motions, acting along a diagonal line across the basemat are shown in Figures 3.7B-136 through 3.7B-138 for the three control motions. The maximum pressure from all cases is due to the CMSI motion for the case of Rock with concrete backfill. When superimposed with the net effect of dead load minus the maximum potential buoyancy effects (corresponding to a downward uniform pressure of 4,800 psf), the maximum uplift configuration in any corner is determined to be a triangle with side dimensions of 146'. This postulated uplift condition is shown in Figure 3.7B-139. This uplift area is subsequently modeled in the 3-D finite element static model of the NI which is used to predict the maximum stresses used for design of the basemat. Figures 3.7B-140 through 3.7B-142 show the envelope of dynamic lateral pressures for all soil cases and all control motions on the embedded walls of the NI for the N-S and E-W directions. The lateral pressures due to horizontal and vertical excitation have been combined by SRSS. These pressure distributions are further increased to consider an additional effect. This is the 3-D torsional effect which is ignored in a 2-D representation. The maximum increase for torsional effects (over all soil cases and all motions) is 29%. It is conservatively applied to the SRSS combination of the pressures due to horizontal and vertical input metions. The resulting pressure distributions due to seismic loading are then combined with static pressures to determine the pressures used for design. In order to determine the effects of SSSI between the Ni and adjacent structures, separate SSI analyses were performed incorporating representation of both the N1 and adjacent structures. In this fashion, the effects of structure-soil-structure interaction is explicitly accounted for. These analyses are further described in Appendix 3.7C. O Asywoved Design Material Onign of SSC Page 3.78-12

l l System 80+ _ oeston controloocument l [ Table 3.7B-1 Seismic Analysis Cases v Seismic Analysis Cases Case SSE Fixed-Base (without backfill) Yes  ! Fixed-Base (with concrete backfill) Yes Al Yes B1 Yes Bl.5 Yes B2 Yes B3.5 Yes B4 Yes C1 Yes C1.5 Yes C2 Yes C3 Yes i Analysis Cutoff Frequencies Horizontal Vertical Case Cutoff (Hz) Cutoff (Hz) ' Fixed-Base 40 40 ' (w/out backfill) Fixed-Base 40 40 > (w/ concrete backfill) At 40 40 , B1 40 40 t Bl.5 40 40 ft2 40 40 33.5 18 40 B4 22 36 d 40 40 C1.5 26 26 C2 12 25 C3 11 18 i O 4prowd outen neateriel Dnion of SSC  !*9e 17813 t

l System 80+ Design Control Document j Table 3.7B-2 Soil Layers and Properties

~

h! Case Al Layer Thictmess S-Wave P-Wave Damping Damping  ! No. (ft.) Velocity Velocity S-Wave P-Wave (ft/sec) (ft/sec) 1 4.75 1774 4345 0.008 0.003 2 5 1847 4522 0.012 0.004 3 5 1880 4585 0.017 0.006 4 5 1880 4585 0.017 0.006 5 5 1913 4666 0.022 0.007 6 5 1913 4666 0.022 0.007 7 5 1949 4754 0.025 0.008 8 5 1949 4754 0.025 0.008 9 5 1994 4867 0.027 0.009 10 $ 1994 4867 0.027 0.009 11 2 1994 4867 0.027 0.009 Table 3.7B-3 Soil Layers and Properties Case B1 Layer Thickness S-Wave Velocity P-Wave Velocity Damping Damping No. (ft.) (ft/sec) (ft/sec) S-Wave P-Wave , 1 4.75 1774 4345 0.008 0.003 2 5 , 1846 4522 U.012 0.004 ) 3 5 1872 4585 0.018 0.006 4 5 1872 4585 0.018 0.006 5 5 1905 4666 0.023 0.008 l 6 5 1905 4666 0.023 0.008 j 7 5 1941 4754 0.026 0.009 ' 8 5 1941 4754 0.026 0.009 9 5 1987 4867 0.027 0.009 10 5 1987 4867 0.027 0.009 11 2 1987 4867 0.027 0.009 12 3 2032 4977 0.029 0.010 13 5 2032 4977 0.029 0.010 14 5 2072 5075 0.031 0.010 15 5 2072 5075 0.031 0.010 16 5 2072 5075 0.031 0.010 17 5 2072 5075 0.031 0.010 18 5 2143 5249 0.033 0.011 19 5 2143 5249 0.033 0.011 20 5 2143 5249 0.033 0.011 21 5 2143 52;9 0.033 0.011 Unit Weight (all layers) = 125 pcf Approved Design Material- Design of SSC Page 3.78-14

Sy~ tem 80 + Design ControlDocument I Table 3.7B-4 Soil Layers and Properties Case Bl.5 Thickness S-Wave P-Wave Layer No. (ft.) Velocity (ft/sec) Velocity (ft/sec) Damping S-Wave Damping P-Wave 1 4.75 1405 3440 0.010 0.003 2 5 1399 3426 0.019 0.006 3 5 1399 3427 0.027 0.009 4 5 1399 3427 0.027 0.009 - 3394 0.036 0.012 5 5 1386 l 6 5 1386 3394 0.036 0.012 7 5 1384 3390 0.041 0.014 8 5 1384 3390 0.041 0.014 9 5 1392 3410 0.045 0.015 10 5 1392 .'110 0.045 0.015 11 2 1392 3410 0.045 0.015 12 3 1412 3460 0.047 0.016 j 13 5 1412 3460 0.047 0.016 14 5 1461 3579 0.047 0.016 i I 15 5 1461 3579 0.047 0.016 16 5 1461 3579 0.047 0.016 17 5 1461 3579 0.047 0.016 18 5 1514 3709 0.049 0.016 19 5 1514 3709 0.049 0.016 20 5 1514 3709 0.049 0.016 l 21 5 1514 3709 0.049 0.016 Unit Weight (all layers) = 125 pcf i l l l 1 Annrosed Deciptr Mosend Desiprs of SSC Page 3.7815

i Syntem 80+ Design controlDocument j I I Table 3.7B-5 Soil Layers and Propert'.es  ! Case B2 S-Wave Velocity P-Wave Damping Layer No. Thickness Velocity (ft/sec) (ft/sec) S-Wave Damping P-Wave (ft.) 1 4.25 996 2440 0.013 0.004 2 4.5 973 2383 0.025 0.008 3 4.5 935 2290 0.039 0.013 4 4.5 935 2290 0.039 0.013 5 4.5 915' 2241 0.048 0.016 6 4.5 915 2241 0.048 0.016 7 4.5 915 2241 0.048 0.016 8 4.5 908 2224 0.055 0.018 9 4.5 908 2224 0.055 0.018 10 4.5 909 2227 0.062 0.021 11 4.5 909 2227 0.062 0.021 12 2.5 909 2227 0.062 0.021 13 4.5 909 2227 0.067 0.022 14 4.5 909 2227 0.067 0.022

                      ~

15 4.5 916 2244 0.072 0.024 16 4.5 916 2244 0.072 0.024 17 4.5 916 2244 0.072 0.024 18 4.5 916 2244 0.072 0.024 19 4.5 945 2315 0.075 0.025 20 4.5 945 2315 0.075 0.025 21 4.5 945 2315 0.075 0.025 22 4.5 945 2315 0.075 0.025 23 3 945 2315 0.075 0.025 Unit Weight (all layers) = 125 pcf 9 Approwd Design Material- Design of SSC Pope 3.78-16

System 80+ Design ControlDocument

   /3 g     Table 3.78-6 Soil Layers and Properties Case B3.5 Layer No. Thickness      S-Wave Velocity P-Wave Velocity Damping S-Wave            Damping P-Wave (ft.)            (ft/sec)      (ft/sec) 1            2.75                567          1389          0.026                           0.009 2               3                567          1389          0.026                           0.009
                                                                                                                                       ]

3 3 503 1232 0.050 0.017 , 4 3 503 1232 0.050 0.017 5 3 445 1090 0.080 0.027 1 6 3 445 1090 0.080 0.027 7 3 445 1090 0.080 0.027 8 3 389 953 0.109 0.036 9 3 389 953 0.107 0.036 10 3 389 953 0.109 0.036 J 11 3 3 94 %5 0.I12 0.037 j 12 3 394 %5 0.I12 0.037 13 3 394 %5 0.112 0.037 14 3 411 1007 0.110 0.037 FD 0.110 0.037 Q 15 3 411 1007 16 3 411 1007 0.I10 0.037 17 2 4?! 1007 0.110 0.037 18 2 411 1007 0.110 0.037 19 8 1301 3187 0.036 0.012 20 8 1328 3253 0.035 0.013 21 8 1328 3253 0.038 0.013 22 8 1328 3253 0.038 0.013 23 8 1363 3339 0.041 0.014 24 8 1363 3339 0.041 0.014 Unit Weight (all layers) = 125 pcf f

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AMweved Desiger Noterial Desiers of SSC Page 3.7817

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System 80+ Design controlDocument I Table 3.7B-7 ScP Layers and Properties h ! Case B4 Layer No. Thickness S-Wave Velocity P. Wave Velocity Damping S-Wave Damping P-Wave (ft.) (ft/sec) (ft/sec) 1 3.75 473 1159 0.025 0.008 ) 2 4 421 1031 0.050 0.017 3 4 421 1031 0.050 0.017 __ 4 3.75 374 916 0.079 0.026 5 3.75 374 916 0.079 0.026 6 3.4 340 833 0.099 0.033 7 3.4 340 833 0.099 0.033 8 3.4 340 833 0.099 0.033 9 3.3 329 806 0.I12 0.037 10 3.3 329 806 0.112 0.037 11 3.3 329 806 0.I12 0.037 12 3.1 310 759 0.128 0.043 13 3.1 310 759 0.128 0.043 14 3.1 310 759 0.128 0.043 15 3.1 310 759 0.128 0.043 16 8 2124 5203 0.019 0.006 17 10 2181 5342 0.021 0.007 18 10 2181 5342 0.021 0.007 19 10 2264 5546 0.022 0.007 20 10 2264 5546 0.022 0.007 Unit Weight (all layers) = 125 pcf 9 Approved Design Afsterial- Design of SSC Page 3.7818

Sy" tem 80+ Design ControlDocument , (A) Table 3.7B-8 Soil Layers and Properties Case C1 Layer No. Thickness S-Wave Velocity P-Wave Velocity Damping S-Wave Damping P-Wave (ft) (ft/sec) (ft/sec) 1 4.75 1770 4335 0.008 0.003 2 7 1844 4518 0.013 0.004 3 8 1866 4571 0.0) 8 0.006 4 8 1901 4656 0.023 0.008 5 8 1937 4744 0.026 0.009 6 8 1937 4744 0.026 0.009 7 8 1976 4839 0.029 0.010 8 8 2001 4902 0.031 0.010 9 10 2046 5011 0.034 0.011

10 10 2046 5011 0.034 0.011 11 10 2138 5236 0.034 0.011 12 10 2138 5236 0.034 0.011 13 10 2310 5658 0.032 0.011 14 10 2310 5658 0.032 0.011 15 10 2266 5550 0.036 0.012 16 10 2266 5550 0.036 0.012 17 10 2359 5778 0.036 0.012 18 10 2359 5778 0.036 0.012 19 10 2450 6001 0.036 0.012 20 10 2450 6001 0.036 0.012 21 10 2424 5938 0.038 0.013 22 10- 2424 5938 0.038 0.013 Unit Weight (all layers) = 125 pcf i

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System 80+ Design ControlDocurnent Table 3.7B-9 Soil Layers and Properties Case C1.5 Layer No. Thickness S-Wave Velocity P-Wave Velocity Damping S-Wave Damping P-Wave (ft.) (ft/sec) (ft/sec) 1 5.75 997 2443 0.012 0.004 2 6 980 2400 0.024 0.008 3 6 961 2353 0.034 0.011 4 6 961 2353 0.034 0.011 5 6 947 2319 0 M3 0.014 6 6 950 2327 C M7 0.016 7 6 950 2327 0.047 0.016 8 6 970 2376 0.049 0.016 9 4 970 2376 0.049 0.016 10 8 976 2391 0.052 0.017 11 8 996 2440 0.056 0.019 12 9 996 2440 0.056 0.019 13 9 1039 2545 0.056 0.019 14 10 1039 2545 0.056 0.019 15 10 1130 2767 0.053 0.018 16 12 1130 2767 0.053 0.018 17 12 1093 2678 0.060 0.020 , 18 14 1093 2678 0.060 0.020 19 14 1153 2823 0.057 0.019 20 14 1207 2956 0.056 0.019 l 21 14 1207 2956 0.056 0.019 22 14 1201 2941 0.057 0.019 Unit Weight (all layers) = 125 pcf l 1 l 4 Oll l AS4 roved Design Material- Desigst of SSC Page 3.7B-20

Sy' tem 80+ Design ControlDocument . Table 3.7B-10 Soil Layers and Properties Case C2 Layer No. Thickness P-Wave Velocity S-Wave Velocity Damping S-Wave Damping P-Wave (ft.) (ft/sec) (ft/sec) 1 5.75 485 1188 0.020 0.007 2 6 450 1102 0.039 0.013 3 6 409 1002 0.062 0.021 4 6 375 919 0.085 0.028 5 6 375 919 0.085 0.028 6 6 365 894 0.093 0.031 7 6 365 894 0.093 0.031 8 6 365 894 0.097 0.032 9 4 365 894 0.097 0.032 10 6 369 904 0.100 0.033 11 6 388 950 0.097 0.032 17 6 388 950 0.097 0.032 13 6 388 950 0.097 0.032 14 6 388 950 0.097 0.032 15 7 400 980 0.100 0.033 b 16 8 400 980 0.100 0.033 l 17 9 481 1178 0.083 0.028-18 9 481 1178 0.083 0.028 19 9 481 1178 0.083 0.028 20 9 469 1149 0.087 0.029 21 9 469 1149 0.087 0.029 l l 1 22 9 495 1212 0.085 0.028 23 9 495 1212 0.085 0.028 24 10 515 1261 0.085 0.028 25 10 515 1261 0.085 0.028 26 10 487 1193 0.093 0.031 27 10 487 1193 0.093 0.031 Unit Weight (all layers) = 125 pcf Apnnrowed Desigru niereria!- Desiger of SSC Pope 3.78-21

System 80+ Design ControlDocument Table 3.7B-11 Soil Layers and Properties Case C3 Layer No. Thickness S-Wave Velocity P-Wave Velocity Damping S-Wave Damping P-Wave (ft.) (ft/sec) (ft/sec) 1 4.25 478 1171 0.023 0.008 2 4.5 433 1061 0.045 0.015 3 4.5 386 946 0.073 0.024 4 4.5 386 946 0.073 0.024 5 4.5 386 946 0.073 0.024 6 3.5 357 874 0.092 0.031 7 3.5 357 874 0.092 0.031 8 3.5 354 868 0.097 0.036 9 3.5 354 868 0.097 0.036 10 3.5 354 868 0.097 0.036 11 3.5 348 852 0.107 0.036 12 3.5 348 852 0.107 0.036 13 2.5 348 852 0.107 0.036 14 2.5 348 852 0.107 0.036 15 3.5 357 876 0.107 0.036 16 3.5 357 876 0.107 0.036 17 4 371 908 0.107 0.036 18 4 371 908 0.107 0.036 19 5 371 908 0.107 0.036 20 5 371 908 0.107 0.036 21 6 371 908 0.107 0.036 22 7 394 %5 0.104 0.035 23 10 394 965 0.104 0.035 24 25 2500 6123 0.015 0.005 25 25 2461 6028 0.019 0.006 26 25 2539 6220 0.021 0.007 27 25 2626 6433 0.022 0.008 Unit Weight (r.Il layers) = 125 pcf 9 Approved Design htsterial. Design of SSC Page 3.78-22

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Syatem 80+ Design controlDocument O Soll Case U 5-Wave Excitation S-Wave Excitation p-Wave Excitation O-180 Direction 90-270 Direction vertical Direction (X) (Y) (Z) U l I V Nodal Response Nodal Response Nodal Response Acceleration Acceleration Acceleration Time Histories Time Histories Time Histories due to due to . Que to x-Excitation y-Excitation z-Excitation X(x) X(y) X(z) Y(x) Y(y) Y(z) Z(x) Z(y) Z(z) XX(x) XX(y) XX(Z) YY(x) YY(y) YY(z) ZZ(x) ZZ(y) ZZ(z) O U Add Nodal Response Acceleration Time Histories < in Time Domain X(x) + X(y) + X(z) - X(combined: Y(x) + Y(y) + Y(z)

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l B Soil Structure Interaction Analysis Methodology f and Results , i Other Seismic Category I Str9ctures f Diesel Fuel Storage Structure and i Component Cooling Water Heat Exchanger Building u l Contents Page p 1.0 Introduction ................................. . . . . . . . . . . . 3.7C- 1 ) Q 2.0 3.0 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Building Models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7C-1 3.7C-1 4.0 Generation of Responses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7C-2 Figures 3.7C-1 Coupled SASSI Soil-Structure-Interaction Model; Nuclear Island and Diesel  ! Fuel Storage Structure, North-South Direction . . . . . . . . . . . . . . . . . . . 3.7C-3 - 3.7C-2 SASSI Soil-Structure-Interaction Model; Diesel Fuel Storage Structure, East-West Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7C-4 3.7C-3 Coupled SASSI Soil-Structure-Interaction Model; Nuclear Island and Component Cooling Water Heat Exchanger Building, North-South Direction . 3.7C-5 3.7C.4 SASSI Soil-Structure-Interaction Model; Component Cooling Water Heat  ; Exchanger Building, East-West Direction . . . . . . . . . . . . . . . . . . . . . . 3.7C-6  ! 3.7C-5 Response Spectra at Elev. 73.75' of CCW, All Soil Cases, All Motions, I Horizontal Model, H1 Motion (E-W), 2% Damping . . . . . . . . . . . . . . . . 3.7C-7  ! 3.7C-6 Response Spectra at Elev. 91.75' of CCW, All Soil Cases, All Motions, i Horizontal Model, HI Motion (E-W), 2% Damping . . . . . . . . . . . . . . , . 3.7C-8 , 3.7C-7 Response Spectra at Elev.111.75' of CCW, All Soil Cases, All Motions,

                    ' Horizontal Model, HI Motion (E-W), 2% Damping . . . . . . . . . . . . . . . . 3.7C-9                               ;

3.7C-8 Response Spectra at Elev. 73.75' of CCW, All Soil Cases, All Motions, j Horizontal Model, H2 Motion (N-S),2% Damping . . . . . . . . . . . . . . 3.7C-10 i G 3.7C-9 Response Spectra at Elev. 91.75' of CCW, All Soil Cases, All Motions,  ; V Horizontal Model, H2 Motion (N-S),2% Damping . . . . . . .. . . . . . . . . 3.7C-11 l I l 1 4preM De@ neeennel Doelpur of SSC Page R l j

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System 80+ Design ControlDocument Figures (Cont'd.) Page 3.7C-10 Response Spectra at Elev.111.75' of CCW, All Soil Cases, All Motions, Horizontal Model, H2 Motion (N-S),2% Damping . .. . . . . 3.7C-12 3.7C-11 Response Spectra at Elev. 73.75' of CCW, All Soil Cases, All Motions, Vertical Model,2% Damping .... ..... .................. 3.7C-13 3.7C-12 Response Spectra at Elev. 91.75' of CCW, All Soil Cases, All Motions, Venical Model,2% Damping ... . .. . ................. 3.7C-14 3.7C-13 Response Spectra at Elev. I11.75' of CCW, All Soil Cases, All Motions, Vertical Model,2% Damping .............. ......... ... 3.7C-15 3.7C-14 Response Spectra at Elev. 78.25' of DFSS, All Soil Cases, All Motions, Horizontal Model, HI Motion (E-W),2% Damping . . . . ........ 3.7C-16 3.7C-15 Response Spectra at Elev.103.25' of DFSS, All Soil Cases, All Motions, Horizontal Model, H1 Motion (E-W), 2% Damping . . ............ 3.7C-17 3.7C-16 Response Spectra at Elev. 78.25' of DFSS, All Soil Cases, All Motions, i Horizontal Model, H2 Motion (N-S),2% Damping . . . . . . ..... . 3.7C-18 3.7C-17 Response Spectra at Elev.103.25' of DFSS, All Soil Cases, All Motions, l Horizontal Model, H2 Motion (N-S),2% Damping ....... ..... 3.7C-19 3.7C-18 Response Spectra at Elev. 78.25' of DFSS, All Soil Cases, All Motions, Venical Model,2% Damping . ... . .... .. ... . .. 3.7C-20 . 3.7C-19 Response Spectra at Elev.103.25' of DFSS, All Soil Cases, All Motions, f Vertical Model,2% Damping . ......... .. . ... ... . 3.7C-21 l 3.7C-20 Response Spectra at Elev. 73.75' of CCW, All Soil Cases, All Motions, l Horizontal Model, H1 Motion (E-W), 5 % Damping . . . ... . .. .. 3.7C-22 l 3.7C-21 Response Spectra at Elev. 91.75' of CCW, All Soil Cases, All Motions, i Horizontal Model, H1 Motion (E-W), 5% Damping . . . . . . .. . ... 3.7C-23 l 3.7C-22 Response Spectra at Eley,111.75' of CCW, All Soil Cases, All Motions, I Horizontal Model, H1 Motion (E-W), 5 % Damping . .... . .. 3.7C-24 l 3.7C-23 Response Spectra at Elev. 73.75' of CCW, All Soil Cases, All Motions, I Horizontal Model, H2 Motion (N-S),5% Damping . . . . . ... . 3.7C-25 3.7C-24 Response Spectra at Elev. 91.75' of CCW, All Soil Cases, All Motions, Horizontal Model, H2 Motion (N-S),5% Damping . . . .. . ... 3.7C-26 3.7C-25 Response Spectra at Elev.111.75' of CCW, All Soil Cases, All Motions, Horizontal Model, H2 Motion (N-S), 5% Damping . . .. .. 3.7C-27 3.7C-26 Response Spectra at Elev. 73.75' of CCW, All Soil Cases, All Motions. Venical Model,5% Damping .... .. . .. ... .. . 3.7C-28 3.7C-27 Response Spectra at Elev. 91.75' of CCW, All Soil Cases, All Motions, Venical Model,5% Damping .... ... ........ ... . . 3.7C-29 3.7C-28 Response Spectra at Elev.111.75' of CCW, All Soil Cases, All Motions, Venical Model,5% Damping ... ....... . ...... . .. 3.7C-30

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3.7C-29 Response Spectra at Elev. 78.25' of DFSS, All Soil Cases, All Motions, Horizontal Model, H1 Motion (E-W), 5% Damping . . . ...... . 3.7C-31 3.7C-30 Response Spectra at Elev.103.25' of DFSS, All Soil Cases, All Motions, Horizontal Model,111 Motion (E-W), 5% Damping . . . . . .. ... 3.7C-32 3.7C-31 Response Spectra at Elev. 78.25' of DFSS, All Soil Cases, All Motions, Horizontal Model, H2 Motion (N-S),5% Damping . . .... ........ 3.7C-33 3.7C-32 Response Spectra at Elev.103.25' of DFSS, All Soil Cases, All Motions, Horizontal Model, H2 Motion (N-S), 5 % Damping .... .. . ... 3.7C-34 Approved Design Materke! Design of SSC Pagetv

i System 80+ Deslan ControlDocument  : t Figures (Cont'd.) Page 3.7C-33 Response Spectra at Elev. 78.25' of DFSS, All Soil Cases, All Motions,  ; 3.7C-35

                                      . Vertical Model,5% Damping ........................ ....

3.7C Response Spectra at Elev.103.25' of DFSS, All Soil Cases, All Motions, Vertical Model,5% Damping ............ ...,............ 3.7C-36 t 4 t i e I J l l I l 1 I Appresent Deate neeenrint - Dee@n of SSC Page v .--_~_ . . . ~ , - , . - - . , _ . _ , . , . , .I

System 80+ Design controlDocument I . 1.0 Introduction This appendix describes the soil-structure-interaction (SSI) methodology used to evaluate the seismic responses for the Other Seismic Category I Structures of the System 80+ Standard Design. These stmetures include the Diesel Fuel Storage Structures (DFSS) and Component Cooling Water Heat Exchanger Buildings (CCWHEB) which have been evaluated. These two buildings are located near the Nuclear Island. The Nuclear Island is a very massive structure. Therefore, the SSI analyses of the DFSS and the CCWHEB incorporate the Soil-Structure-Soilinteraction effects of the Nuclear Island. The SSI analyses are performed for all the generic sites used in the SSI analyses of the Nuclear Island (described in Appendix 3.7B) and the three Safe Shutdown Earthquake (SSE) control motions CMS 1, CMS 2, and CMS 3. 2.0 Methodology The methodology used in the SSI analyses of the DFSS and the CCWHEB is based on the methodology of the computer code SASSI. SASSI is the computer code that was also used for the SSI analyses of the Nuclear Island. The two stmetures are essentially symmetrical about their center lines, with insigni'icant eccentricities between the centers of mass, and vertical and horizontal centers of rigidity. Therefou, the seismic results can be obtained using two-dimensional SSI analyses. Two two-dhensional SASSI analyses are performed for each structure. One SASSI analysis is for a two-dimensional model of a slice through the Nuclear Island, along the N-S axis, and either the DFSS or the CCWHEB. This model includes both the Nuclear Island structure model coupled through the soil elements with either the DFSS /" or the CCWHEB model (Figures 3.7C-1 and 3.7C-3, respectively). This model is utilized to compute (3), responses due to the N-S and Vertical components of the three control motions CMS 1, CMS 2, and CMS 3. The other SASSI analysis is for a slice through either the DFSS or the CCWHEB, along the E-W axis (Figures 3.7C-2 and 3.7C-4, respectively). This model is used to compute the responses due to the E-W component of the three control motions. The DFSS and the CCWHEB are both modeled as embedded structures. The DFSS have embedments of 13.5 feet below grade. The CCWHEB have embedments of 18.0 feet below grade. The foundations cf each structure are modeled as rigid. A layer of soil elements is added around the perimeter of the rigid foundation of both structures in order to obtain dynamic soil pressures directly from the SSI analyses. In this manner the interaction effects from the Nuclear Island can be determined and incorporated into the design of the exterior walls located below grade. For the rock case, the two buildings are analyzed as " fixed-base" structures with no embedment. The

      " fixed-base" analyses are based on the three-dimensional lumped mass " stick" models of the structures, with simultaneous application of the three components of earthquake motion (i.e., N-S, E-W, and Vertical). The " fixed-base" analyses are performed for the three rock control motions, CMS 1, CMS 2, and CMS 3. The computer code SAP 90 is used in the " fixed-base" analyses to determine structure responses.

3.0 Building Models The dynamic models of the DFSS and CCWHEB are lumped mass " stick" models supported on a fixed-p base (rock) or soil. The models consist of lumped masses at each floor level, consisting of the mass of the floor, half the mass of the walls above and below, weight of equipment plus 25% of the specified live () Apemme onw neww- ony or ssc rege 3.7c-s

l 1 System 80+ Design ControlDocument load. The lumped masses are connected by linear elastic beam type elements which account for coupling l between the horizontal and vertical components of motion. Damping for the SSE is specified as 7% of  ! critical for both building models. General Arrangement of the DFSS and CCWHEB are provided in Section 1.2.16.7 and 1.2.16.8, respectively. l Figures 3.7C-1 throy a 3.7C-4 show schematics of the coupled SSI model of the Nuclear Island and each f of the DFSS and CC'GIEB, respectively. 4.0 Generdion of Responses The response acceleration time histories from the fixed-base and SSI analyses are used as input to the program RESPEC to compute in-structure response spectra for two damping values: 2% and 5%. For the SSI analyses, the response acceleration of each structure in the N-S direction is obtained by algebraically adding in time the N-S response acceleration time history due to N-S excitation and the N-S response acceleration time history due to vertical excitation. The response acceleration in the E-W direction is due to the E-W excitation only. The response acceleration in the vertical direction is obtained by algebraically adding in time the vertical response acceleration time history due to vertical excitation and the vertical response acceleration time history due to N-S excitation. For the fixed-base analyses, the E-W, N-S, and vertical response acceleration time histories include the effects from all three directions, since the control motions in the E-W, N-S, and vertical are applied simultaneously. Figures 3.7C-5 through 3.7C-13 and 3.7C-14 through 3.7C-19 are envelopes of the 2% damping in-structure response spectra, for all soil cases and the three control motions, in the N-S, E-W, and Vertical directions at the base and roof of the CCWHEB and DFSS, respectively. Figures 3.7C-20 through 3.7C-28 and 3.7C-29 through 3.7C-34 are the 5 % damping curves for the CCWHEB and DFSS, respectively. O AMvoved Design Mateniet Deslyn of SSC Page 3.7C-2

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