ML20065G866

to Nonproprietary Design Assessment Rept

ML20065G866
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Site:	Limerick
Issue date:	09/30/1982
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LIMERICK GENERATING STATION UNITS 1 & 2 DESIGN ASSESSMENT REPORT REVISION 1 PAGE CHANGES m

The attached pages, tables, and figures are considered part of a controlled copy of the Limerick Generating Station DAR. This material should be incorporated into the DAR by following the instructions below.

REMOVE INSERT

, VOLUME 1 Pages S-i, -iii, -iv, -v Pages S-i, -iii, -iv, -v Pages 1.4-1 6 -2 Pages 1.4-1 & -2 i Figure 1.4-3 Figure 1.4-3 Page 2-i Page 2-i Pages 2.1-1 & -2 Pages 2.1-1 & -2 Pages 2.2-1 thru -3 Pages 2.2-1 thru -3 Page 4-iii _

Page 4-iii Pages 4.2-5 thru -8 Pages 4.2-5 thru -8

! Table 4.2-4 Table 4.2-4 Tables 4.2-6 6 -7 Tables 4.2-6 & -7 Table 4.2-9 Table 4.2-9 Figure 4.2-16 Figure 4.2-16 I

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N-Pages 5-i & -ii

.Page 5.2-1 Table 5.2-1 Pages 5-i & -ii Page 5.2-1 Table 5.2-1 l Page 5.3-1 Page 5.3-1 Table 5.3-1 Tables 5.3-1 & -2 Page 5.5-1 Page 5.5-1 Pages 5.6-1 & -2 Pages 5.6-1 & -2 Table 5.6-1 Table 5.6-1 Figure 5.6-2 Figure 5.6-2 i Page 5.8-1 Page 5.8-1 Table 5.8-1 Table 5.8-1 Table 5.10-1 (pg 1) Table 5.10-1 (pg 1)

Page 6-i Page 6-i Pages 6.1-1 thru 6.4-1 Pages 6.1-1 thru 6.4-1 Page 6.6-1 Page 6.6-1 1 Page 6.8-1 Page 6.8-1 l

Pages 7-i thru Table 7.1-2 Pages 7-i thru Table 7.1-2 Figures 7.1-3 thru -8 Figures 7.1-3 thru -8 Figures 7.1-11 thru -15 Figure 37.1-11 thru -15 Figure 7.1-17 Figure 7.1-17

'Pages 7.2-1 thru -3 Pages 7.2-1 thru -6 l Figure 7.2-1 Figure 7.2-1 (after tab for Chapter 8) Tab " Response to NRC Questions" j thru tab 600 i VOLUME 2 j

Figure B.1-5 Figure B.1-5 Figures B. 2-71 thru -82 Figures B.2-71 thru -82 Page D-1 Pages D-ii thru Figure D.1-25 8210040272 820930 1 PDR ADOCK 05000352 _1_

! A PDR l_ ___ ._. _ _ _

LGS DAR Revision 1 Page Change Instructions REMOVE INSERT 1 VOLUME 2 Page D-2 Page D.2-1 thru Figure D.2-11 Pages E-i & E-1 Pages E-i thru Figure E.1-29

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LGS DAR

SUMMARY

TABLE OF CONTENTS

( CHAPTER 1 GENERAL INFORMATION 1.1 PURPOSE OF REPORT

1.2 BACKGROUND

1.3 MARK II CONTAINMENT PROGRAM 1.3.1 ~ References 1.4 PLANT DESCRIPTION 1.4.1 Primary Containment 1.4.1.1 Penetrations 1.4.1.2 Internal Structures CHAPTER 2

SUMMARY

2.1 LOAD DEFINITION

SUMMARY

O 2.1.1 SRV Load Definition Summary 2.1.2 LOCA Load Definition Summary 2.2 DESIGN ASSESSMENT

SUMMARY

2.2.1 Containment Structure, Reactor Enclosure and Control Structure Assessment Summary 2.2.1.1 Containment Structure Assessment Summary 2.2.1.2 Reactor Enclosure and Control Structure Assessment l dummary 2.2.2 Containment Submerged Structures Assessment Summary 2.2.3 Piping Systems Assessment Summary 2.2.4 NSSS Assessment Summary 2.2.5 Equipment Assessment Summary 2.2.6 Electrical Raceway System Assessment Summary 2.2.7 HVAC Duct System Assessment Summary 2.2.8 Suppression Pool Temperature Monitoring System Assessment Summary O

S-i Rev. 1, 09/82

(' LGS DAR

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4.2.4.1 Design Basis Accident (DBA) Transient 4.2.4.2 Intermediate Break Accident (IBA) Transients 4.2.4.3 Small Break Accident (SBA) Transients 4.2.5 LOCA Loading Histories for LGS Containment Components 4.2.5.1 LOCA Loads on the Containment Wall and Pedestal 4.2.5.2 LOCA Loads on the Basemat and Liner Plate 4.2.5.3 LOCA Loads on the Drywell and Drywell Floor 4.2.5.4 LOCA Lcads on the Columns

4. 2. 5. ' LOCA Loads on the Downcomers 4.2.5., LOCA Loads on the Downcomer Bracing 4.2.5.'/ LOCA Loads on Wetwell Piping 4.2.6 References CHAPTER 5 LOAD COMBINATIONS FOR STRUCTURES, PIPING, AND EQUIPMENT

5.1 INTRODUCTION

5.2 LOAD COMBINATIONS FOR CONCRETE DESIGN IN CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL g- STRUCTURE _

( 5.2.1 References 5.3 STRUCTURAL STEEL AND ASME CLASS MC STEEL COMPONENTS LOAD COMBINATIONS 5.4 LINER PLATE LOAD COMBINATIONS 5.5 DOWNCOMER LOAD COMBINATIONS 5.6 PIPING, OUENCHER, AND QUENCHER SUPPORT LOAD l COMBINATIONS 5.6.1 Load Considerations for Piping Inside the Drywell 5.6.2 Load Considerations for Piping Inside the Wetwell 5.6.3 Quencher and Quencher Support Load Considerations 5.6.4 Load Considerations for Piping in the Reactor Enclosure 5.7 NSSS LOAD COMBINATIONS 5.8 EQUIPMENT LOAD COMBINATIONS 5.9 ELECTRICAL RACEWAY SYSTEM LOAD COMBINATIONS 5.10 HVAC DUCT SYSTEM LOAD COMBINATIONS O

D S-iii Rev. 1, 09/82

() LGS DAR CHAPTER 6 DESIGN CAPABILITY ASSESSMENT CRITERIA

6.1 INTRODUCTION

6.2 CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL l STRUCTURE CAPABILITY ASSESSMENT CRITERIA 6.2.1 Containment Structure Capability Assessment Criteria 6.2.2 Reactor Enclosure and Control Structure Capability l Assessment Criteria 6.3 STRUCTURAL STEEL CAPABILITY ASSESSMENT CRITERIA 6.4 LINER PLATE CAPABILITY ASSESSMENT CRITERIA 6.4.1 References 6.5 DOWNCOMER CAPABILITY ASSESSMENT CRITERIA 6.6 PIPING, QUENCHER AND QUENCHER SUPPORT CAPABILITY ASSESSMENT CRITERIA 6.7 NSSS CAPABILITY ASSESSMENT CRITERIA 6.8 EQUIPMENT CAPABILITY ASSESSMENT CRITERIA 6.9 ELECTRICAL RACEWAY SYSTEM CAPABILITY ASSESSMENT CRITERIA 6.10 HVAC DUCT SYSTEM CAPABILITY ASSESSMENT CRITERIA CHAPTER 7 DESIGN ASSESSMENT 7.1 ASSESSMENT METHODOLOGY 7.1.1 Containment, Reactor Enclosure, and Control l Structure Assessment Methodology 7.1.1.1 Containment Structure 7.1.1.2 Reactor Enclosure and Control Structure l 7.1.2 Structural Steel Assessment Methodology 7.1.2.1 Suppression Chamber Columns 7.1.2.2 Downcomer Bracing 7.1.2.3 ASME Class MC Steel Components l 7.1.3 Liner Plate Assessment Methodology 7.1.4 Downcomer Assessment Methodology O.. 7.1.4.1 Structural Model 7.1.4.2 Loads 7.1.4.3 Analysis S-iv Rev. 1, 09/82

LGS DAR 7.1.4.4 Design Assessmerit 7.1.4.5 Fatigue Evaluation of Downcomers in Wetwell Airspace 7.1.5 Piping and SRV Eystems Assessment Methodology 7.1.6 NSSS Assessment Methodology 7.1.7 Equipment Assessment Methodology 7.1.7.1 Hydrodynamic Loads 7.1.7.2 Seismic Loads 7.1.7.3 Other Loads 7.1.7.4 Qualification Methods 7.1.8 Electrical Raceway System Assessment Methodology 7.1.9 HVAC Duct System Assessment Methodology 7.1.10 References 7.2 DESIGN CAPABILITY MARGINS 7.2.1 Stress Margins 7.2.1.1 Containment Structure 7.2.1.2 Reactor Enclosure and Control Structure l 7.2.1.3 Suppression Chamber Columns 7.2.1.4 Downcomer Bracing I I 7.2.1.5 Liner Plate

7.2.1.6 Downcomers 7.2.1.7 Electrical Raceway System 7.2.1.8 HVAC Duct System 7.2.2 Acceleration Response Spectra 7.2.2.1 Containment Structure 7.2.2.2 Reactor Enclosure and Control Structure l CHAPTER 8 MARK II T-0UENCHER VERIFICATION TEST (See Proprietary Section) l l

O S-v Rev. 1, 09/82

LGS DAR 1.4 PLANT DESCRIPTION The Limerick Generating Station (Units 1 and 2) is located on the east bank of the Schuylkill River in Limerick Township of Montgomery County, Pennsylvania, approximately 1.7 miles southeast of the limits of the Borough of Pottstown and approximately 20.7 miles northwest of the Philadelphia city limits.

Each of the LGS units employs a General Electric Company boiling-water reacter (BWR) designed to operate at a rated core thermal power of 3293 MWt (100% steam flow) with a correspondir.g gross electrical output of 1092 MWe. Approximately 37 MWe are used for auxiliary power, resulting in a net electrical output of 1055 MWe.

Commercial operation of LGS Unit 1 is scheduled for April 198' and Unit 2 for October 1987.

1.4.1 PRIMARY CONTAINMENT

)

The containment is a reinforced concrete structure consisting of a cylindrical suppression chamber beneath a truncated conical drywell. Figures 1.4-1 and 1.4-2 show the cross section of the containment and suppression chamber (including pedestal),

respectively. The conical portion of the primary containment (drywell) encloses the reactor vessel, reactor coolant recirculation loops, and associated components of the reactor coolant system. The drywell is separated from the wetwell, i.e, the pressure suppression chamber and pool, by the drywell floor, also named the diaphragm slab. The cone and cylinder form a structurally integrated reinforced concrete vessel, lined with steel plate and closed at the top of the drywell with a steel domed head. The carbon steel liner plate is anchored to the concrete by structural steel members embedded in the concrete and welded to the liner plate.

The entire containment is structurally separated from the surrounding reactor enclosure except at the base foundation slab (a reinforced concrete mat, top lined with a carbon steel liner plate) where a seismic gap filled with roloform is provided between the two adjoinino foundation slabs. The containment structure dimensions and mrameters are listed in Tables 1.4-1

() and 1.4-2.

1.4-1 Rev. 1, 09/82

LGS DAR Major systems and components in the containment include the vent pipe system (downcomers) connecting the drywell and wetwell, vacuum relief system, containment cooling system, and main steam relief valve (MSRV) discharge piping and associated quencher components. Figure 1.4-3 shows the locations and orientation of the quenchers and discharge piping.

1.4.1.1 Penetrations Services and communications between the inside and the outside of the containment are performed through penetrations. Basic penetration types include pipe penetrations, electrical penetrations, and access hatches (equipment hatches, personnel

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lock, suppression chamber access hatches, and control rod drive (CRD) removal hatch). Each penetration consists of a pipe sleeve with an annular ring welded to it. The ring is embedded in the concrete wall and provides an anchorage for the penetration to resist normal operating and accident loads. The pipe sleeve is also welded to the containment liner plate to provide a leaktight penetration.

1.4.1.2 Internal Structures The internal structures consist of reinforced concrete and structural steel and have the major functions of supporting and shielding the reactor vessel, supporting the piping and equipment, and forming the pressure suppression boundary. These structures include the diaphragm slab, the reactor pedestal (a concentric cylindrical reinforced concrete shell resting on the containment base foundation slab and supporting the reactor vessel; Figure 1.4-2 shows pedestal cross section), the reactor shield wall, the suppression chamber columns (hollow steel pipe columns supporting the diaphragm slab), the drywell platforms, the seismic trusses, the quencher supports, and the reactor steam supply system supports.

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• A D.S VALVES H.M.K.E.S LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT QUENCHER LOCATIONS AND ORIENTATION FIGURE 1.4-3 R EV 1,09/82

d O LGS DAR V

CHAPTER 2

SUMMARY

TABLE OF CONTENTS SECTION TITLE 2.1 LOAD DEFINITION

SUMMARY

2.1.1 SRV Load Definition Summary 2.1.2 LOCA Load Definition Summary 2.2 DESIGN ASSESSMENT

SUMMARY

2.2.1 Containment Structure, Reactor Enclosure, and Control n 2.2.1.1 Structure Assessment Summary Containment Structure Assessment Summary 2.2.1.2 Reactor Enclosure and Control Structure Assessment l Summary 2.2.2 Containment Submerged Structures Assessment Summary 2.2.3 Piping Systems Assessment Summary 2.2.4 NSSS Assessment Summary

2.2.5 Equipment Assessment Summary 2.2.6 Electrical Raceway System Assessment Summary 2.2.7 HVAC Duct System Assessment Summary 2.2.8 Suppression Pool Temperature Monitoring System Assessment Summary t

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O 2-i Rev. 1, 09/82

/3 LGS DAR C

CHAPTER 2

SUMMARY

2.1 SRV LOAD DEFINITION

SUMMARY

2.1.1 SRV LOAD DEFINITION

SUMMARY

Hydrodynamic loads resulting from SRV actuation fall into two categories: loads on the SRV system itself (the discharge line and the discharge quencher device), and the loads on the suppression pool walls and submerged structures.

Loads on the SRV system during SRV actuation include loads on the SRV piping due to effects of steady backpressure, transient water slug clearing, and SRV line temperature. Determination of loading on the quencher body, arms, and support is based on

,s transients resulting from valve opening (water clearing and air

( ) clearing), valve closing, and operation of an adjacent quencher.

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Air clearing loads are examined for four loading cases: symmetric (all-valve) SRV actuation, asymmetric adjacent SRV actuation, single SRV actuation, and automatic depressurization system (ADS-five valves) actuation. Dynamic forcing functione for loading of the containment walls, pedestal, basemat, and submerged structures are developed using techniques discussed in Section 4.1. Loads on the SRV system due to SRV actuation are discussed in Section 4.1.3, and loads on suppression pool walls and submerged structures due to SRV actuation are discussed in Section 4.1.4. A full-scale, unit cell test program was.

conducted at the KWU laboratories to verify these SRV loading specifications. These tests are described in Chapter 8.

Adjacent structures indirectly affected by SRV loads include the reactor enclosure, control structure, and associated equipment and components. The assessment methodology used in determining the SRV load effect on these adjacent structures is described in Section 7.1.1.2.

2.1.2 LOCA LOAD DEFINITION

SUMMARY

C':

'LJ The spectrum of LOCA-induced loads acting on the LGS containment structure is characterized by LOCA loads associated with 2.1-1 Rev. 1, 09/82

LGS DAR poolswell and condensation oscillation and chugging, as well as long-term and secondary LOCA loads.

The LOCA loads associated with poolswell result from short duration transients and include.downcomer clearing loads, water jet loads, poolswell impact and drag loads, pool fallback drag loads, poolswell air bubble loads, and loads due to drywell and wetwell temperature and pressure transients. Techniques used to evaluate these loads are described in Section 4.2.1.

Condensation oscillations result from mixed flow (air / steam) and pure steam flow effects in the suppression pool. Chugging loads result from low mass flux pure steam condensation. The load definitions from these phenomena are contained in Section 4.2.2.

Long-term LOCA loads result from those wetwell and drywell pressure and temperature transients associated with design basis accidents (DBA), intermediate break accidents (IBA), and small break accidents (SBA). Their load definitions are contained in Section 4.2.4.

Structures directly affected by LOCA loads include the drywell walls and floor, wetwell walls, RPV pedestal, basemat, liner plate, columns, downcomers, downcomer bracing system, and wetwell piping. Their loading conditions are described in Section 4.2.5.

Adjacent structures indirectly afft ted by LOCA loads include the reactor enclosure, control structure, and associated equipment and components. The assessment methodology used in determining the LOCA load effect on these adjacent structures is described in Section 7.1.1.2.

O Rev. 1, 09/82 2.1-2

LGS DAR 2.2 DESIGN ASSESSMENT

SUMMARY

Design assessment of the LGS structures and components is achieved by analyzing the response of the structures and components to the load combinations explained in Chapter 5. In Chapter 7, predicted stresses and responses (from the loads defined in Chapter 4 and combined as described in Chapter 5) are compared with the applicable code allowable values identified in Chapter 6.

2.2.1 CONTAINMENT STRUCTURE, tEACTOR ENCLOSURE, AND CONTROL l STRUCTURE ASSESSMENT

SUMMARY

2.2.1.1 Containment Structure Assessment Summary The primary containment walls, base slab, diaphragm slab, reactor pedestal and reactor shield are analyzed for the effects of SRV and LOCA in accordance with Table 5.2-1. The ANSYS finite element program is used for the dynamic analysis of structures.

Response spectra curves are developed at various locations within the containment structure to assess the adequacy of components.

Stress resultants due to dynamic loads are combined with other loads in accordance with Table 5.2-1 to evaluate rebar and concrete stresses. Design safety margins are defined by comparing the actual concrete and rebar stresses at critical sections with the code allowable values. The assessment methodology of the containment structure is given in Section.7.1.1.1.

The containment mode shapes, modal frequencies, and hydrodynamic response spectra are given in Appendix A.

The results of the structural assessment of the containment structure are given in Appendix D.

2.2.1.2 Reactor Enclosure and Control Structure Assessment l Summary The reactor enclosure and control structure are assessed for the g effects of SRV and LOCA loads in accordance with Table 5.2-1 and Table 5.3-1.

2.2-1 Rev. 1, 09/82

LGS DAR Pressure time histories in the wetwell are used to investigate the reactor enclocure and control structure response to SRV and LOCA loads. Maximum time history force responses and broadened response spectra curves are approximately used to assess the adequacy of associated structural components. The assessment methodology of the reactor enclosure and control structure is presented in Section 7.1.1.2.

The mode shapes, modal frequencies, and hydrodynamic response spectra of the reactor enclosure and control structure are presented in Appendix B.

The results of the structural assessment are summarized in Appendix E.

2.2.2 CONTAINMENT SUBMERGED STRUCTURES ASSESSMENT

SUMMARY

Load combinations for the downcomer bracing and suppression chamber columns are presented in Table 5.3-1. Load combinations for the downcomers are presented in Table 5.5-1. The hydrodynamic design assessment methodology for the downcomers, bracing, and columns is presented in Sections 7.1.2 and 7.1.4.

The results of the analysis are presented in Appendix D.

The suppression pool liner plate loads are combined in accordance with Table 5.2-1. Results from the analysis indicate that no structural modification is required (see Sections 7.1.3 and 7.2.1.5).

2.2.3 PIPING SYSTEMS ASSESSMENT

SUMMARY

Containment and reactor enclosure piping systems are being analyzed by the methods presented in Section 7.1.5. The load combinations for piping are described in Table 5.6-1. The results of the analysis are presented in Appendix F.

2.2.4 NSSS ASSESSMENT

SUMMARY

To be provided later.

O Rev. 1, 09/82 2.2-2

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LGS DAR 2.2.5 EQUIPMENT ASSESSMENT

SUMMARY

Non-NSSS safety related equipment in the containment, reactor enclosure, and control structure are assessed by the methods contained in Section 7.1.7. Loads are combined as shown in Table 5.8-1. The results of the analysis are presented in Appendix H.

2.2.6 ELECTRICAL RACEWAY SYSTEM ASSESSMENT

SUMMARY

Electrical raceway system loads are combined in accordance with Table 5.9-1. The assessment methodology and analysis results are presented in Chapter 7.

2.2.7 HVAC DUCT SYSTEM ASSESSMENT

SUMMARY

i HVAC duct system loads are combined in accordance with Table 5.10-1. The assessment methodology and analysis results

,O are presented in Chapter 7.

i 2.2.8 SUPPRESSION POOL TEMPERATURE MONITORING SYSTEM (SPTMS)

ASSESSMENT

SUMMARY

SPTMS adequacy assessment and suppression pool temperature response to SRV discharge are presented in Appendix I.

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O j 2.2-3 Rev. 1, 09/82

O LGS DAR CHAPTER 4 TABLES Number Title 4.1-1 These tables are proprietary and are through located in the proprietary supplement to this DAR 4.1-41 4.2-1 Short-Term LOCA Loads Associated with Poolswell 4.2-2 Short-Term Drywell Pressures During Poolswell 4.2-3 LGS Plant-Unique Poolswell Code Input Data t

4.2-4 Input Data For LGS LOCA Transients 4.2-5 LOCA Water Jet Loads 4.2-6 Deleted O 4.2-7 Poolswell Air Bubble Loads l

I 4.2-8 Poolswell Water Friction Drag Loads i

4.2-9 Deleted l 4.2-10 Maximum Load on Submerged Structures 4.2-11 Component LOCA Load Chart for LGS

4.2-12 Wetwell Piping LOCA Loading Situations i

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! 4-iii Rev. 1, 09/82

/~T LGS DAR b

When the jet is predicted to dissipate, the sphere is traveling at the final jet velocity at the point of maximum jet penetration. This condition is used as the final load calculation point. The final jet velocity is that of the jet front just before the last parcicle leaving the vent reaches the jet front. The velocity of the last particle is disregarded.

The largest water jet loads on affected components are given in Table 4.2-5.

- Boundary-Loads During Poolswell 4.{.1.4 s

Ducing the poolswell transient, the high pressure air bubble that forms in the vicinity of the vent exit creates an increase in pressure on all suppression pool boundaries below the vent exit as well as those walls with which it is in direct contact.

Boundaries that are between the bubble location and the point of maximum pool elevation also experience increased pressure loads corresponding to the increased pressure in the wetwell airspace, es as well as the hydrostatic contribution of the water slug.

V Reference 1.3-1, section 4.2.5, and Reference 1.3-5, nection 2.1.2.5, describe the methodology for specification of these boundary loads. The poolswell analytical model is used to determine the maximum values of bubble pressure and wetwell airspace pressure. The analysis takes the maximum pool elevation

as 1.5 times the initial submergence. Using this data, a static loading is applied to the containment structure as follows

a. For the basemat - uniform pressure equal to the maximum bubble pressure superimposed on the hydrostatic load corresponding to a submergence from vent exit to the basemat

b. For the containment walls below the vent exit - maximum

, bubble pressure plus hydrostatic head correspor. ding to vertical distance from vent exit.

l c. For the containment walls between the vent exit and

, maximum pool elevation - linear variation between t

maximum bubble pressure and maximum wetwell airspace l pressure x_-

l 4.2-5

LGS DAR

d. For the containment walls above maximum pool elevation -

maximum wetwell airspace pressure.

The pressure distribution used for the LGS analysis is shown in Figure 4.2-2.

4.2.1.5 Poolwell Asymmetric Air Bubble Load The methodology used in Section 4.2.1.4 assumes that the air flow rate in each downcomer is equal, leading to a symmetric loading of the containment boundary. Concern has been expressed (Reference 1.3-2, subsection III.B.3.e) that circumferential variations in the downcomer air flow rate can occur, due to drywell air / steam mixture variation, that would result in variations in the bubble pressure load on the wetwell wall. This asymmetric loading condition is calculated by statically applying the maximur air bubble pressure, obtained from the PSAM computer code, to half of the submerged boundary and statically applying the hydrostatic pressure of the water column to the other half of the submerged boundary. The pressure load on the basemat and wetwell walls below the vent exit is the sum of the air pressure and the hydrostatic pressure. For the portion of the wall above the vent exit, the pressure increase due to the air bubble is linearly attenuated from the bubble pressure at the vent exit to zero at the pool surface. This increase is then added to the local hydrostatic pressure to obtain the total pressure. The time period of application of the load is from the termination of vent clearing until the maximum swell height is reached.

These loading conditions are conservative with respect to the NRC's long-term criteria for asymmetric bubble loads (Ref. 1.3-5, Appendix A).

4.2.1.6 Poolswell Impact Load As the pool rises during poolswell, structures located between the initial suppression pool surface and the peak poolswell height are subject to the poolswell impact load. The poolswell maximum elevation is determined by the poolswell analytical model with polycropic exponent of 1.2 for wetwell air compression t a maximum swell height which is the greater of 1.5 times the maximum vent submergence or the elevation corresponding to the drywell floor uplift pressure of 2.5 psid (Ref. 1.3-1 and 1.3-5).

For LGS, Reference 1.3-1 separates all impacted structures into two classes:

Rev. 1, 09/82 4.2-6

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a. Impact loads on small structures (one dimension

< 20 in.)

b. Impact loads on large structures (both dimensions

> 20 in.). These structures are treated on a case-by-case basis.

Poolswell impact loads on small structures are determined as specified in Reference 1.3-5, Appendix A.

The PSAM computer runs summary is provided on Figures 4.2-3 through 4.2-7. These graphs present various poolswell plant-unique characteristics, including pressure-time, AP-time, velocity-time, velocity-height, and height-time parameters.

(% s

\sI 4.2.1.7 LOCA Air Bubble Submerced Structure Load During the drywell air purge phase of a LOCA, an expanding bubble is created at the downcomer exits. These rapidly expanding bubbles create three-dimensional velocity and acceleration fields.

l To determine the drag loads, the system was modeled acoustically by the inhomogeneous wave equation (Reference 4.2-8). A bubble source was developed from 4T test data and qualitative information. Table 4.2-7 presents major LOCA air bubble loads.

4.2.1.8 Poolswell Draa Load Subsequent to bubble contact, all bubbles are assumed to coalesce into a blanket of air, and the poolswell drag loads are due to the slug of water rapidly accelerating upward. The loads act in the vertical direction only (except for lift forces that act in l the transverse direction to the flow). The one-dimensional l poolswell model is used to predict the velocity and acceleration i ,

/ at the structure location. As recommended in References 1.3-5 and 1.3-2 and consistent with Section 4.2.3.5 of Reference 1.3-1, the velocity is increased by 10% for additional conservatism to 4.2-7 Rev. 1, 09/82

LGS DAR account for possible bubble asymmetry. Once the flow field is known, the drag forces are calculated by the methods of Appendix C. This methodology conservatively estimates a standard drag coefficient for unsteady flow. This drag load applies to any structure located between the elevation of the vent exit and the peak poolswell height. The duration of the drag load begins when the vent clears, except for structures that are originally not submerged. For structures that are not submerged, the drag load duration is based on the slug transient time (Reference 4.2-6, page 4-78, step 3). Friction drag forces on vertical piping, downcomers, and columns are given in Table 4.2-8.

4.2.1.9 Poolswell Fallback Load After the termination of poolswell, the slug of water falls under the influence of gravity, causing drag forces on structures located between the peak poolswell height and the vent exit. The motion of the water is described by the following equations:

2 H(t) =H - 1 gt (4.2-3) max 2 V (t) = gt FB V =g FB where:

g = the acceleration of gravity H't) = the height above initial water level at time t Hmax = the maximum swell height t = the time (starting with t=0) at maximum swell height The drag load is then calculated from the methods of Appendix C.

The loading stops when H(t) has fallen below the structure or when H(t) has returned to the normal water level, whichever is calculated to occur first.

4.2-8

LGS DAR TABLE 4.2-4 INPUT DATA FOR LGS LOCA TRANSIENTS Drywell free air volume (including downcomers) 248,950 ft3-Wetwell free air volume 149,425 fta Maximum downcomer submergence 12.25 ft Downcomer flow area (total) 256.5 ftz Downcomer loss coefficient 2.11 Initial drywell pressure 14.8 psia Initial wetwell pressure 15.45 psia l Initial drywell humidity 100%

Initial pool temperature 900F l

Estimated DBA break size 3.538 ft2 i

. Number of vents 87 I

Minimum suppression pool mass 5.83 x 106 lb Initial vessel pressure 1,055 psia l Vessel and internals mass 2,940,300 lb Vessel and internals overall heat 484.9 Btu /sec cp Vessel and internals specific heat 0.123 Btu /lb l

Rev. 1, 09/82

LGS DAR TABLE 4.2-6 i

O DELETED l O

Rev. 1, 09/82

i I

i iO LGS DAR TABLE 4.2-7 l POOLSWELL AIR BUBBLE LOADS 1

i Water volume in downcomers 3142.16 ft3 4

Pool surface area (outside pedestal) 4973.89 ft2 Maximum poolswell after water discharge 18.88 ft j Height of downcomer water in the pool 7.58 in.(0.632 ft)

Maximum poolswell height (18.88 + 0.632 ft) 19.51 ft i

j Basemat hydrostatic pressure 10.51 psig l Downcomer tip hydrostatic pressure 5.20 psig

{ Maximum air bubble pressure 48.25 psia Maximum pressurs asemat 58.76 psia

) x- Maximum pressurs .c downcomer tip 48.25 psia i

! Maximum poolswell inside the pedestal 212 ft-9 in. l (6.62 ft above i the high water i level) i i

i I

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LGS DAR TABLE 4.2-9 O

DELETED l

\

\

Rev. 1, 09/82

s ,,

I i

~

8.0. SLAB

-  : EL. 234' 5" REACTOR j<

PEDESTAL

._ / VACUUM E L. 229' BREAKER - 0"

.: T.O. PLAT FORM

~ ,, Y - - -

T -~ E,L. 226' 9" ,

MAXIMUM POOL SWELL EL. 225' 8.12" MAXffAUM POOL SWELL k'

HEIGHT = 19.51' \ /

!s

.i

. *r " .

HIGH NATER LEVEL f') . .'l n o

v

" [ EL. 206' 2" i i l V BRACING * % NORM WATER LEVEL

~, , EL. 203' 5" EL. 204' 11"

- 4 >_4 r_

L MAXIMUM VENT SUBMERGENCE

= 12' - 3" B.O. VENT PIPE, q , i i E L.193' - 11" ,,

7 DI APHR AGM SLAB WETWELL SUPPORT COLUMN PIPING "

12*-0"

r. , ,

3'6" T.O. SL AB

{ .

i .' 't *

.. l*

EL.181* 11" l LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT COMPON ENTS AFFECTED BY LOCA LOADS FIGURE 4.216 R EV 1,09/82

i

(' LGS DAR (s,T/ .

CHAPTER 5 LOAD COMBINATIONS FOR STRUCTURES, PIPING, AND EQUIPMENT TABLE OF CONTENTS

't Section Title

5.1 INTRODUCTION

5.2 LOAD COMBINATIONS FOR CONCRETE DESIGN IN CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL STRUCTURE l 5.3 STRUCTURAL STEEL AND ASME CLASS MC STEEL COMPONENTS LOAD COMBINATIONS 5.4 LINER PLATE LOAD COMBINATIONS 5.5 DOWNCOMER LOAD COMBI!iATIONS 5.6 PIPING, QUENCHER, AND QUENCHER SUPPORT LOAD COMBINATIONS l 5.6.1 Load Considerations for Piping Inside the Drywell O' 5.6.2 5.6.3 Load Considerations for Piping Inside the Wetwell Quencher and Quencher Support Load Considerations 5.6.4 Load Considerations for Piping in the Reactor Enclosure l 5.7 NSSS LOAD COMBINATIONS 5.8 EQUIPMENT LOAD COMBINATIONS 5.9 ELECTRICAL RACEWAY SYSTEM LOAD COMBINATIONS j 5.10 HVAC DUCT SYSTEM LOAD COMBINATIONS i

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5-i Rev. 1, 09/82

i LGS DAR 1

CHAPTER 5 TABLES Number Title 5.2-1 Load Combinations for Concrete Design in Containment, Reacto- Enclosure, and Control Structure (Considering Hydrodynamic Loads) 1 5.3-1 Load Combinations and Allowable Stresses for Structural

Steel Components

, 5.3-2 Load Combinations and Allowable Stresses for ASME Class

MC Components 5.5-1 Load Combinations and Allowable Stresses for Downcomers i

5.6-1 Load Combinations and Stress Limits for Piping Systems l

j 5.8-1 Load Combinations and Damping Values for Non-NSSS

()

Safety-Related Equipment in the Primary Containment, Reactor Enclosure, and Control Structure

5.9-1 Load Combinations and Allowable Stresses for Electrical Raceway System i

j 5.10-1 Load Combinations and Allowable Stresses for HVAC Duct

Systems

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f LGS DAR k

5.2 LOAD COMBINATIONS FOR CONCRETE DESIGN IN CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL STRUCTURE l The loads on the containment, its concrete internals (i.e., RPV pedestal, diaphragm slab), reactor enclosure, and control structure are combined to assess the structural integrity in accordance with the design load combinations given in Table 5.2-1. The factored load approach is used in the assessment,of the concrete structural components. The load factors adopted are based on the degree of certainty and probability of occurrence for the individual loads as discussed in Reference 1.3-1, section 5.1.2. l The loss-of-coolant accidents are characterized by several phenomena that result in non-concurrent loadings on the structures. Time sequences of occurrence of the various time dependent loads, as shown in Figures 5-5 through 5-20 of Reference 1.3-1, are taken into account to determine the most critical loading conditions.

O l

6 5.2-1 Rev. 1, 09/82

n .-

LOAD COMBINATIONS FOR CONCRETE I l

Equa- Load P T R E tion Condition D L o o o o 1 Normal w/o Temp. 1.4 1. 7 1.0 - - --

2 Normal w/ Temp. 1.0 1. 3 1.0 1.0 1.0 -

3 Normal Sev. Env. 1.0 1. 0 1.0 1.0 1.0 1.2 4 Abnormal 1.0 1.0 - - - -

4a Abnormal 1.0 1.0 - - - -

5 Abnormal Sev. Env. 1.0 1.0 - - -

1.1 Sa Abnormal Sev. Env. 1.0 1.0 - - -

1.1 6 Normal Ext. Env. 1.0 1.0 1.0 1.0 1.0 -

7 Abnormal Ext. Env. 1.0 1.0 - - - -

7a Abnormal Ext. Env. 1.0 1.0 - - - -

Load Description D = Dead Loads L = Live Loads P = Operating Pressure Loads o

T = Operating Temperature Loads o

R = Operating Pipe Reactions o

SRV = Safety Relieve Valve Loads

LGS DAR TABLE 5.2-1 (Page 1 of 2)

$ SIGN IN CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL STRUCTURE q CONSIDERING llYDRODYNAMIC LOADS) j E P P T B R Single ss B A A A V SRV AOT(1) ADS ASYM Valve LOCA(3) j

1. 5 X(2) X X -

j 1.3 X -

X -

B - - - - -

% 25 X -

X -

1.25 -

1.0 1. 0 -

1.25 -

X X -

X 1.25 1.0 1.0 -

1.0 - - -

X X 1.1 -

1.0 1.0 -

1.1 -

X X -

X 1.1 1.0 1.0 -

1. 0 - - '-

X X 14 0 - - - - -

1.0 X -

X - -

1.0 1.0 -

1.0 1.0 1.0 1. 0 -

X X -

X 1.0 -

1.0 1. 0 1. 0 1.0 1. 0 - - -

X X i

Rev. 1, 09/82

~m E = Operating-Basis Earthquake o

E = Safe Shutdown Earthquake ss P = SRA or IBA (LOCA) Pressure Load B

B = Pipe Break Temperatures Reaction Loads A

P = DBA (LOCA) Pressure Load A

T = Pipe Break Temperature Load A

R = Reaction and jet forces associated with V

AOT = Abnormal Operating Transient ADS = Automatic Depressurization System ASYM = Asymmetric (1) For columns designated AOT, ADS, ASYM, @

columns may be included in the load comb Equation 1, either AOT or ASYM may be cc ASYM simultaneously.

(2) X indicates applicability for the design (3) LOCA includes chugging, condensation osc l

[

t

^

LGS DAR (Page 2 of 2)

TABLE 5.2-1 (Cont'd) the pipe break I

nd Single Valve, only one of the four possible j ination for any one equation. For example, in ,

asidered w' th the other loads but not both AOT and ated load combination. l illation, and large air bubble loads. l Rev. 1, 09/82

I LGS DAR

, 5.3 STRUCTURAL STEEL AND ASME CLASS MC STEEL COMPONENTS LO' .~~

COMBINATIONS 1

The load combinations for structural steel in the containment, reactor enclosure, and control structure are given in Table 5.3-1. These combinations apply to the suppression chamber l steel columns, the downcomer bracing, and miscellaneous structural steel within the containment, reactor enclosure, and l

control structure.

The loss-of-coolant accidents are characterized by several phenomena that result in non-concurrent loadings on the structures. Time sequences of occurrence of the various time dependent loads, as shown in figures S-5 through 5-20 in Reference 1.3-1, are taken into account to determine the most critical loading conditions.

The load combinations for the ASME Class MC steel components in the concrete containment are given in Table 5.3-2. These i combinations apply to the drywell head assembly, equipment-

. hatches, personnel lock, suppression chamber access hatches, s control rod drive removal hatch, and piping and electrical penetrations.

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9 LGS DAR TABLE 5.3-1 (Page 1 of 2)

LOAD COMBINATIONS AND ALLOWABLE STRESSES FOR STRUCTURAL STEEL COMPONENTS (Suppression Chamber Columns, Downcomer Bracing, and Reactor Building Structural Steel)

Equation Condition Load Combination Allowable Stress 1 Normal D+L+P +SRV F w/o Temp. o s l 2 Normal D+L+P +T +SRV F w/ Temp. o o s 3 Normal / D+L+P +T +E+SRV 1.25 F o o Severe s

4 Normal / D+L+P +T +E'+SRV (1)

Extreme o o

! 5 Abnormal D+L+P+(T +T )+R (1) o a

+SRV+LOCA l 6 Abnormal / D+L+P+(T +T )+R+E (1)

Severe o a

+SRV+LOCA l 7 Abnormal / D+L+P+(T +T )+R+E' (1)

. Extreme o a I +SRV+LOCA (1) In no case shall the allovable stress exceed 0.90 Fy in bending, 0.85 Fy in axial tension or compression, and C.50 Fy in shear. Where the design is governed by i

requirements of stability (local or lateral buckling),

the actual stress shall not exceed 1.5F .

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LGS DAR TABLE 5.3-1 (Cont'd) (Page 2 of 2)

Notations:

F = Allowable stress according to the AISC, i s " Specification for the Design, Fabrication, and

, Erection of Structural Steel for Buildings," dated 1969, Part 1 Fy = Minimum specified yield strength D = Dead load

]

L = Live load To = Thermal effects during normal operating conditions

including temperature gradients and equipment and pipe reactions T = Added thermal effects (over and above operating 4 a thermal effects) that occur during a design accident Po = Operating Pressure Load l f P = Design basis accident pressure load R = Local force or pressure on structure due to postulated pipe rupture including the effects of steam / water jet impingement, pipe whip, and pipe reaction E = Load due to operating basis earthquake E' = Load due to safe shutdown earthquake SRV = Safety relief valve loads ,

LOCA = Loads due to loss-of-coolant accident conditions *

(chugging, condensation oscillation, or large air.

bubble loads)  !

4 i

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Rev. 1, 09/82

LGS DAR TABLE 5.3-2 (Page 1 of 2) l LOAD COMBINATIONS AND ALLOWABLE STRESSES FOR ASME CLASS MC COMPONENTS The drywell head assembly, equipment hatches, personnel lock suppression chamber access hatches, CRD removal hatch, and piping and electrical penetrations are designed for the following loading combinations and allowable stresses:

Equation Condition Load Combination Stress Limits 1 Normal D+L+1.15P 1.15 times ASME Section III, Class B 2 Normal D+L+T +P ASME Section III, A Class B 3 Emergency D+L+T +P+H +R+E ASME Section III, A A Summer 1970 Addenda, Figure N-414

() 4 Faulted D+L+T +P+H +R+E' A A ASME Section III, Summer 1970 Addenda, Figure N-414 5 Normal D+L+T +SRV ASME Section III, w/ Temp. o Class MC Components 6 Abnormal / D+L+T +P+H +R+E ASME Section III, Severe A A Fig. NB-3224-1 for

+SRV+LOCA " Emergency Conditions" 7 Abnormal / D+L+T +P+H +R+E ASME Section III, Extreme A A Fig. NB-3225-1 for

+SRVdLOCA " Faulted Conditions" Definitions D = Dead load L = Live Load T = Thermal effects due to temperature gradient o through the wall, under accident conditions

/)

(s/

T A

= Thermal effects due to temperature gradient through the wall, under accident conditions P = Design basis accident pressure load Rev. 1, 09/82

] LGS DAR J

TABLE 5.3-2 (Page 2 of 2)

R = Steam / water jet forces or reactions resulting from the rupture of process piping E = Load due to the operating basis earthquake (OBE)

E' = Load due to the design basis earthquake (SSE)

B = Hydrostatic loading due to post-accident flooding of the primary containment to the level of the reactor core H = Force on the structure due to thermal A expansion of pipes, under accident conditions SRV = Safety / relief valve loads LOCA = Loads due to loss-of-coolant accident conditions (chugging, condensation oscillation, annulus pressurization or large air bubble loads)  ;

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Rev. 1, 09/82

LGS DAR 5.5 DOWNCOMER LOAD COMBINATIONS Load combinations and stress allowables for the downcomers are given in Table 5.5-1. These load combinations are based on the load combinations given in table 5-2 of Reference 1.3-1.

The loss-of-coolant accidents are characterized by several phenomena that result in non-concurrent loadings on the structures. Time seg;ences of occurrence of the various time dependent loads, as .thown in figures 5-5 through 5-20 in Reference 1.3-1, are taken into account to determine the most critical loading conditions.

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/ LGS DAR N]'

5.6 PIPING, QUENCHER, AND OUENCHER SUPPORT LOAD COMBINATIONS LOCA loads considered on piping systems include poolswell impact loads, poolswell drag loads, downcomer water jet loads, poolswell air bubble loads, fallback drag loads, condensation oscillation loads, chugging loads, and inertial loading due to the acceleration of the containment structure produced by LOCA loads.

Loads due to SRV discharge on piping systems include water clearing loads, air clearing loads, fluid transient loads on SRV discharge piping, reaction forces at the quencher, and inertial loading due to the acceleration of the containment structure produced by SRV discharge * #ds.

The load combinations and stress limits for piping systems are given in Table 5.6-1.

5.6.1 LOAD CONSIDERATIONS FOR PIPING INSIDE THE DRYWELL Piping systems inside the drywell are subjected to inertial (c). loading due to the acceleration of the containment produced by LOCA and SRV discharge. loads in the wetwell. The SRV discharge piping in the drywell is also subjected to fluid transient forces due to SRV discharge.

5.6.2 LOAD CONSIDERATIONS FOR PIPING INSIDE THE WETWELL All piping in the wetwell is subject to the inertial loading due to LOCA and SRV discharge.

Drag and impact loads due to LOCA and SRV discharge on individual pipes in the wetwell depend on the physical location of the I piping. Other SRV discharge and LOCA loads applicable to piping in the wetwell are discussed in the paragraphs that follow.

Piping systems located below the suppression chamber water level are shown on Figures 5.6-1 and 5.6-2. In addition to the inertial loads, these piping systems are subjected to SRV air bubble and LOCA air bubble loads, condensation oscillation loads, and chugging loads. The SRV pipinq, quencher, and quencher support are also subject to fluid ;ransient forces due to SRV discharge. Piping systems located within the jet impingement

{s_)T cone of the downcomer are also subjected to downcomer water jet loads.

5.6-1

LGS DAR Piping systems within the poolswell zone a e shown on Figures 5.6-2, 5.6-3, and 5.6-4. All horizontal runs of these pipes are above the suppression chamber water level. The following loads, in addition to the inertial loads, act on these systems:

a. The horizontal runs of pipe below elevation 225'-8',

experience poolswell impact, poolswell drag, and fallback drag loads.

b. The vertical portions of pipe in the water below elevation 225'-8" experience poolswell drag and fallback drag loads.

5.6.3 OUENCHER AND QUENCHER SUPPORT LOAD CONSIDERATIONS The quencher and quencher supports are subjected to the following hydrodynamic loads in addition to the pressure, weight, thermal, and seismic-loads:

a. Unbalanced loads on the quencher due to SRV water O

clearing and air clearing transients, irregular condensation, and steady-state blowdown.

l l

l b. Drag loads due to SRV discharge and LOCA.

c. SRV piping end loads.

d. Inertial loading due to the acceleration of the containment produced by SRV discharge and LOCA.

5.6.4 LOAD CONSIDERATIONS FOR PIPING IN THE REACTOR ENCLOSURE The effects of the inertial loading due to acceleration of the containment produced by SRV discharge and LOCA loads are evaluated for this piping.

Rev. 1, 09/82 5.6-2

l l

l l

LGS DAR TABLE 5.6-1 LOAD COMBINATIONS AND STRESS LIMITS FOR PIPING SYSTEMS Stress Equation Condition Load Combination Limit l 1 Design PD NB-3652, NC-3600, ND-3600 2 Normal PD + DW NB-3654, NC-3600, ND-3600 3 Upset (a) PO+DW+(OBE2+SRV2)a/2 NB-3654, (b) PO+DW+(RVC2+0BE2)*/2 NC-3600, (c) PO+DW+FV ND-3600 (d) PO+DW+0BE+RVO 4 Emergency (a) PO+DW+(OBE2+SRV2 +SBA2) /2 NB-3655, ADS NC-3600, (b) PO+DW+(FV2+0BE2)1/2 ND-3600 5 Faulted (a) PO+DW+(OBE2+SRV2 +IBA2)1/2 NB-3656, ADS ASME Code (b) PO+DW+(SSE2+SRV2 +IBA2)1/2 Case 1606 ADS (c) PO+DW+(SSE2+DBA2) /2 Notations:

PD = Design pressure PO = Operating pressure DW = Dead weight OBE = Operating basis earthquake (inertia portion)

SSE = Safe shutdown earthquake (inertia portion)

SRV = Loads due to safety relief valve blow, axisymmetric x or asymmetric i SRV = Load due to automatic depressurization SRV blow, t ADS axisymmetric l SBA = Small break accident IBA = Intermediate break accident DBA = Design basis accident FV = Transient response of the piping system associated with fast valve closure (transients associated with valve closure t.mes less than 5 seconds are considered)

RVC = Transient response of the piping system associated with relief valve opening in a closed system RVO = Sustained load or response of the piping system

(-)g

(, associated with relief valve opening in an open system or last segment of the closed system with steady state load l Rev. 1, 09/82

av DWG PENE- TYPE OF ELEVATION N o. LINE No. QTY SYSTEM TRATION PENE-No. TRATION A B B 4"-HBD 187 1 HPCI X 212 EMBEDDED 207'-6" 199'11" h B 4" HBD 188 1 HPCI X-236 EMBEDDED 207'-6" 199'11" h A 24" HBD 189 1 HPCI X-210 SLEEVE 207'-6" 192'-8" h B 4"-HBD 171 2 kkRfY (:l$ EMBEDDED 201'-6" 199'-11" B 10" HBD 169 2 hkRIY ):lM EMBEDDED 207'-6" 199'11" d A 18"-GBD-143 2 RHR X 204A,B SLEEVE 219'.0" 199'-11" -

A 4" GBD 144 2 RHR X 226A SLEEVE 207'-6" 199'-11" -

A 12".HBD 173 1 RCIC .x.215 SLEEVE 207*-6" 199'11" -

B 6" HisB-139 1 RHR X 240 EMBEDDED 207'-3 1/4" 199'-11" h A 10".HBB 140 1 RHR X 238 SLEEVE 207'-9" 199'11" -

A 10" HBB 140 1 RHR X-239 SLEEVE 207'-1" 199'11" -

C 4".HCB-106 1 N6$$kTgOLID X 231 A SLEEVE 207'7" 205'1" 2 C 4~ HCB-107 1 koh$kTg LlD X 231B SLEEVE 207'9" 204'-6" 2 B 2" HBD 357 1 hokh $N bObLING X 217 EMBEDDED 207'-6" 199'11" B 2"-HBD-356 1 bdyLNkON hObLING X-216 EMDMD 207'-6" 199*-11" -

W w

~

X ..

X w .

• . *l

. jf EL.A 4'

-4 - - - -

EL.A 6 n '; -

. -h

• N p 4

,.b ..,

EL.B #

EL B DRAWING A DR AWING B

~

_. . -----.-_--,.n. ,, . . .

!i l

i i

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i4 j ,

' 1 LGS DAR l

5.8 EQUIPMENT LOAD COMBINATIONS l I

4 Safety-related equipment located within the primary containment, l reactor enclosure, and control structure are assessed for the j governing load combinations shown in Table 5.8-1.

i l

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i j

l i

i 1

i -

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I f i

I i 1 <

l I  !

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5.8-1 Rev. 1, 09/82

/~' LGS DAR b)

TABLE 5.8-1 i

LOAD COMBINATIONS AND DAMPING VALUES FOR NON-NSSS SAFETY-RELATED EQUIPMENT IN THE PRIMARY CONTAINMENT, REACTOR ENCLOSURE, AND l CONTROL STRUCTURE

! Equation Condition Load Combination Dampina(1) l 1 Upset a. N+[OBE2 + SRV2)1/2 2%

. b. N+0BE 0.5%

2 Emergency a. N+[OBE2 + SRV2 + SBA2)2/2 2%

3 Faulted a. N+[OBE2 + SRV2 + IBA2]1/2 2g

b. N+[SSE2 + SRV2 + IBA2]1/2 2%

c. N+[SSE2 + DBA2]1/2 2%

d. Envelope of a, b& c 2%

e. N+SSE 0.5%

4 Worst a. Envelope of la, 2 and 3d 2%

Notations:

N = Normal loads (dead weight + operating temp + operating press.) l OBE = Operating basis earthquake loads l SSE = Safe shutdown earthquake loads SRV = Safety relief valve discharge loads SBA = Small break accident loads IBA = Intermediate break accident loads DBA = Design basis accident loads (1) Where justified, a higher damping value may be used.

O

'V Rev. 1, 09/82

~N LGS DAR TABLE 5.10-1 (Page 1 of 2)

LOAD COMBINATIONS AND ALLOWABLE STRESSES FOR HVAC DUCT SYSTEMS

~

DUCTS Allowable Equation Condition Load Combination Stress 1 Normal D+L+SRV Fs 2 Normal D+P +SRV Fs M

3 Abnormal D+P 1.25F T s 4 Normal / Severe D+P +E 1.25F (1)

M s 5 Normal / Severe D+P +E+SRV 1.25F M s 6 Normal D+Po Fs 4

7 Normal / Severe D+Po+E 1.25F s

i 8 Normal / Extreme- D+Po+E' (2) 9 Normal / Extreme D+P +E'+SRV (2) h.

N/

4 10 Extreme / Abnormal M

D+P +P +E'+SRV+LOCA (2) 4 O A 11 Extreme / Abnormal When protection against tornado depressurization is required:

12 Extreme / Abnormal D+P +W +SRV+LOCA (2)

O D i For ducts inside drywell of containment, the fol-lowing additional load combination is also applicable:

D+H +P +P +E'+SRV+LOCA (2)

A O A DUCT SUPPORTS Allowable Equation Condition Load Combination Stress

1 Normal D+L+SRV Fs

g 2 Normal / Severe D+E 1.25F (1) s l 3 Normsl/ Severe D+E+SRV (2) 4 Extreme / Abnormal D+E'+SRV+LOCA (2)

Rev. 1, 09/82 I

LGS DAR CHAPTER 6 DESIGN CAPABILITY ASSESSMENT CRITERIA TABLE OF CONTENTS

6.1 INTRODUCTION

6.2 CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL STRUCTURE l i CAPABILITY ASSESSMENT CRITERIA 6.2.1 Containment Structure Capability Assessment Criteria 6.2.2 Reactor Enclosure and Control Structure Capability l Assessment Criteria 6.3 STRUCTURAL STEEL AND ASME CLASS MC STEEL COMPONENTS CAPABILITY ASSESSMENT CRITERIA 6.4 LINER PLATE CAPABILITY ASSESSMENT CRITERIA 6.4.1 References 6.5 DOWNCOMER CAPABILITY ASSESSMENT CRITERIA I h 6.6 PIPING, OUENCHER, AND OUENCHER SUPPORT CAPABILITY

\'- ASSESSMENT CRITERIA 6.7 NSSS CAPABILITY ASSESSMENT CRITERIA 6.8 EQUIPMENT CAPABILITY ASSESSMENT CRITERIA 6.9 ELECTRICAL RACEWAY SYSTEM CAPABILITY ASSESSMENT CRITERIA 6.10 HVAC DUCT SYSTEM CAPABILITY ASSESSMENT CRITERIA O

6-1 Rev. 1, 09/82

LGS DAR 6.2 CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL STRUCTURE l CAPABILITY ASSESSMENT CRITERIA 6.2.1 CONTAINMENT STRUCTURE CAPABILITY ASSESSMENT CRITERIA The acceptance criteria detailed in the LGS FSAR Section 3.8.1.5 have been used to assess the structural integrity of the containment and internal structures. No changes are made in these acceptance criteria when the effects of the dynamic SRV discharge 7d LOCA loads are included.

6.2.2 REACTOR ENCLOSURE AND CONTROL STRUCTURE CAPABILITY l ASSESSMENT CRITERIA The acceptance criteria for seismic Category I structures presented in the LGS FSAR Section 3.8.4.5 have been used to assess the structural integrity of the reactor enclosure, control structure, and their components. No changes are made in these acceptance criteria when the effects of the dynamic SRV discharge and LOCA loads are included, f)%

u.

6.2-1 Rev. 1, 09/82

LGS DAR 6.3 STRUCTURAL STEEL AND ASME CLASS MC STEEL COMPONENT CAPABILITY ASSESSMENT CRITERIA The allowable stresses for structural steel in the containment, reactor enclosure, and cont:ol structure are given in Table 5.3-

1. These criteria apply to the suppression chamber steel columns, the downcomer bracing, and miscellaneous structural steel within the containment, reactor enclosure, and control structure.

The allowable stresses for ASME Class MC steel components in the concrete containment are given in Table 5.3-2. These allowable stresses apply to the drywell head assembly, equipment hatches, personnel lock, suppression chamber access hatches, and piping and electrical penetrations.

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1 Rev. 1, 09/82 6.3-1

LGS DAR 6.4 LINER PLATE CAPABILITY ASSESSMENT CRITERIA The strains in the liner plate and anchorage system (welds and anchors) from self-limiting loads such as dead load, creep, shrinkage, and thermal effects are limited to the allowable values specified in Table CC-3720-1 of Reference 6.4-1. The.

displacements of the liner anchorage are limited to the displacement values of Table CC-3730-1 of Reference 6.4-1.

Stresses in the liner plate and anchorage system (welds and anchors) from mechanical loads such as SRV discharge and chugging are checked according to Reference 6.4-2. Specifically, primary plus secondary membrane plus bending stresses are checked according to subsection NE-3222.2. Fatigue strength evaluation is based on subsection NE-3222.4. Allowable design stress intensity values, design fatigue curves, and material properties that are used conform to subsection NA, Appendix 1. l The capacity of the liner plate anchorage is limited by the

< -s s concrete pull-out to the service load allowable for oncrete as

y ) specified in Reference 6.4-3.

l 6.

4.1 REFERENCES

6.4-1 ASME Boiler and Pressure Vessel Code,.Section III, Division 2, 1975 Edition l l

6.4-2 ASME Boiler and Pressure Vessel Code,Section III, Division 1, 1974.

6.4-3 ACI 318, " Building Code Requirements for Reinforced Concrete", 1971 Edition.

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{d 6.4-1 Rev. 1, 09/82

O LGS DAR d

6.6 PIPING, OUENCHER, AND OUENCHER SUPPORT CAPABILITY SUPPORT ASSESSMENT CRITERIA Piping systems in the containment and reactor enclosure are analyzed in accordance with ASME Section III, Division 1 (1971 Edition wi.ie Addenda through Winter 1972), subsections NB-3600, NC-3600, and ND-3600, and ANSI B31.1 (Power Piping Code) for the ,

loading described in Table 5.6-1. In addition to these code requirements, when piping is required to deliver rated flow during or following an emergency or faulted event, the functional capability requirement shall be met for the load combinations with the event.

The quencher and quencher support are designed in accordance with ASME Section III, Division 1 (1977 Edition with Addenda through Summer 1979), subsections NC-3200 and NF-3000, respectively, for the loading discussed in Section 5.6.3.

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l lO 6.6-1 Rev. 1, 09/82 1

LGS DAR 6.8 EQUIPMENT CAPABILITY ASSESSMENT CRITERIA 6.8.1 ANALYSIS Safety-related equipment located in the primary containment, reactor enclosure, and control structure are analyzed to satisfy l load combinations la, Ib, 2, 3d, and 3e of Table 5.8-1. The maximum load effects result from simultaneous exitation in all three principal directions for all combinations involving dynamic loads as detailed in Section 7.1.7.4.1.3. The operability of active components required to operate during a dynamic event is also considered.

6.8.2 TESTING When safety-related equipment is qualified by testing, a test response spectrum (TRS) is derived to envelope the required response spectrum (RRS) for load combinations Ib, 3e, and 4 of

-}

'y Table 5.8-1. The minimum test sequence is to perform five runs of the TRS for load combination Ib, followed by one run of load combination 3e, then one run of load combination 4.

Qualification is achieved if the equipment does not fail or malfunction during the test. Operability is verified before and after the test sequence. Active components required to function during a dynamic event are also operated during the test.

An example of a combined RRS and an enveloping TRS are presented in Appendix H.

6.8.3 COMBINED ANALYSIS TEST Some equipment is qualified by a combination of analysis and testing procedures. Details of this method, as well as further documentation of the equipment qualification program, are presented in Appendix H.

) '

l 6.8-1 Rev. 1, 09/82

LGS DAR CHAPTER 7 DESIGN ASSESSMENT TABLE OF CONTENTS Number Title 7.1 ASSESSMENT METHODOLOGY 7.1.1 Containment, Reactor Enclosure, and Control l Structure Assessment Methodology 7.1.1.1 Containment Structure 7.1.1.1.1 Hydrodynamic Loads l 7.1.1.1.1.1 Structural Models 7.1.1.1.1.2 Damping 7.1.1.1.1.3 Fluid-Structure Interaction 7.1.1.1.1.4 Supplementary Computer Programs 7.1.1.1.1.5 Load Application 7.1.1.1.1.5.1 SRV Discharge Loads 7.1.1.1.1.5.2 LOCA Related Loads 7.1.1.1.1.6 Analysis

/T 7.1.1.1.1.6.1 Response Spectra Generation (ms/ 7.1.1.1.1.6.2 Stress Analysis 7.1.1.1.2 Seismic Loads 7.1.1.1.3 Static and Thermal Loads 7.1.1.1.4 Load Combinations 7.1.1.1.5 Design Assessment 7.1.1.2 Reactor Enclosure and Control Structure l 7.1.1.2.1 Hydrodynamic Loads 7.1.1.2.1.1 Load Definitions 7.1.1.2.1.2 Hydrodynamic Analysis Models 7.1.1.2.1.2.1 SRV Analysis Models 7.1.1.2.1.2.2 CO Analysis Model 7.1.1.2.1.2.3 CHUG Analysis Models 7.1.1.2.1.2.4 Control Structure Floor / Local Models 7.1.1.2.1.3 Hydrodynamic Analysis 7.1.1.2.1.3.1 Analysis Procedures 7.1.1.2.1.3.2 Generation of Response Data 7.1.1.2.2 Seismic Loads 7.1.1.2.3 Static Loads 7.1.1.2.4 Load Combinations 7.1.1.2.5 Design Assessment 7.1.2 Structural Steel and ASME Class MC Steel Components Assessment Methodology 7.1.2.1 Suppression Chamber Columns 7.1. 2.1.1 Structural Models 7.1.2.1.2 Loads

, ,/ 7.1.2.1.2.1 SRV Discharge Loads 7.1.2.1.2.2 LOCA Related Loads 7.1.2.1.2.3 Seismic Loads 7-1 Rev. 1, 09/82

LGSR DAR i

CHAPTER 7 l TABLE OF CONTENTS (Cont'd)

Number Title 7.1.2.1.2.4 Static Loads 7.1.2.1.2.5 Load Combinations 7.1.2.1.2.6 Design Assessment 7.1.2.2 Downcomer Bracing 7.1.2.2.1 Bracing System Description 7.1.2.2.2 Loads 7.1.2.2.2.1 SRV Discharge Loads 7.1.2.2.2.2 LOCA Related Loads 7.1.2.2.2.3 Seismic Loads 7.1.2.2.2.4 Static Loads 7.1.2.2.2.5 Thermal Load 7.1.2.2.2.6 Load Combinations 7.1.2.2.3 Design Assessment 7.1.2.3 ASME Class MC Steel Components l 7.1.3 Liner Plate Assessment Methodology 7.1.4 Downcomer Assessment Methodology 7.1.4.1 Structural Model 7.1.4.2 Loads N 7.1.4.3 Analysis 7.1.4.4 Design Assessment 7.1.4.5 Fatigue Evaluation of Downcomers in Wetwell Airspace 7.1.5 Piping and SRV Systems Assessment Methodology 7.1.6 NSSS Assessment Methodology 7.1.7 Equipment Assessment Methodology 7.1.7.1 Hydrodynamic Loads 7.1.7.1.1 SRV Discharge Loads 7.1.7.1.2 LOCA Related Loads 7.1.7.2 Seismic Loads 7.1.7.3 Other Loads 7.1.7.4 Qualification Methods 7.1.7.4.1 Dynamic Analysis 7.1.7.4.1.1 Methods and Procedures 7.1.7.4.1.2 Appropriate Damping Values 7.1.7.4.1.3 Three Components of Dynamic Motions 7.1.7.4.2 Testing 7.1.7.4.3 Combined Analysis and Testing 7.1.8 Electrical Raceway System Assessment Methodology 7.1.9 HVAC Duct System Assessment Methodology 7.1.10 References O

7-ii Rev. 1, 09/82

i

] -LGSR DAR l

CHAPTER 7 TABLE OF CONTENTS (Cont'd) 1 i

Number Title 7.2 DESIGN CAPABILITY MARGINS l 7.2.1 Stress Margins 7.2.1.1 Containment Structure 7.2.1.2 Reactor Enclosure and Control Structure l 7.2.1.3 Suppression Chamber Columns 7.2.1.4 Downcomer Bracing i 7.2.1.5 Liner Plate

7.2.1.6 Downcomers 7.2.1.7 Electrical Raceway System 7.2.1.8 HVAC Duct System 7.2.2 Acceleration Response Spectra 7.2.2.1 Containment Structure 7.2.2.2 Reactor Enclosure and Control Structure l O

i l

l l

i i

i l

?

i

!O i

i 7-iii Rev. 1, 09/82

} - - . . - . - - - - . _ - . -- _

LGS DAR O CHAPTER 7 TABLES Number Title 7.1-1 Reactor Enclosure and Control Structure: l Summary of Hydrodynamic Analyses and Corresponding Math Models 7.1-2 Control Structure Floor Model Material Properties 7.2-1 Maximum Spectral Accelerations of Containment Due to SRV and LOCA Loads at 1% Damping.

O O

7-iv Rev. 1, 09/82

LGS DAR CHAPTER 7 FIGURES Number Title 7.1-1 3-D Containment Finite Element Model (ANSYS Model) 7.1-2 Equivalent Modal Damping Ratio Vs. Modal Frequency for Structural Stiffness Proportional Damping (Containment Building) 7.1-3 Reactor Enclosure and Control Structure l Vertical Axisymmetric Coupled Model (FESS) 7.1-4 Reactor Enclosure and Control Structure l Vertical Stick Model 7.1-5 Reactor Enclosure and Control Structure l Horizontal Stick Model 7.1-6 Control Structure Floot " Half Model" 7.1-7 Control Structure Floor " Quarter Model" 7.1-8 Equivalent Modal Damping Ratio Vs. Modal Frequency for Structural Damping (Reactor Enclosure and Control Structure) 7.1-9 Downcomer Bracing System 7.1-10 Downcomer Bracing System Details 7.1-11 Deleted l 7.1-12 Liner Plate Pressures - Normal Condition 7.1-13 Liner Plate Pressures - Abnormal Condition 7.1-14 Liner Plate Pressures - Abnormal Condition 7.1-15 Liner Plate Pressures - Abnormal Condition O

7-v Rev. 1, 09/82

LGS DAR CHAPTER 7 FIGURES (Cont'd)

Number Title 7.1-16 Downcomer Analytical Model 7.1-17 Suppression Chamber Column Analytical Model 7 . '>.- 1 Deleted l O

O 7-vi Rev. 1, 09/82

p LGS DAR Q

CHAPTER 7 DESIGN ASSESSMENT 7.1 ASSESSMENT METHODOLOGY Loads on LGS structures, piping, and equipment are defined in Chapter 4. The methods by which these loads are combined are discussed in Chapter 5. The criteria for establishing design capability are stated in Chapter 6.

This section describer, the assessment methodology used in the final evaluation of structures, piping, and equipment.

7.1.1 CONTAINMENT, REACTOR ENCLOSURE, AND CONTROL STRUCTURE l ASSESSMENT METHODOLOGY 7.1.1.1 Containment Structure

(}

7.1.1.1.1 Hydrodynamic Loads 7.1.1.1.1.1 Structural Models The dynamic analysis for the structural response of the containment and internal structures due to the SRV discharge loads and LOCA loads is performed using the finite element method. The ANSYS (FSAR Section 3.8.7) finite element computer program was chosen for the transient dynamic analysis.

Figure 7.1-1 shows the ANSYS finite element model. The concrEt, containment walls, slabs, RPV, RPV pedestal, and shield wall are modeled with shell elements. The refueling bellows and stabilizer truss are modeled with spar elements. The RPV internals are modeled with beam elements. The suppression pool fluid mass is modeled with lumped mass elements. The ANSYS model includes a total of 797 elements and 206 dynamic degrees of freedom.

The scil structure interaction is taken into consideration by modelling the soil using a series of discrete springs and dampers Os.

in three directions as shown in Figure 7.1-1. e properties of the discrete springs and dampers are calculatea based on the 7.1-1 Rev. 1, 09/82

LGS DAR formulae for lumped parameter foundations found in Reference 7.1-1.

7.1.1.1.1.2 Damping

a. Structural Damping The equations of motion for a discretized structure must include a ter..i to account for viscous damping that is linearly proportional to the velocity. The equations of motion for a damped system are:

[M] {hi+[C] {r} + [K] {r} = {R(t (7.1-1) where [C] is the viscous damping matrix. A viscous damping matrix of the form [C] =a (M] + a [K) (7.1-2) O was used (Reference 7.1-2) where o and a are proportionality constants that relate damping to the velocity of the nodes and the strain rates, respectively. This damping matrix leads to the following relation between a and p and the damping ratio of the ith mode C : i C = c/2w + sw /2 (7.1-3) i i i where w is the natural circular frequency of the ith mode. For the usual case of only structural damping, o = 0 and therefore p = 2C /w . i i Because only a single value of a is permitted in the ANSYS input, the most dominant natural frequency of the structure is selected for the computation of a (Reference 7.1-3). Rev. 1, 09/82 7.1-2

Q V LGS DAR A value of a equal to 0.00063 is used in the ANSYS model whicn :orresponds to structural modal damping of approximately 4 percent of critical at 20 Hz which is the most dominant natural frequency of the structure. Figure 7.1-2 shows modal damping ratio versus modal frequency for structural stiffness-proportional-damping.

b. Soil Springs and Radiation Damping The elastic half-space theory as described by Reference 7.1-1 was used to compute the values of the spring constants and dampers in the horizontal, l vertical, and rocking directions (K , K, K,C,C,C). l H V t H V t The following parameters are used to represent the rock foundation:

Shear modulus of foundation medium () G =

            =     1.154 x 103   KSI v    =    Poisson's ratio of foundation medium
            =    0.3

p = Material density of foundation medium s

            =    0.00481 K-sec2/ft' V    =    Shear wave velocity S
            =     6180  ft/sec                                   l From which we get the following:                          l l       K    =    3.37 x 10*    K/in l         H C    =     1.57 x 104   K-sec/in O

7.1-3 Rev. 1, 09/82

LGS DAR K = 3.96 x 106 K/in V C = 2.72 x 10* K-sec/in V K e 9.5 x 1011 K-in/ Rad

           +

C = 2.29 x 10' K-in-sec/ Rad

           +

The above lumped foundation springs and dampers were then distributed to every node on the basemat according to the tributary area. 7.1.1.1.1.3 Fluid-Structure Interaction The ANSYS finite element model with appropriate fluid - structure coupling was developed for the analysis of the containment structure. The water mass constitutes only 1/7 of the total mass of the reinforced concrete structure. The model used considers fluid - structure coupling by lumping the water mass in the suppression pool at each node of the wetted surface. The weighted area approach was considered to determine the fluid mass at each node of the suppression pool. 7.1.1.1.1.4 Supplementary Computer Programs Supplementary computer programs were used for preprocessing and postprocessing of data generated for or by the ANSYS computer program. Preprocessing programs called PREPRC1, PREPRC2, and PREPRC3 were developed to convert the SRV, condensation oscillation, and chugging pressure time histories into force time histories, respectively, acting at the associated nodes of the ANSYS model. The programs write the nodal force time histories onto a file for processing by ANSYS. A postprocessor program was developed to calculate the nodal acceleration time history. This program is called DISOGE. It reads the structural response displacement time histories generated from ANSYS (displacement pass option), scans for the maximum displacements, and generates the acceleration time histories using the Fast Fourier Transformation method. Rev. 1, 09/82 7.1-4

LGS DAR pl The enveloped response spectra are furnished in two sets of damping values, the low and the high damping. The low damping values are 0.5, 1, 2, and 5 percent of critical. The high damping values are 7, 10, 15, and 20 percent of critical. The spectra are broadened by 115 percent to account for the uncertainties in the structural modeling techniques and material properties. 7.1.1.1.1.6.2 Stress Analysis The ANSYS computer program (stress pass option) is used to compute the force and moment resultants due to SRV and LOCA - related loads. A postprocessor program called SCALE is used to scan for the maximum absolute values of forces and moments in the circumferential and meridional directions. The forces and moments due to chugging and condensation oscillation loads are considered for the load combinations including the LOCA loads. The governing forces and moments from the six different frequencies are used in the stress analysis.

 /3 V

7.1.1.1.2 Seismic Loads Seismic loads constitute a significant loading in the structural assessment. The same seismic loads as those used in the initial building design are used. In that design, a dynamic analysis was made using discrete mathematical idealization of the entire structure using lumped masses. The resulting axial forces, moments, and shear forces at various levels due to the operating basis earthquake and the safe shutdown earthquake are used (FSAR Section 3.7). The effects of the seismic overturning moment and vertical accelerations are converted into forces at the elements. 7.1.1.1.3 Static and Thermal Loads The loads under consideration are the static loads (dead load and accident pressure) and temperature loads (operating and accident temperature) which are all axisymmetrical. , f

a. To analyze the above static loads, an in-house computer e- program, FINEL (FSAR Section 3.8.7), is used. Moments,

(g/ axial forces, and shear forces are computed by FINEL in an uncracked axisymmetric finite element containment model. 7.1-7

LGS DAR

b. The operating and accident temperature gradients are computed using ME 620 (FSAR Section 3.8.7) computer program (Bechtel 7rogram).

c. The results from a, b, and the hydrodynamic / seismic analysis are combined and applied to a containment element. The element contains data relative to rebar location, direction, and quantity and concrete properties. Within that wall element, force equilibrium and strain compatibility between the rebar and concrete is established by allowing the concrete to crack in tension. In this way, the stresses in the rebar and concrete are determined. The program used for this analysis is called CECAP (FSAR Section 3.8.7).

7.1.1.1.4 Load Combinations All load combinations from equations 1 through 7a as presented in Table 5.2-1 have been analyzed. The reversible nature of the structural responses due to the pool O dynamic loads and seismic loads is taken into account by considering the peak positive and negative magnitudes of the response forces and maximizing the total positive and negative forces and moments governing the design. Seismic and pool dynamic load effects are combined by conservatively summing the peak responses of each load by the absolute sum (ABS) method. Even though the square root sum of squares (SRSS) method is more appropriate because the peak effects of all loads may not occur simultaneously (Reference 7.1-4), the conservative ABS method is used in the design assessment of the containment and internal concrete structures to expedite licensing. 7.1.1.1.5 Design Assessment Material stresses at the critical sections in the primary containment and internal concrete structure are analyzed using the CECAP computer program. Critical sections for bending moment, axial force and shear in three directions are located throughout the containment structure. Liner plate is not considered as a structural element. The CECAP program considers concrete cracking in the analysis of reinforced concrete Rev. 1, 09/82 7.1-8

LGS DAR J sections. CECAP uses an iterative technique to obtain stresses considering redistribution of forces due to. cracking and, in the process, it reduces the thermal stresses due to the relieving effect of concrete cracking. The program is also capable of describing the spiral and transverse reinforcement stresses directly. The input data for the program consists of the uncracked forces, moments and shears calculated by FINEL, ANSYS, and seismic analysis. The loads are then combined in accordance with Table 5.1-1 with appropriate load factors. The stress margins are calculated in Section 7.2. 7.1.1.2 Reactor Enclosure and Control Structure l 7.1.1.2.1 Hydrodynamic Loads 7.1.1.2.1.1 Load Definitions The reactor enclosure and control structure were analyzed for both the SRV discharge load and the LOCA condensation oscillation (~}

 \-

and chugging loads. Description of the different load cases are presented in Section 7.1.1.1.1.5. 7.1.1.2.1.2 Hydrodynamic Analysis Models For the hydrodynamic loads described in Section 7.1.1.2.1.1, different mathematical models are constructed for the determination of the reactor enclosure and control structure hydrodynamic responses. The mathematical models are presented in detail in the following sections and are summarized in Table 7.1-1. 7.1.1.2.1.2.1 SRV Analysis Models The reactor enclosure and control structure were modeled to simultate global structural response during SRV actuation. Included in the analyses were an axisymmetric model for axisymmetric SRV loads and flexible base vertical, N-S, and E-W , stick models for the asymmetric SRV loads. The latter uses the ANSYS containment finite element model response as input. The mathematical models and analysis procedures are described below. O 7.1-9 Rev. 1, 09/82

LGS DAR 7.1.1.2.1.2.1.1 Axisymmetric SRV Analysis Model An axisymmetric model, based on Bechtel proprietary code CE971-FESS, was created to generate vertical response data for the NSSS new loads' structure and equipment adequacy assessment. The axisymmetric model has been closely correlated with in-plant test data (Reference 7.1-5). The model represents a containment system, adjacent structure (including reactor enclosure and control structure), and the soil medium as shown in Figure 7.1-3. Figure 7.1-8 shows a mass-proportional and stiffness-proportional damping simulation. The containment system and soil medium were modeled as FESS axisymmetric finite elements, whereas the adjacent structure was simulated by a coupled stick model. Altogether, the model has a combination of 673 dynamic degrees of freedom. The model was modified to simulate as-built conditions (i.e., concrete aging effect, etc) and normal plant operating conditions (i.e., RPV mass, etc) for generation of response data used for associated equipment adequacy evaluation. The analytical elements have the material properties as shown below: Element Young's Material Shear Material Modulus, E Density, p Poisson's Wave, Vs Type Kip /ft2 Kip.s2/ft* Ratio (Ft/s) Concrete 0.0936E+6* 0.00446 0.22 - Steel 0.4176E+7 0.01524 0.33 - Soil Medium 0.432E+6 0.00481 0.30 5950**

• The modulus represents a dynamic modulus of elasticity. h
  • The shear wave velocity, Vs, is used to simulate a soil shear mndulus (G=Vs zp), equal to 0.166 E+6 Kip /ft2, Rev. 1, 09/82 7.1-10

A LGS DAR V 7.1.1.2.1.2.1.2 Asymmetric SRV Analysis Models Analysis models for the asymmetric load include the combined use of the ANSYS finite element containment model response as input to the flexible base vertical, N-S horizontal and E-W horizontal stick models of the reactor enclosure and control structure. The l ANSYS containment model is shown in Figure 7.1-1, and the stick models are shown in Figures 7.1-4 and 7.1-5. The stick model damping uses the composite damping method (Reference 7.1-1). The vertical stick model was taken from the verified axisymmetric (FESS) coupled model. This model has 46 dynamic degrees of freedom. The flexible base was similated by a soil spring and a damper as recommended in the Bechtel design guide (Ref. 7.1-1). The N-S and E-W analytical stick models were similar to those used in the seismic analyses. Each stick model has 12 dynamic degrees of freedom. Input load data were taken from associated ANSYS containment {N) s_, analysis output data. This includes use of a vertical input time-history at the adjacent structure base equal to an average vertical response acceleration time history (from ANSYS) at the containment wall base, multiplied by an attenuation factor and use of horizontal input acceleration time history at the adjacent structure base equal to the gross motion generated from the associated containment ANSYS output data. 7.1.1.2.1.2.2 CO Analysis Model The reactor enclosure and control structure were modeled to simulate global structural response due to CO loads. Included in the analyses were an axisymmetric model for basic CO load case and CO-ADS load case, as was used in the axisymmetric SRV analysis described in Section 7.1.1.2.1.2.1.1. 7.1.1.2.1.2.3 CHUG Analysis Models The reactor enclosure and control structure were modeled to simulate global structural response during various CHUG events.

 ,- g Included in the time-history analyses were flexible base stick

, (j models presented in Section 7.1.1.2.1.2.1.2, which use the ANSYS containment model response as input for the CHUG asymmetric loads, and an axisymmetric model for the CHUG equivalent 7.1-11 Rev. 1, 09/82

LGS DAR axisymmetric loads. The mathematical models and analytical procedures are described below. 7.1.1.2.1.2.3.1 CHUG Asymmetric Analysis Models Analysis models for the CHUG asymmetric loads, as were used for SRV asymmetric loads, include the combined use of the ANSYS finite element containment model response as input to the flexible base vertical, N-S horizontal and E-W horizontal stick models of the reactor enclosure and control structure. The ANSYS containment model is shown in Figure 7.1-1, and the stick models are shown in Figures 7.1-4 and 7.1-5. The stick model damping used the composite damping method (Reference 7.1-1). The vertical stick model used was taken from the verified axisymmetric (FESS) coupled model. This model has 46 dynamic degrees of freedom. The flexible base was simulated by a soil spring and damper as recommended in the Bechtel design guide (Ref. 7.1-1). The N-S and E-W analytical stick models were the same as were used in the seismic analyses. Each stick model has 12 dynamic degrees of freedom. Input load data were taken from associated ANSYS containment analysis output data. This includes the use of a vertical input time history at the reactor enclosure and control structure base equal to an average vertical response acceleration time history (from ANSYS) at the containment wall base, multiplied by an attenuation factor, and the use of horizontal input acceleration time history at the reactor enclosure and control structure base equal to the gross motion generated from the associated containment ANSYS output data (no attenuation factor being used). 7.1.1.2.1.2.3.2 CHUG Axisymmetric Analysis Model An axisymmetric model, based on Bechtel proprietary code CE971-FESS, was created to generate vertical response data for the NSSS new loads' structure and equipment adequacy assessment. Similar to the axisymmetric SRV analysis model (Section 7.1.1.2.1.2.1.1) and the axisymmetric CD analysis model h (Section 7.1.1.2.1.2.2), CHUG axisymmetric analysis model represents a containment system, an adjacent structure (including Rev. 1, 09/82 7.1-12

LGS DAR {~ s m-reactor enclosure and control structure), and the soil medium as l shown in Figure 7.1-3. The containment system and soil medium were modeled as FESS axisymmetric finite elements, whereas the adjacent structure was simulated by a coupled stick model. The model has a combination of 673 dynamic degrees of freedom. The model was modified to simulate as-built conditions (i.e., concrete aging effect, etc) and normal plant operating conditions (i.e., RPV mass, etc), for generation of response data used for associated equipment adequacy evaluation. The analytical elements have the material properties as shown in the table of Section 7.1.1.2.1.2.1.1. 7.1.1.2.1.2.4 Control Structure Floor / Local Models Based on the excitation source at floor-wall junctions, analytical models for the selected floors were constructed to generate floor vertical response. Each floor considered was as a finite element model, with boundaries at walls simulated by 73 clamped edges. Along the line of symmetry (N-S direction), (x/ ) symmetric boundary conditions were imposed in the construction of a " half model" for the transient analyses, i.e., SRV and CHUG loads. To deal with dynamic problems of larger load duration, i.e., CO, CO-ADS, and seismic loads, a " quarter model" was formed by taking a symmetric line in the E-W direction of the " half model". Symmetric boundary were imposed similarly. The half model (Figure 7.1-6) consists of 42 model nodes and 30 quadrilateral elements. By choosing five dynamic degrees of freedom (DDOF) to each interior node and three DDOF to each symmetric node, the model has 115 DDOF. Similarly, the quarter model has 9 model nodes and 4 quadrilateral elements (Figure 7.1-7), with 12 DDOF selected for analysis. All floor models considered have identical nodal coordinates and similar model material properties (i.e., equivalent floor element thickness and mass density to take into account the beam-slab system action). Floor-supporting steel girders have contributed a substantial ( ()S portion of equivalent floor element thickness calculated for the slab-beam system. The contribution of the girders are different in magnitude, depending upon girder size and junction with or 7.1-13 Rev. 1, 09/82

LGS DAR l without shear connectors. The floors, except that El. 269 feet (control room), were built with shear of connectors. The floor model at El 269 feet (control room) was verified by data correlation / comparison with an in-plant test. In addition, the models were modified to simulate as-built conditions (e.g., concrete aging effect, etc). To deal with seismic events, the models were further modified to consider cracking effects. Floor model material properties are shown in Table 7.1-2. 7.1.1.2.1.3 Hydrodynamic Analysis 7.1.1.2.1.3.1 Analysis Procedures 7.1.1.2.1.3.1.1 Axisymmetric Analysis Procedure The axisymmetric analysis general procedure is to perform a time history analysis using equivalent axisymmetric input forcing vectors described in Sections 7.1.1.1.1.5.1 and 7.1.1.1.1.5.2, using Bechtel proprietary code CE971-FESS. Acceleration response spectra (ARS) data are generated using the acceleration response time histories obtained from the time history analysis using Bechtel proprietary code CE789-MSPEC. All associated ARS data are enveloped, widened 115 percent, and plotted, using Bechtel proprietary codes ENVLPS and MSPEC. 7.1.1.2.1.3.1.2 Asymmetric Analysis Procedure The general analytical procedure for asymmetric analysis consists of generating input load vectors to ANSYS model from appropriate use of the load definition and applying ANSYS transient response for asymmetric loadings to adjacent structure decoupled stick models (N-S, E-W, and vertical). A transient analysis is performed using decoupled BSAP stick models for each load case. The acceleration response spectra (ARS) data are generated using the response acceleration time histories and Bechtel proprietary code CZ789-MSPEC. All associated ARS data are enveloped, widened 15 percent, and plotted. O Rev. 1, 09/82 7.1-14

LGS DAR O 7.1.1.2.1.3.1.3 Floor / Local Model Analysis Procedure The floor model analysis general procedure is to perform a time history analysis using input forcing vectors taken from the output of stick model analyses described in Sections 7.1.1.2.1.3.1.1 and 7.1.1.2.1.3.1.2 and using the model according to Bechtel proprietary code CE800-BSAP. ARS data are developed using the acceleration response time histories and Bechtel proprietary codes CE789-MSPEC and ENVLPS. 7.1.1.2.1.3.2 Generation of Response Data 7.1.1.2.1.3.2.1 Acceleration Response Spectra Data 7.1.1.2.1.3.2.1.1 SRV ARS Data Two sets of ARS data were generated. One set is for SRV

'    axisymmetric analysis and the other set is for SRV asymmetric analysis. The ARS data, enveloped from associated data and (A-]- broadened 115 percent at peak frequencies, represent global response, applicable to structural assessment and NSSS equipment (or other safety-related equipment) adequacy evaluations located at or near the adjacent structure walls and/or columns. The ARS at selected typical locations on the reactor enclosure and control structure are presented in Appendix B.

7.1.1.2.1.3.2.1.2 CO ARS Data Two sets of ARS data are generated. One set is for basic CO load case analysis and the other set is for the CO-ADS load case. l Again, the ARS data, enveloped from associated data and broadened 115 percent at all peak frequencies, represent global response. The data are applicable to structure and/or equipment adequancy assessment located at or near the adjacent structure walls and/or columns. The ARS at selected locations are presented in Appendix B. 7.1.1.2.1.3.2.1.3 CHUG ARS Data (~$g Two sets of broadened ARS data are presented in Appendix B for (_,/ appropriate use in structure and equipment adequacy assessment. Set one is for CHUG asymmetric analysis case, and set two is for the CHUG equivalent axisymmetric analysis case. 7.1-15 Rev. 1, 09/82

LGS DAR The ARS data for the CHUG asymmetric case were developed and plotted similar to the SRV asymmetric analysis case. The data picts include the broadened ARS data in the three global directions (vertical, N-S, and E-W axes). The CHUG asymmetric vertical ARS data provide responses for the applicable areas for the NSSS equipment adequacy assessment. The N-S and E-W ARS data apply to all NSSS equipment situated in any location of the reactor enclosure and control structure. The ARS data for CHUG equivalent axisymmetric analysis cases were developed and plotted similar to SRV axisymmetric analysis cases. Again, the data represent only gicbal response, applicable to the NSSS equipment adequacy evaluations located at or near the adjacent structure wal'.s and/or columns. Local / floor models are required for generating vertical ARS data for some floor-mounted equipment. 7.1.1.2.1.3.2.1.4 Hydrodynamic Local ARS Data The local ARS data in the control structure were generated based on the floor / local model analytical procedure described in Sections 7.1.1.2.1.2.4 and 7.1.1.2.1.3.1.3. The data was broadened i15 percent at all peaks of the data enveloped from associated dynamic events. The hydrodynamic events considered in the enveloping were SRV, CHUG, CO (basic), and CO-ADS. The hydrodynamic Iccal ARS data are used for the structures, components, and floor-mounted equipment where the global ARS data are not applicable. 7.1.1.2.2 Seismic Loads The seismic analysis methodology is discussed in FSAR Section 3.7.2.1. A seismic local model (Section 7.1.1.2.1.2.4) was developed to generate local ARS data for the floors of the control structure. Rev. 1, 09/82 7.1-16

O LGS DAR V 7.1.1.2.3 Static Loads The static loads are discussed in FSAR Section 3.8.4.3. 7.1.1.2.4 Load Combinations All individual loads for concrete structures are combined with the appropriate load factors, as shown in Table 5.2-1, for analysis of all loading combinations. Steel structures are checked for the load combinations listed in Table 5.3-1. 7.1.1.2.5 Design Assessment Critical sections for bending moment, axial force, and shear in all three directions are located throughout the reactor enclosure l and control structure. Design capability at the critical [N-)/ sections is determined, and then the design capability is compared with the actual forces and moments acting on the sections under all the load combinations. This comparison yields design margins. The design margins are discussed in Section 7.2. 7.1.2 STRUCTURAL STEEL AND ASME CLASS MC STEEL COMPONENTS ASSESSMENT METHODOLOGY 7.1.2.1 Suppression Chamber Columns There are 12 suppression chamber columns, which are 42-inch diameter pipe with 1-1/4 inch wall thickness. The columns are attached at the underside of the diaphragm slab at El. 234 ft-2 in. and at the basemat at El. 181 ft-11 in. 7.1.2.1.1 Structural Models l The columns were independently analyzed for static and dynamic loads. The analytical methods used for nonhydrodynamic loads () ( ,e such as dead, live, pressure, temperature, seismic, and pipe rupture loads are described in FSAR Section 3.8.3.4.5. 7.1-17 Rev. 1, 09/82

LGS DAR To deal with dynamic effects from seismic and hydrodynamic events, two analytical approaches were used. The ANSYS containment model (Section 7.1.1.1), in which the columns were also modeled, was used for LOCA load cases. For seismic and SRV loads, the BSAP beam model (Figure 7.1-17) was used. The beam model has 13 beam elements and 14 nodes, with effective water mass in the submerged portion. The column ends were modeled as clamped edges. 7.1.2.1.2 Loads The columns, partially submerged in the suppression pool, are subjected to direct pressure loads from air bubble oscillation, etc, and inertia loads due to building response (or movement) from dynamic loads (seismic and hydrodynamic). Thermal loads are induced due to the rise of temperature during hydrodynamic LOCA events. 7.1.2.1.2.1 SRV Discharge Loads The SRV discharge p. essure-time histories are considered as acting on the submerged portions of the columns. The inertia forces from building response due to SRV discharge load are included by using the response spectra shown in Appendix A. 7.1.2.1.2.2 LOCA Related Loads The manner in which the LOCA related loads are applied to the column is the same as described for SRV loads in Section 7.1.2.1.2.1. 7.1.2.1.2.3 Seismic Loads The seismic loads on the column were obtained by response spectrum method. The response spectra used are developed for OBE and SSE as described in FSAR Section 3.7. O Rev. 1, 09/82 7.1-18

LGS DAR g]) b 7.1.2.1.2.4 Static Load l Static loads, including dead load and thermal load, were considered in the column analysis. 7.1.2.1.2.5 Load Combinations l The load combinations and allowable stresses are in accordance with Section 5.3. Member forces and moments obtained from dynamic loads are combined by the SRSS method. The resulting combined dynamic loads are combined with the static loads by the absolute sum technique. 7.1.2.1.2.6 Design Assessment l The combined stresses due to axial force and bending moment were calculated and compared with allowable stresses. lO

 '-              Downcomer Bracing 7.1.2.2 The following covers the methodology used in the assessment of the bracing system at EL. 203' - 5" in the primary containment suppression pool.

7.1.2.2.1 Bracing System Description The downcomer bracing system is designed as a two-dimensional truss system to provide horizontal rupport for 87 downcomers, 14 MSRV discharge lines, and other miscellaneous piping in the suppression pool. The bracing system is supported vertically by the 87 downcomers and at 12 anchor points around the RPV pedestal wall. The bracing system is made of stainless steel members connected to carbon steel collars at the downcomers and embeddment plates at the pedestal wall by high-strength stainless steel bolts. The bracing members consist of 10-inch and 12-inch diameter schedule 160 pipe sections, and 3-1/4 inch end l connection plates. The bracing system is designed in accordance ! with Reference 7.1-10. The bracing system layout and typical connection details are shown in Figures 7.1-9 and 7.1-10. The mathematical model used l l 7.1-19 Rev. 1, 09/82

LGS DAR I in the bracing system is presented in Figure D.2-10 of Appendix D. 7.1.2.2.2 Loads The bracing system is assessed for all plant operation induced loads described below. The basis for all hydrodynamic loads considered in the analysis is presented in Chapter 4. 7.1.2.2.2.1 SRV Discharge Loads Discharge through the SRV discharge pipe creates horizontal as well as vertical loading on the bracing system due to unbalanced pressures. The horizontal (lateral) load is considered as acting on the downcomers and the SRV discharge pipes. The vertical load is considered acting on the bracing members alone. These loads are applied to the bracing system by considering them as equivalent static loads using a dynamic magnification factor which is obtained from the dynamic analysis of the downcomer, as described in Section 7.1.4. The SRV discharge also induces hydrodynamic forces in the containment structure. Inertial forces of the bracing system,

due to the response of the containment structure, are considered l using hydrodynamic reponse spectra of the containment structure shown in Appendix A.

1 The lateral loads and the containment structure response form the complete SRV discharge load set on the bracing system. 7.1.2.2.2.2 LOCA Related Loads Loss-of-coolant accidents are characterized by several phenomena that result with non-concurrent loadings on the bracing system as described in Section 4.2. These hydrodynamic loads induce accelerations of the containment structure, which in turn induce additional loads on the bracing system. These loads are obtained from the hydrodynamic acceleration response spectra shown in Appendix A. In addition, the LOCA event induces lateral forces on the i submerged portion and tip of downcomers. The loads include drag loads, pressure loads, and chugging tip load. The hydrodynamic Rev. 1, 09/82 7.1-20

i LGS DAR analysis of a single downcomer for the lateral loads is presented in Section 7.1.4. The resulting reaction forces at the bracing support are applied as equivalent static load in accordance with section 3.1 of Reference 7.1-6. 7.1.2.2.2.3 Seismic Loads The forces due to the seismic accelerations of the downcomers, the SRV lines, and the bracing members are obtained by analysis of these structures using the response spectra developed for OBE and SSE as described in FSAR Section 3.7.2. 7.1.2.2.2.4 Static Loads The dead load of the bracing members is considered with allowance for buoyancy.

     -      7.1.2.2.2.5     Thermal Load s   J The operating and accident temperature considered is 90 and 2100F, respectively. The reference temperature of the system is assumed to be 600F.

7.1.2.2.2.6 Load Combinations l The load combinations and allowable stresses are described in Table 5.3-1. Although the loads on the bracing system under consideration act in random horizontal directions, each individual, load is applied on the system in the worst possible direction to find the maximum resultant forces. I 7.1.2.2.3 Design Assessment The two-dimensional truss model of the bracing system is analyzed for the static, thermal, and equivalent static hydrodynamic loads using the computer program STRUDL. The containment structure inertial load is analyzed for seismic and hydrodynamic responses using the computer program ANSYS. The bracing member forces due l O to the various loading conditions are combined by the absolute sum method and assessed for the conditions specified in Table 5.3-1. l 7.1-21 Rev. 1, 09/82

LGS DAR 7.1.2.3 ASME Class MC Steel Components The assessment methodology used for hydrodynamic loads on MC components will be provided later. 7.1.3 LINER PLATE ASSESSMENT METHODOLOGY FSAR Section 3.8.1.1.2 provides a description of the containment liner plate and its anchorage system. The analysis and design of the liner plate anchorages for nonhydrodynamic loads is in accordance with Reference 7.1-7. For the analysis of the liner plate and anchorages for hydrodynamic suction pressure loads, the contributing load on the liner is that due to the net negative pressure load. The net negative pressure load is determined from the dynamic negative pressure due to SRV actuation and/or LOCA chugging minus the static positive pressure due to the wetwell hydrostatic pressure and/or LOCA wetwell pressure. Figures 7.1-12 through 7.1-15 describe the loads on the suppression chamber liner plate for the normal and abnormal load conditions. For the normal condition, the hydrostatic pressure on the base mat liner is 10.4 pst (positive) and the maximum negative pressure due to the actuation of all SRVs is 11.67 psi (negative). The distribution of these pressures on the suppression chamber wall is shown in Figure 7.1-12. The maximum net negative pressure is 1.27 psi (negative). For the abnormal condition, the total positive pressure on the basemat liner is 35.4 psi which consists of 10.4 psi (positive) from hydrostatic pressure plus 25.0 (positive) from a small or intermediate break LOCA (Figure 7.1-13). The total maximum negative pressure on the basemat liner is 16.9 psi (negative) due to the axisymmetric chugging and SRV loads (Figure 7.1-14). The maximum negative pressures from SRV actuation and chugging are combined for conservatism. It is recognized that the probability of these two phenomena producing peak negative pressures at the same time is very low. The combined pressure distribution due to hydrostatic, LOCA, SRV, and chugging is shown in Figure 7.1-15. O Rev. 1, 09/82 7.1-22

/7 LGS DAR V 7.1.4 DOWNCOMER ASSESSMENT METHODOLOGY 7.1.4.1 Structural Model There are 87, 24-inch OD, steel pipe downcomers running vertically down from the diaphragm slab. The downcomers are embedded in the diaphragm slab and extend downward to El. 193'-11", which is approximately 12 feet below high water level, as shcwn in Figure 1.4-2. All downcomers are supported laterally at El 203 '-5 by the downcomer bracing system. Any vertical loads are transmitted by the bracing system to the downcomers and therefore to the diaphragm slab. The structural model considers the downcomer as a vertical pipe fixed at the underside of the diaphragm slab with a spring in the horizontal direction at bracing level. This model is shown in Figure 7.1-16. The inertial effect of the water in the submerged portion of the downcomer (12 feet) was approximated by the addition of a equivalent mass of water lumped at the appropriate

,_s  nodal points. The model is evaluated for three rpring values for a representative support stiffness provided by the bracing system

(\_)s to the downcomers. The bracing spring is set to 50 k/in, 50 k/in, and 15000 k/in to represent the tangential mode, the radial mode, and rigid response of the bracing system. 7.1.4.2 Loads The downcomer is subjected to static and dynamic loads due to normal, upset, emergency, and faulted conditions. Loading cases and combinations are described in Table 5.5-1. The basis for all hydrodynamic loads considered in the analysis is presented in Chapter 4. 7.1.4.3 Analysis Downcomers are analyzed for the specified loading conditions using the Bechtel computer program BSAP. The downcomers are J analyzed for both the hydrodynamic loads acting directly on the submerged portions and the inertial forces due to containment responses to the hydrodynamic and seismic loads. () The hydrodynamic load analyses, due to SRV discharge and LOCA related loads acting on the submerged portion of the downcomers, are performed using the mode-superposition time history 7.1-23 Rev. 1, 09/82

LGS DAR technique. The seismic and hydrodynamic load analyses, due to containment responses, are performed using the response-spectrum analysis procedure. Damping values used are equal to 2 percent of critical for OBE and SRV loads, and 7 percent of critical for SSE and LOCA loads. 7.1.4.4 Design Assessment The resultant stresses in the downcomers due to the load combinations described in Table 5.5-1 are compared with the allowable stresses in accordance with the criteria given in Reference 6.4-2. 7.1.4.5 Fatique Evaluation Of Downcomers In Wetwell Air Space A fatigue analysis of the downcomers was conducted in accordance with ASME Section III, Division 1 (1979 Summer Addendum), subsection NB-3650. Only that portion of the downcomer in the air space of the suppression chamt er need be evaluated for fatigue. Figures D.2-8 and D.2-9 of Appendix D show the number of cycles considered and the load histogram, respectively. l 7.1.5 PIPING AND SRV SYSTEMS ASSESSMENT METHODOLOGY The piping and SRV systems will be analyzed for the load combinations described in Table 5.6-1 using Bechtel computer programs ME101 and ME632. These programs are described in FSAR Section 3.9. Static and dynamic analysis of the piping and SRV systems are performed as described in the paragraphs below. Static analysis techniques are used to determine the stresses due to steady state loads and/or dynamic loads having equivalent static loads. Response spectra at the piping anchors are obtained from the dynamic analysis of the containment subjected to LOCA and SRV loading. Piping systems are then analyzed for these response spectra following the method described in Reference 7.1-8. Time history dynamic analysis of the SRV discharge piping subjected to fluid transient forces in the pipe due to relief valve opening is performed using Bechtel computer code ME632. Rev. 1, 09/82 7.1-24

("N LGS DAR b) 7.1.6 NSSS ASSESSMENT METHODOLOGY To be provided later. 7.1.7 EQUIPMENT ASSESSMENT METHODOLOGY Safety-related equipment located within the containment and the reactor enclosure and control structure are subjected to hydrodynamic loads due to SRV and LOCA discharge effects principally originating in the suppression pool of the containment structure. The equipment and equipment supports are assessed to verify their adequacy to withstand these hydrodynamic loads in combination with seismic and all other applicable loads in accordance with the load combinations given in Table 5.8-1. 7.1.7.1 Hydrodynamic loads 7.1.7.1.1 SRV Discharge Loads Ov Loadings associated with the axisymmetric and asymmetric SRV discharges are described in Chapters 3 and 4. Acceleration response spectra at the various elevations where the equipment are located have been generated for all appropriate pressure history traces (Figures 4.1-25 through 4.1-27) for damping values of 1/2, 1, 2 and 5 percent. These have been enveloped into a single curve for each of the above damping values. Such enveloped curves are generated for each of the N-S, E-W and vertical directions. These curves form the basis for the SRV loads for equipment assessment. 7.1.7.1.2 LOCA Related Loads Loadings associated with loss-of-coolant accident (LOCA) are described in Chapters 3 and 4. The various LOCA loadings considered include condensation oscillation and chugging (Section 4.2.2). Acceleration response spectra at various elevations where the equipment are located have been generated for the above LOCA loads for damping values of 1/2, 1, 2 and 5 percent. These have been enveloped into a single curve for each of the above damping values. Such enveloped curves are generated (} \s - for each of the N-S, E-W and vertical directions. These curves form the basis for the LOCA loads for equipment assessment. 7.1-25 Rev. 1, 09/82

LGS DAR 7.1.7.2 Seismic Loads The details of seismic input and seisrcic loads are discussed in FSAR Section 3.7. The effects of both operating basis earthquake l (OBE) and safe shutdown earthquake (SSE) are considered. These  ! loads are provided in the form of acceleration response spectra at each floor for damping values of 1/2, 1, 2, and 5 percent for each of N-S, E-W and vertical directions. 7.1.7.3 Other Loads In addition to hydrodynamic and seismic loads, other loads such as dead loads, live loads, operating loads, pressure loads, thermal loads, nozzle loads and equipment piping interaction loads, as applicable, are also considered. 7.1.7.4 Qualification Methods The adequacy of the design of the equipment is assessed by one of the following: f

a. Dynamic analysis

b. Testing under simulated conditions

c. Combination of testing and analysis.

The choice is based on P.he practicality of the method depending upon function, type, siae, shape, and complexity of the equipment and the reliability of the qualification method. In general, the requirements outlined in Reference 7.1-9 are followed for the qualification of equipment. 7.1.7.4.1 Dynamic Analysis 7.1.7.4.1.1 Methods and Procedures Rev. 1, 09/82 7.1-26

p LGS DAR V The dynamic analysis of various equipment is classified into three groups according to the relative rigidity of the equipment based on the magnitude of the fundamental natural frequency described below.

a. Structurally simple equipment - comprised of that equipment which can be adequately represented by one degree of freedom system.

b. Structurally rigid equipment - Comprised of that equipment whose fundamental frequency is:

1) greater than 33 Hz for the consideration of seismic loads, and,

2) greater than 100 Hz for the consideration of hydrodynamic loads.

Structurally complex equipment - Comprised of that equipment (~ s-- ') c. which cannot be classified as structurally simple or structurally rigid. When the equipment is structurally simple or rigid in one direction but complex in the other, each direction may be classified separately to determine the dynamic loads. The appropriate response spectra for specific equipment are obtained from the response spectra for the elevation at which the equipment is located in a building for OBE, SSE, and hydrodynamic loads. This includes the vertical as well as both the N-S and E-W horizontal directions. For equipment that is structurally simple, the dynamic loading (either seismic or hydrodynamic) consists of a static load corresponding to the equipment weight times the acceleration selected from the appropriate response spectrum. The acceleration selected corresponds to the equipment's natural frequency, if the equipment's natural frequency is known. If the equipment's natural frequency is not known, the acceleration selected corresponds to the maximum value of the response

  ~]    spectra.

(V 7.1-27 Rev. 1, 09/82

LGS DAR For equipment that is structurally rigid, the seismic load consists of a static load corresponding to the equipment weight times the acceleration at 33 Hz, selected from the appropriate response spectrum and the hydrodynamic loading consists of a static load corresponding to the equipment weight times the acceleration at 100 Hz, selected from the appropriate response spectrum. For the analysis of structurally complex equipment, the equipment is idealized by a mathematical model that adequately predicts-the dynamic properties of the equipment, and a dynamic analysis is performed using any standard analysis procedure. An acceptable alternative method of analysis is by static coefficient analysis for verifying structural integrity of frame type structures such as members physically similar to beams and columns that can be represented by a simple model. No determination of natural frequencies is made, and the response of the equipment is assumed to be the peak of the response spectrum at damping values in accordance with Section 7.1.7.4.1.2. This response is then multiplied by a static coefficient of 1.5 to take into account the effects of both multifrequency excitation and multimode response. 7.1.7.4.1.2 Appropriate Damping Values O The following damping values are used for the design assessment:

a. Load combinations involving OBE but not hydrodynamic loads -1/2%

b. Load combinations involving SSE but not hydrodynamic loads -1/2%

c. Load combinations involving hydrodynamic loads, or seismic and hydrodynamic loads - 2%

Higher damping values may be used where justified. 7.1.7.4.1.3 Three Components of Dynamic Motions The responses such as internal forces, stresses, and deformations h at any point from the three principal orthogonal directions of the dynamic loads are combined as follows. Rev. 1, 09/82 7.1-28

p LGS DAR V The response value used shall be the maximum value obtained by adding the response due to vertical earthquake with the larger value of the responses due to one of the horizontal earthquakes by the absolute sum method. For the other dynamic loads, the response value shall be obatained by combining the response due to three orthogonal directions of an individual load by the square root of the sum of the squares (SRSS) method. 7.1.7.4.2 Testing In lieu of performing dynamic analysis, dynamic adequacy is established by providing dynamic test data. Such data must conform to one of the following:

a. Performance data of equipment that has been subjected to equal or greater dynamic loads (considering appropriate frequency range) than those to be experienced under the

() specified dynamic loading conditions,

b. Test data from comparable equipment previously tested under similar conditions that has been subjected to equal or greater dynamic loads than those specified.

c. Actual testing of equipment in operating conditions simulating, as closely as possible, the actual installation, the required loadings and load combinations.

A continuous sinusoidal test, sine beat test, or decaying sinusoidal test is used when the applicable floor acceleration spectrum is a narrow band response spectrum. Otherwise, random motion test (or equivalent) with broad frequency content is used. The equipment to be tested is mounted in a manner that simulates the actual service mounting. Sufficient monitoring devices are used to evaluate the performance of the equipment. With the appropriate test method selected, the equipment is considered to be qualified when the test response spectra (TRS) envelopes the required response spectra (RRS) and the equipment does not (~N malfunction or fail. A new test does not need to be conducted if (_,) equipment requires only minor modifications such as additional bracings or change in switch model, etc, and if proper 7.1-29 Rev. 1, 09/82

LGS DAR justification is given to show that the modifications would not jeopardize the strength and function of the equipment. 7.1.7.4.3 Combined Analysis and Testing There are several instances where the qualification of equipment by analysis alone or testing alone is not practical or adequate because of its size, or its complexity, or large number of similar configurations. In these instances, a combination of analysis and testing is the most practical. The following are general approaches:

a. An analysis is conducted on the overall assembly to determine its stress ' level and the transmissibility of motion f rom the base of the equipment to the critical components. The critical components are removed from the assembly and subjected to a simulation of the environment on a test table.

b. Experimental methods are used to aid in the formulation of the mathematical model for any piece of equipment. Mode shapes and frequencies are determined experimentally and incorporated into a mathematical model of the equipemnt.

7.1.8 ELECTRICAL RACEWAY SYSTEM ASSESSMENT METHODOLOGY To be provided later. 7.1.9 HVAC DUCT SYSTEM ASSESSMENT METHODOLOGY To be provided later. 7.1.10 REFERENCES 7.1-1 " Seismic Analyses of Structures and Equipment for Nuclear Power Plants," BC-TOP-4A, Bechtel Power Corporation, November 1974. 7.1-2 Wilson, E. L, "A Computer Program for the Dynamic Stress Analysis of Underground Structures," USAEWES, Control Report 1-175, January 1968. Rev. 1, 09/82 7.1-30

i 1 LGS DAR I i 7.1-3 Desai and Abel, " Introduction to the Finite Element l Method," Van Nostroid Reinold Co., 1972 7.1-4 " Technical Bases for the Use of SRSS Method for 1 Combining Dynamic Loads for Mark II Plants," NEDE-24010-P, General Electric Co, July 1977. 7.1-5 SRV In-Plant Test Report 7.1-6 Davis, W. M., "MK II Main Vent Lateral Loads Summary Report," NEDE-23806-P, General Electric Co., October 1978. 7.1-7 T. E. Johnson, et al., " Containment Building Liner Plate i Design Report," BC-TOP-1, Bechtel Corporation, San Francisco, December 1972. 1 7.1-8 " Seismic Analysis of Piping Systems," BP-TOP-1, j [ Revision 2, Bechtel Power Corporation, San Francisco,

              January 1975.

7.1-9 IEEE Standard 344-1975, " Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations." 7.1-10 American Institute of Steel Construction, Manual of Steel Construction, 7th Edition, 1970. i 1 l i i 7.1-31 Rev. 1, 09/82

SUMMA) l LOAD l 1 CASE j i L l MODEL j l(FIGUhE NO.) lj i i i I l Axisymmetric l l " FESS" Vertical l l Coupled Mcdel l l (Fig. 7.1-3) l l 1 1 l l Vertical l j Flexible Base l l Stick Model l l (Fig. 7.1-4) l l 1 i i l Horizontal j l tiexible Base l Stick Model l (Fig. 7.1 - 5) l l- 1 1 I l Ccntrol l l Structure Floor j j IIalt Mcdel l l (Fig. 7.1-6) l l 1 1 1 l Control l j Structure Floor l l Luarter Model l l (Fig. 7.1-7) l 1 1 ~

O l LGS DAR TABLE 7.1-2 CONTROL STRUCTURE FLOCR MODEL MATERIAL PROPERTIES

Equiv. Floar Element (1)

Slab Thickness Thickness, t Iloor Element

.         Control structure                     t                                          eff             Mass Density p' Floor Elevation Ift)                    c (ft)                      (ft)                         Kip.S2/ft*

l 1.25 2.66 .002554 El. 217 2.125, 2.5 3.26 .003334 El. 239 1.0 2.93 .003241 i El. 253 1. 0 2.61 .002538 l l El. 269 1.5 2.63 .003219 El. 289 1. 5 2.96 .002610 El. 304 1.0 2.50 .002145

!         El. 331                               2,   1.5                      3.595                           .0040821 2.965                           .0044573 i

i (1) Equivalent floor element thickness and mass density p' to take into account the beam-slab system action. I

 ;                                                                                                        Rev. 1, 09/82 I

i i

_ p - (TYPICAL) E L. 410'

                                     "

• O' YPICAL) t- J 68.25' ~'

(^\ 58' ' 12.5' 7 j 0,8 l i: 40.25' I l o 13 12 9 _ < y E L. 352' Z 19'34 28 h f. .a p.jI29 EL 333' f j 33 32 21 19.5'

           "                        2#

35 4 43 R 313'__ _L 564>. E L 304' 61.95* 57 21' 61 9 60 59 58 9 , E L 283' L.a>+3p.J 76 Q, p 774> f EL 269' 82 ' 78 I 6, 81 80 79 >y

                    -                       {- 0+4                              g 103          34, E L 239' jg   ,      , , ,

m iO4 m I:: 105 22' 108 107 1059 um RIGID MEMBER

     '89.55'        -                   -

gy+ o.J { E L 217' o PIN CONNECTION

                    -                   -                         127           '

129 j 6' e NOOAL POINT W/ MASS

9 "

EL 201'

                    ~g           =g- a+<            130 p                                  e NODAL POINT W/OUT MASS
                    --+

o +- 128

 /                                                                                 ,

(% ~ n o o 9i 154 .159 h EL 177'

       -h,h                                                                                                                 >Y kV I

it h

         @E I gy TYPICAL 1     DAMPER n

m I & l l

h. _ _._ __ _ , _ _ . _ _._ .. _.

                                                                                                                  'l
                                          ~

I~ 6@S' ~ l ~ 4@5' l ' 8@5' ~ l iO'l ~ 20' ~l 30' ~l~ 30' LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT i f)

 \/                                                                                                       REACTOR ENCLOSURE AND CONTROL STRUCTUR E VERTICAL AXISYMMETRIC COUPLED MODEL (FESS)

FIGURE 7.1-3 R E V 1,09/82

25' 2

 /                                                          l 3. (TYPIC AL) u  _

40' (TYPICAL) 6.25' TYPICAL 58' w+12 5 ';  ; 68 25' r-3 16 7 2 6. 1 5 21 8% g a 9 fg 'O 21 i l 4 >O ,, 3 , m cd

   ]         -,91k3         ,,0               >--@

19 5' 20 h

              @         24 23               2t t            =            '

25 6 26 @ 27 t 9" Q4> ' tp .. 304' 2s o 33 h 4>29 2,. LEGEND azz RIGID MEMBER p f~ ~ 32 8 f, 7

                                           'A To- '

b 283' O PIN CONNECTION 9 NOOAL POINT W/M ASS 34 i4 e NODAL POINT W/OUT M ASS 4 269' @ BEAM ID NO. Qf-- , do g 3. 3. h TRUSS ID NO.

             )

S , g NODAL POINT ID No.

             ' igi>           0,            dh7--h                          253'
           ~

3

                                              @4p* '-            14-4 k---

42 h 239' h 43 22' 2 11 46 (k 45

                                          '4 k -.
                                              "4      @           

217' 47 so 4k: " O AF -- 29 o 201' 4e as b b LIMERICK GENERATING STATION UNITS 1 AND 2

                    $2 @ 53              51 DESIGN ASSESSMENT REPORT f1,1:a1-nh                                                     177' REACTOR ENCLOSURE AND
                         '""                                                                   CONTROL STRUCTURE VERTICAL STICK MODEL FIGURE 7.1-4                REV 1, 09/82

i

                                                                                       ^r_^-_     _*r_        -^

ELEVATION 12 -- 410' r i

                                           -b                                                                  352' ELEVATION          11 332'                                                                                     333'   I 9'I         l                        M 313' 304'
                                                    ],                                                               f
                                                                              }\

289' [ 8

                                          ;;        /

1 k\ 283' 269' / '

                                                 //                              \\

j l i 254' j 253' o 237' - lf \\

                                /          ]                                          !

200* - 2 N -S ___ DIR ECTION g:  ; 177' s (Y YIII)

                                  ///////f/

INPUT MOTION - ACCELERATION TIME HISTORY FROM E-W DIRECTION 3-D ANSYS MODEL ANALYSIS (ASYMM.) l LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O REACTOR ENCLOSURE AND CONTROL STRUCTURE HORIZONTAL STICK MODEL FIGURE 7.1-5 R EV 1,09/82

C O O N b U 20 21 1. 22 23 24 25 26 l NORTH-SIDED WALL N n i i l

                                                                                                                                           ~

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FIGURE 7.14 R EV 1,09/82

O O l LIMERICK GENER ATING STATION UNITS 1 AND 2 ' ! DESIGN ASSESSMENT REPORT DELETED l FIGURE 7.111 R E V.1,09/82

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o 44 > h l 3< > 1f 2 l 5 ' 1 1I tirr H7 ORIGIN i LIMERICK GENERATING STATION UNITS 1 AND 2 , DESIGN ASSESSMENT REPORT O SUPPRESSION CHAMBER COLUMN ANALYTICAL MODEL i FIGURE 7.117 R E V.1, 09/82

1 I i LGS DAR O, 7.2 DESIGN CAPABILITY MARGINS This section describes the design margins for structures, piping, and equipment resulting from the LGS design assessment which uses the methods of Section 7.1 7.2.1 STRESS MARGINS Stresses at the critical sections for all of the structures, piping, and equipment described in Section 7.1 are evaluated for the loading combinations presented in Chapter 5. The stress margin (SM) in percent is defined as follows: l SM = (1 - SR) x 100 l where SR represents the stress ration. SR is calculated by dividing the factored stress (C f ) by the associated stress n n allowable (F ) or, mathematically, [ SR = I (C f /F ) nn n 7.2.1,1 Containment Structure The detailed results from the structural assessment of the containment structure are summarized in Appendix D.1. Figure D.1-1 shows the design sections in the basemat, shield walls, containment walls, reactor pedestal, and the diaphragm slab that were considered in the structural assessment. Figures D.1-2 through D.1-25 give the calculated maximum design stresses for i the load combinations listed in Table 5.2-1. l Both rebar stresses and concrete stresses are calculated based on the applicable load combination equations. The stresses in the drywell wall are calculated at design sections 1 to 5 and are tabulated in Figures D.1-2 through D.1-5. The stresses in the wetwell wall are calculated at design sectionc 6 to 11 and are tabulated in Figures D.1-6 through D.1-9. The stresses in the

    . shield wall are calculated at design sections 12 and 13 and are tabulated in Figures D.1-10 and D.1-11, respectively. The RPV

(' pedestal stresses are calculated at design sections 14 to 20 and are tabulated in Figures D.1-12 through D.1-16. The stresses in 7.2-1 Rev. 1, 09/82

LGS DAR the diaphgram slab are calculated at design sections 21 to 25 and are tabulated in Figures D.1-17 through D.1-20. The stresses in the basemat are calculated at design sections 26 to 30 and are tabulated in Figures D.1-21 through D.1-25. The containment assessment is summarized as follows:

a. The calculated stress level is very low for load combination equation 1 (an operating condition), i.e.,

rebar stresses are far less than 20 ksi.

b. The maximum rebar stress is predicted as 53.9 ksi at design sections 6 and 11, located in the wetwell vertical direction. The magnitude is within the rebar stress allowable (0.9 Fy = 54 ksi).

c. In general, rebar stresses and concrete compressive stresses are within stress allowables.

7.2.1.2 Reactor Enclosure and Control Structure Results of the structural assessment of the reactor enclosure and control structure are summarized in Appendix E. Figures E.1-1 through E.1-21 show the selected structural elements and sections where stresses were calculated. Appendix E contains tabulations of predicted stresses, stress allowables, and design margins for critical loading combinations considered. The sections selected for assessment were considered to be the most critical based on previous seismic calculations. The critical load combinations are tabulated considering critical locations / sections related to reactor enclosure and control structure shear walls, foundations, floor slabs and supporting steel, steel platforms, and floor support columns. Emphasis is placed on margins of principal resisting structural elements, with reinforcing bar stresses for reinforced concrete structures and axial and/or bending stresses for steel structures. Rev. 1, 09/82 7.2-2

LGS DAR Also included in Appendix E are diagrams of axial forces, N-S shear forces, N-S overturning moments, E-W shear forces, E-W overturning moments for reactor enclosure and control structure as shown in Figures E.1-22 through E.1-31. The reactor enclosure floor system stress margins were calculated for both slabs and floor support steel beams, including floors at El. 201, 217, 253, 283, 313, 333, and 352 ft. Calculated slab stress levels were generally governed by either Equation 1 or 7a , of Table 5.2-1. The highest reinforcing bar stress was found at

the floor of El. 253 ft, having a stress intensity of 51.26 ksi and an associated stress margin of approximately 5 percent.

Figure E.1-32 shows rebar stresses and related stress margins of the aforementioned floors. In addition, the stresses and related stress margins of floor support steel beams are presented in Figure E.1-33. The governing equations were Equations 1 and 7 of 1 Table 5.3-1. Stress levels were generally low.  ! In the case of reactor enclosure support columns, load combination 7 of Table 5.3-1 governs the column stress

interaction. Stress interaction calculations were performed and show that columns were generally understressed (Figure E.1-34).

, The column at column lines 30.5 and E of El. 217 to 253 ft has a 1 fully stressed situation. The reactor enclosure shear wall sections close to the base (El. 177 ft) were assessed as shown in Figure E.1-35. The highest stress conditions occurred in the walls of column lines , 14.1 (west wall) and 31.9 (east wall) due to shearing effect at the base. The corresponding stress margin was approximately 1 l percent. The floor system of the control structure, including the concrete slabs and their supporting steel beams, are shown in Figure E.1-9  ; through E.1-17, while the stress margins are listed in Figures 1 l E.1-36 and E.1-37. In general, none of those selected critical sections were found overstressed in the control structure. All concrete floors were assessed. The concrete slabs are governed by the normal load conditions, Equation 1 of Table 5.2-1. The steel floor beams supporting the concrete slabs are governed by the abnormal extreme environmental load conditions, Equation 7 of Table 5.2-1. j O Generally, the concrete slabs have a higher stress margin than the supporting steel beams. 7.2-3 Rev. 1, 09/82

LGS DAR For the control structure shear walls, the stress levels are critical in the walls close to the base due to seismic loads. The stress margins for the shear walls at column lines 19.4 and 26.6, as shown in Figure E.1-38, were found most critical under the abnormal extreme environmental load condition including DBE and seismic torsional effects. The steel platforms at El. 313, 322, 340, and 350 ft were also assessed. The dynamic loads applied on the steel frames which support the platforms were found less significant than the normal loads. All the steel frames are governed by the normal load condition, Equation 2 of Table 5.3-1, with its associated allowable stresses. Those assessed steel members are shown in Figures E.1-18 through E.1-21. As demonstrated in Figure E.1-39, steel frames are generally understressed. 7.2.1.3 Suppression Chamber Columns The column vibration mode shapes are calculated using computer program BSAP. The mode shapes are shown in Appendix D, Figure D.2-1. The equivalent water mass is equal to the column volume. j g The stresses at the top and bottom of the suppression chamber columns were calculated and combined in accordance with the load combinations shown in Table 5.3-1. The maximum stresses in the column are governed by load combination Equation 7. The maximum stresses in the column (42-inch diameter pipe), top anchorage, and bottom anchorage are shown in Figure D.2-2. The lowest stress margin in the column structure is 10 percent. l 7.2.1.4 Downcomer Bracina l The bracing member forces and the corresponding design margins due to the governing load combinations are given in Figure D.2-11 for the critical bracing members. I 7.2.1.5 Liner Plate l 1

For the normal load condition, maximum negative pressure l (suction) on the pressure boundary portion of the liner plate occurs on the basemat and lower portions of the containment wall and RPV pedestal. The magnitude is 1.27 psi (negative). There is a large stress margin because the liner plates were designed for resisting a large suction, i.e., 5 psi (negative).

Rev. 1, 09/82 7.2-4

LGS DAR For the abnormal load condition, the liner plate does not experience net negative pressure, as can be observed from Figure 7.1-15. 7.2.1.6 Downcomers The downcomer vibration mode shapes are calculated for the modal analyses using computer program BSAP. The mode shapes are shown in Appendix D, Figures D.2-3 through D.2-5, for the three representative bracing system spring stiffnesses. The eqi'ivalent water mass included in the model is equal to the downcomec volume. The downcomers were assessed in accordance with ASME Section III, Division 1, subsection NB-3652, using load combinations in Table 5.5-1. Stresses and design margins are given in Appendix D, Figure D.2-6. Downcomer fatigue at three critical locations were also checked. Loads are combined by the absolute sum method. Figure D.2-7 x_ shows the fatigue usage factors at these critical locations, computed in accordance with ASME Section III, Division 1, subsection NB-3650 (1979 Summer Addenda). Downcomers are adequate for fatigue considerations. 7.2.1.7 Electrical Raceway System To be provided later. 7.2.1.8 HVAC Duct System To be provided later. 7.2.2 ACCELERATION RESPONSE SPECTRA 7.2.2.1 Containment Structure The method of analysis and load description for the acceleration O response spectrum generation are outlined in Section 7.1.1.1.1.6.1. From a review of the acceleration response spectra curves for the containment structure, the 7.2-5 Rev. 1, 09/82

i l l O i i 4 1 k I I I I 4 I f 4 ij l I t i l, I i l i 1 LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT , DELETED l FIGURE 7.21 R E V.1, 09/82 i

LGS DAR QUESTION 220.16 (DAR Section 5.3) In Table 5.3-1 load combinations 1, 2 and 3 do not contain the term Po, the operating pressure loads. Since the load combinations listed in this table also applies to reactor building structural steel, not only to containment internal structures, provide your justification for not including Po in these load combinations. Indicate if the containment will be inerted for hydrogen control. In load combination 6, the sign before LOCA is minus (-). Is this a typographical error?

RESPONSE

During power operation of the reactor, the containment atmosphere is inerted with nitrogen gas to preclude the possibility of a combustible mixture of hydrogen and oxygen accumulation in the primary containment. The method provided for inerting the containment is described in FSAR Section 9.4.5.1. The effect of the operating pressure load, where applicable, has

 \-     been considered in the assessment of structural steel within the containment and reactor enclosure. DAR Table 5.3-1 has been changed to include operating pressure loads in load combinations 1, 2, 3, and 4. In addition, the allowable stress in load combination 4 and the typographical errors in load combinations 5 and 6 have been corrected.

220.16-1 Rev. 1, 09/82

LGS DAR QUESTION 220.17 (DAR Section 7.2) In Section 7.2, Design Capacity Margins, it is stated that you are going to provide the pertinent information on margins of various structures at a later date. Indicate when you will be able to provide the necessary information.

RESPONSE

DAR Sections 7.2.1.1, 7.2.1.2, 7.2.1.3, and 7.2.1.6 have been added to provide pertinent information on the design capacity margins of the containment structure, reactor enclosure and control structure, suppression chamber columns, and downcomers, respectively. Information concerning the design capacity margins of the raceway system (Section 7.2.1.7) and HVAC duct system (Section 7.2.1.8) will be provided in the first quarter of 1983. O O i 220.17-1 Rev. 1, 09/82 I

LGS DAR QUESTION 480.62 Although FSAR Section 6.2.2.2 states that the RHR intake strainers are designed to withstand all hydrodynamic loads postulated to occur in the suppression pool, concerns arise due to the close proximity of the downcomer discharges to the intake strainers. Provide a list of all loads used in the design of the strainers and also provide additional information on your analyses that demonstrate the capability of the strainers to accommodate the hydrodynamic loads from downcomer discharges.

RESPONSE

The requested information will be provided in the fourth quarter of 1982. O O 480.62-1 Rev. 1, 09/82

LGS DAR QUESTION 480.67 Chapter 8 of the Design Assessment Report (DAR) that addresses the T-quencher verification test (proprietary) has not been submitted. We request that a copy of this chapter be submitted for our review.

RESPONSE

Volume 3 (proprietary) of the Design Assessment Report containing Chapter 8 was submitted to the NRC with Amendment 35 to the Limerick License Application by letter from E. J. Bradley to H. R. Denton, dated June 30, 1982. O O 480.67-1 Rev. 1, 09/82

i i i LGS DAR QUESTION 480.68 Provide the pool temperature analysis for the transient involving the actuation of one or more SRV's. For additional guidance, your attention is directed to NUREG-0873, " Pool Temperature Transients for BWR." l'

RESPONSE i

t i The requested information will be provided in the first quarter , of 1983. i s

O I

l , i 1 1 1 i l ! I l [ () ' I P l 480.68-1 Rev. 1, 09/82 l L

LGS DAR OUESTION 480.69 Table 1.3-2 of the DAR indicates that the quencher arm loads, the total quencher loads during SRV opening, and loads during irregular condensation are under evaluation. Provide these load specifications. RT.SPONSE The quencher load specifications are provided in DAR Volume 3 (Proprietary), Section 4.1. DAR Volume 3 was submitted to the NRC with Amendment 35 to the Limerick License Application by letter from E.J. Bradley to H.R. Denton, dated June 30, 1982.

                                                                                               . r O

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            ?                  480.69-1              Rev. 1, 09/82 n..                                                 _ _ _ _ _ _ _ - - . _ . _. __.

LGS DAR QUESTION 480.70 Concerns regarding the capability of the vacuum breaker to perform its function during the pool swell and chugging phases of LOCA have been raised. Provide the design changes, if any, that have been implemented to resolve this concern.

RESPONSE

A redesign and requalification program that considers the effects of the poolswell and chugging events has been initiated. The design changes will be implemented on Limerick during the second and third quarter of 1983 and will be provided in the DAR at that time. O l !O 480.70-1 Rev. 1, 09/82 l l

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0.25 j' O.00 g, 2 4 6 5 2 4 '> '3 39 g 2 4 6 8 3 g 100 GE00ENCY-LFS Accelera tion Spectra for REACTOR ENCL., CONTROL STRUCTURE Load Case: ASYN1ETRIC CHUGGING GE 700 SERIEF ENVELOPE (UIDENED - 15%) Node: 2 Di rec tion : HORI7. E-W Elev: 201'-0 Dampi ng: 0.005,0.01,0.02,0.03,0.05 LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O

                                                                                                   ~

REACTOR ENCLOSURE AND CONTROL STRUCTURE GLOBAL RESPONSE SPECTRA, E W HORIZONTAL, CHUG ASYMMETRIC FIGURE B.2 72 REV 1,09/82

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Node: 4 Direction: HORI7 E-W Elev: 230'-0 _ Danping: 0.005,0,01,0.02,0.03,0.05 LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O REACTOR ENCLOSURE AND CONTROL STRUCTURE GLOBAL RESPONSE SPECTR A, E-W HORIZONTAL, CHUG ASYMMETRIC FIGURE B.2-74 R EV 1,09/82 l- _ ,_. _.- _ _ . _ . _ _ _ _ - _ _ _ _ _ - _ - - - _ - - - . - - - -

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DESIGN ASSESSMENT REPORT

       \

i REACTOR ENCLOSURE AND CONTROL STRUCTURE GLOBAL RESPONSE SPECTRA, E-W HORIZONTAL, CHUG ASYMMETRIC FIGURE B.2 75 REV 1,09/82

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Accelera tion Spectra for REACTOR ENCL., CONTROL STRUCTURE Load Case: ASYMMETRIC CHUGGING GE 700 SERIES ENVELOPE (WIDENED - 15%) Node: 6 Direction: HORIZ E-W Elev: 269'-0 Damping: 0.005,0,01,0.02,0.03,0.05 LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O REACTOR ENCLOSURE AND CONTROL STRUCTURE GLOBAL RESPONSE SPECTR A, E W HORIZONTAL, CHUG ASYMMETRIC FIGURE B.2-76 R EV.1, 09/82

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                                                                                         RIOD SEC.

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Node: 11 Direction: HORIZ E-W Elev: 352'-0 Damping: 0.005,0.01,0.02,0.03,0.05 I l LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O REACTOR ENCLOSURE AND CONTROL STRUCTURE GLOBAL RESPONSE SPECTRA, E W HORIZONTAL, CHUG ASYMMETRIC FIGURE 8.2 81 REV 1,09/82

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

4 d t j i LGS DAR APPENDIX D l FIGURES l }  ! j NUMBER TITLE l  ! i  ; Containment Assessment Section Location

D.1-1 l 1

1 D.1-2 through D.1-5 Containment Stresses - Drywell Wall l j t D.1-6 through D.1-9 Containment Stresses - Wetwell Wall l 1 D.1-10 through D.1-11 Containment Stresses - Shield Wall l I D.1-12 through D.1-16 Containment Stresses - RPV Pedestal l D.1-17 through D.1-20 Containment Stresses - Diaphragm Slab l 2  : j D.1-21 through D.1-25 Containment Stresses - Base Slab l l 1 4 i i I' 1

)

1 i J j . i 4 4 D-il Rev. 1, 09/82 l

1. - ._ . _ . _ -. - _ _ _ _. - _ _ - . _. --. - _ _ - . - -_ -

1 l l l } LGS DAR i  !

  !                                                                  APPENDIX D                   !

j FIGURES l NUMBER TITLE l

]

j D.2-1 Suppression Chamber Columns Mode Shapes l D.2-2 Suppression Chamber Columns Design Margins f l i D.2-3 Downcomer Mode Shapes, K = 50 k/in l D.2-4 Downcomer Mode Shapes, K = 350 k/in l l D.2.5 Downcomer Mode Shapes, K = 15000 k/in l

 )

D.2-6 Downcomer Design Margins l D.2-7 Downcomer Fatigue Usage Factor l l D.2-8 Downcomer Fatigue Cycles l

i

() D.2-9 Downcomer Fatigue Histogram l j D.2-10 Downcomer Bracing System Mathematical Model l f D.2-11 Downcomer Bracing System Design Margins l 1 i i l i. i D-lii Rev. 1, 09/82

LGS DAR D.1 CONTAINMENT STRUCTURAL DESIGN ASSESSMENT Figure D.1-1 indicates the containment structural elements and cross sections where stresses are determined, and Figures D.1-2 through D.1-25 contain a tabulation of the predicted stresses and 1 allowable stresses for each loading combination considered. i Load combinations, taken from Table 5.2-1, are tabulated to cover all of the critical sections in the containment concrete

structures. Load combination Equation 2 for all sections and l Equations I and/or 3 and 6 for some sections are not executed because they do not represent the governing cases.

h i l 1 l I i 1 i i 1

O I

D.1-1 Rev. 1, 09/82

SECTIONS LOCATIONS 15 DRYWELL WALL 6, WETWELL WALL 12-1L SHIELD WALL PEDESTAL  % AXIS OF SYMMETRY [O 14-20 21-25 26-30 DI APHR AGM SLAB B ASL SL AB e i* s? s

• 1& . ,*

2 E L. 309'4" , {

                                                                      .                                I
                                                                       -     E L. 308' 3"               .
                                                                      ~
                                                                                                                     --               SHIE LD W ALL I

o . DRYWELL WALL 3 , E'. . 276'4" E L. 267 *-9" n

                                                                  -  wg
                                                                        ,-             .          u
                                                                          %.                      . s' 4                                                 -h E L. 262'-O"                                                E L. 259 *-0" BOTTOM                               l REBAR                       E L. 23S '-3 "

5 o, b;. 25' E L. 239'4" . T[OUTSIDEI[t$ih^" , b ', r- J w

                                                                    ~-                              .,.                           . . . . ..

OUTSIDE - ,ji R=8' R = 19' R =36 E L. 232'-0" R=25 DIAFHRAGM i OUTSIDE {REBAR INSIDE > l REBAR l . RPV 4 7

                               )4                      llNSIDE
                                                    , ,REBAR                                        '

l l g E L. 203'-6" 'q

                               %g g          (     ,

_-- E L. 200' 9" - (j'03 E L.198 " 6l h E L.185'-3 " R =27', R =42' WETWELL fWALL w , R = 8- R=16' R*33'

                                        ~
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T P EB R BASE SLAB h; (30 BOTTOM REBAR / 1 R ANSVE RSE (OUTSIDE) SHEAR TIESITYP) l l SECTION WHERE LIMERICK GENERATING STATION STRESSES ARE ASSESSED UNITS 1 AND 2 DESIGN ASSESSMENT REPORT 1 l CONTAINMENT ASSESSMENT SECTION LOCATION FIGURE D.1 1 REV 1,09/82 l l

O v DRYWELL WALL SECTIONS: 1, 2 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop Diaq. (2)(3) 1 - - - - - - - 3, 6 - - - - - - - 4, 4a 18.57 31.36 5.82 13.90 11.17 6.50 -0.233 4T, 4aT 7.14 6.83 13.4 19.75 16.83 10.99 -0.967 5, Sa, 7, 7a 25.66 30.04 9.95 13.45 20.82 4.60 -0.257 ST, Sa T , 7T, 11.36 -4.66 16.34 24.41 32.46 11.67 -1.542 fT V 7aT NOTES: (1) Allowable Rein forcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equa tions are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT v CONTAINMENT STRESSES DRYWELL WALL FIGURE D.1-2 R E V.1, 09/82

DRYWELL WALL SECTION: 3 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop Diaq. (2)(3) 1 -0.46 -0.07 -0.55 -0.04 -0.28 0.12 -0.432 3, 6 11.4 6.26 11.3 4.2 15.7 15.1 -0.200 4, 4a 9.97 43.0 14.8 34.5 17.4 'l.4

                                                                     .             -0.218 4T, 4aT        3.45    18.7   23.8    28.0    24.5         12.2           -0.926 5,  Sa, 7, 7a      24.1     40.2   21.3    17.4    36.9         20.1           -0.460 N                                     34.5            52.0 ST, Sa T , 7T,     14.9     15.9           27.4                 17.6           -1.38 7aT l

NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. l LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES DRYWELL WALL FIGURE D.1-3 R EV.1,09/82

i w O DRYWELL WALL SECTION: 4 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop Diaq. (2)(3)' 1 -0.55 -0.02 -0.69 0.06 -0.35 0.15 -0.097 3, 6 13.0 6.44 13.0 4.2 17.1 20.3 -0.230 4, 4a 8.49 41.7 21.8 20.9 23.2 10.8 -0.202 4T, 4aT 4.43 20.3 29.2 30.2 32.4 11.4 -0.822 5, Sa, 7, 7a 26.3 39.3 28.7 17.5 39.7 24.9 -0.522 ST, Sa T , 7T, 16.8 16.3 40.6 28.6 48.0 21.8 -1.431 [ 7aT

 \

NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "" fcc Compressive Stress (4) , Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. i 1 1 l l l LIMERICK GENERATING STATION j UNITS 1 AND 2 l DESIGN ASSESSMENT REPORT U l CONTAINMENT STRESSES DRYWELL WALL l FIGURE D.1-4 REV.1,09/82 l 1 1

O DRYWELL WALL SECTION: 5 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop Diaq. (2)(3) 1 -0.73 1.04 -0.80 1.0 -0.35 0.16 -0.106 3, 6 15.5 10.3 14.6 5.5 20.0 18.6 -0.294 4, 4a 31.2 33.7 21.6 6.9 14.6 39.8 -0.671 4T, 4aT 22.6 13.2 24.1 24.4 22.7 36.5 -0.671 5, Sa , 7, 7a 43.6 33.2 32.8 9.5 37.6 54.0 -0.931 ST, sat, 7T, 30.4 9.9 45.5 22.1 47.7 46.2 -1.71 7aT J NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI

                                                                                            )

(2) Allowable Cencrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress: " -" for Compressive Stress (4) Load Combination Equations are taken from 'I ;ble 5. 2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. l l l 1 I i l 1 LIMERICK GENERATING STATION < UNITS 1 AND 2 O DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES DRYWELL WALL l FIGURE D 1-5 REV.1,09/82

U WRTWELL WALL SECTION: 6 Max. MAXIMUM RERAR STRESSES,_KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination l Ties MSI Equations (4) Vert l Hoop Vert floop Diag. (2)(3) 1 -1.2 1.0 n.76 1.7 0.63 0.161 -0.99 3, 6 16.9 17.7 16.5 4.8 20.6 0.52 -0.361 4, 47 31.1 39.6 26.8 9.2 18.6 43.7 -0.582 4T, 4aT 26.0 48.7 26.7 28.1 28.2 35.0 -0.718 5, Sa, 7, 7a 50.1 43.1 36.0 12.9 45.7 44.8 -1.009 ST, Sa T , 7T, 24.9 48.8 . 53.9 l 26.7 47.6 27.5 -1.592 7a7 NOTES: (1) A1loweble Rein forcing Steel Stress = 54 KSI (2) A11cwahle Concreto Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equa tions rire taken from Table 5.2-1, 4T, 4a?, ST, sat, 7T, 7aT include thermal components. LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT A CONTAINMENT STRESSES l WETWELL WALL l FIGURE D.1-6 R E V.1, 09/82

(J ) WETWELL WALL SECTIONS: 7, 8, 9 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop' Diaq. (2)(3) 1 -1.36 9.7 -1.4 4.8 2.09 0.89 -0.210 3, 6 25.5 20.3 23.3 6.8 28.4 5.3 -0.427 4, 4a 14.8 38.4 26.8 25.5 26.2 13.6 -0.616 4T, 4aT 12.8 46.2 34.6 33.0 33.8 14.0 -1.31 5, Sa, 7, 7a 37.7 37.0 47.9 21.8 48.6 15.2 -0.819 ST, sat, 7T, 33.2 41.0 50.0 46.3 53.9 17.3 -2.12 m 7aT i l l V NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, 5T, sat, 7T, 7aT include thermal components. LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT v CONTAINMENT STRESSES WETWELL WALL FIGURE D.17 R EV.1, 09/82

i WETWELL WALL SECTION: 10 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination l Ties KSI Equations (4) Vert l Hoop Vert Hoop Diag. (2)(3) 1 -1.68 15.8 -1.5 7.4 3.35 1.1 -0.254 3, 6 27.5 30.7 25.5 7.58 31.1 0.70 -0.503 4, 4a 16.6 42.4 29. 1 !35.3 31.4 5.3 -0.744 4T, 4aT 12.2 35.6 38.0 39.7 37.9 8.13 -1.50 5, Sa, 7, 7a 37.5 40.1 43.6 27.5 50 6.7 -1.13 ST, Sa T , 7T, 29.4 46.7 c'.R 35.6 52.4 7.4 -2.25 7aT NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "- " for Compressive Stress , (4) Load Combination Equa tions are taken from Table 5.2-1,

\

4T, 4aT, ST, sat, 7T, 7aT include thermal components. , j LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT J CONTAINMENT STRESSES WETWELL WALL FIGURE D.1-8 R E V.1, 09/82

O WETWELL WALL SECTION: 11 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop Diaq. (2)(3) 1 -1.57 4.95 -1.5 2.96 1.16 2.81 -0.233 3, 6 29.8 21.2 27.1 8.48 34.3 15.3 -0.527 4, 4a 38.1 35.5 33.2 6.48' 20.3 42.9 -0.702 4T, 4aT 36.1 18.5 38.2 11.2 25.1 44.5 -0.990 5, Sc , 7, 7a 53.9 32.9 46.0 9.0 45.0 45.0 -1.04 ST, sat, 7T, 47.2 40.6 51.2 17.0 47.4 45.4 -1.69 p

 \j 7aT l

NOTES: (1) Allowable Rein forcinq Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress , (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. i LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O CONTAINMENT STRESSES WETWELL WALL FIGURE D.1-9 REV.1,09/82 i

i Y t SHIELD WALL SECTION: 12

                                                                                                  . Max.

l MAXIMUM STEEL STRESSES, KSI (1)(3) Concrete INNER PLATE OUTER PLATE Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop (2)(3) 1 0.39 3.6 -0.11 1.2 1.0 -0.071

<                              3,   6          7.1      8.7    2.1     2.9        1.4           -0.293 4,   4a         2.2      9.2   -0.64    3.7        7.5           -0.265 4T, 4aT         2.0      8.8    0.R1    3.3        7.5           -0.265 5,  Sa , 7, 7a       8.5     12.8     2.7    5.1        9.5           -0.407 ST, Sa T , 7T,       R.3     12.4     2.5    4.7        9.5           -0.407 7aT J

, O NOTES: (1) Allowable Reinforcing Steel Stress = 30. 0 KSI i (2) Allowable Concroto nmpressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. ! LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT b G CONTAINMENT STRESSES SHIELD WALL FIGURE D.1-10 R E V.1, 09/82

O V SHIELD WALL SECTION: 13 Max. MAXIMUM STEEL STRESSES, KSI (1)(3) Concrete INNER PLATE OUTER PLATE Transverse Stress, Load Combination Ties KSI Equations (4) Vert Hoop Vert Hoop (2)(3) l 1 -0.28 0.08' -0.57 -0.10 0.128 -0.077 3, 6 9.7 3.5 2.9 1.1 0.63 -0.404 4, 4a -0.65 0.29 -0.94 -0.15 0.26 -0.12R 4T, 4aT -1.45 -1.73 -0.53 1.03 0.26 -0.128 5, Sa, 7, 7a 10.7 3.6 2.9 1.1 2.4 -0.444 ST, sat, 7T, 9.9 1.9 3.3 2.1 2.4 -0.444 7aT l 1 ' p NOTES: (1) Allowable Reinforcing Steel Stress =30.6 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT Qv' CONTAINMENT STRESSES SHIELD WALL FIGU RE D.1-11 R EV.1, 09/82

s RPV PEDESTAL SECTION: 14 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 -1.0 1.0 -1.2 1.2 0.34 -0.157 3, 6 17.2 13.4 29.2 17.4 3.9 -0.352 4, 4a -1.3 2.4 -1.7 2.0 0.31 -0.230 4T, 4aT 7.98 7.0 -2.31 4.97 0.31 -0.230 5, Sa, 7, 7a 17 . 2 14.7 25.6 17.2 3.3 -0.432 ST, Sa T , 7T, 25.7 19.3 25.0 20.2 3.3 -0.432 I NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES RPV PEDESTAL FIGURE D.1 12 R E V.1, 09/82

O RPV PEDESTAL SECTION: 15 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) , Radial Honn Dadia l Hoop (2)(3) 1 -1.5 0.94 -2.2 0.32 0.35 -0.290 3, 6 43.9 27.2 52.5 33.6 4.7 u.649 4, 4a 4.5 32.2 6.1 47.0 21.1 -0.474 4T, 4aT 14.5 -4.R 4.9 -5.9 48.3 -0.910 5, Sa, 7, 7a 52.9 50.1 52.9 51.8 39.4 -0.856 ST, sat, 7T, 49.9 8.2 51.9 -4.0 27.4 -1.017 {*s} s 7aT l t NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 l DESIGN ASSESSMENT REPORT O CONTAINMENT STRESSES RPV PEDESTAL FIGURE D.1-13 R E V.1, 09/82

O V RPV PEDESTAL SECTION: 16 Max. MAXIMUM RERAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Padial Hoop Radial Hoop- (2)(3) 1 - 1. 5 1.1 -2.0 ' .1 0.34 -0.266 3, 6 30.3 13.0 39.4 29.1 0.86 -0.526 4, 4a 6.9 12.6 -4.8 30.4 7.9 -0.678 4 T, 4aT 13.3 13.3 -5.7 28.0 15.0 -1.051 5, Sa, 7, 7a 45.0 26.1 42.7 36.8 19.6 -0.931 ST, sat, 7T, 37.3 16.0 22.7 15.5 27.9 -1.249 O V 7aT NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES RPV PEDESTAL FIGURE D.1 14 R E V.1, 09/82

[^\ s RPV PEDESTAL SECTIONS: 17, 18 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 -2.1 5.0 -2.7 12.9 9.0 -0.382 3, 6 9.9 8.5 10.5 17.0 12.9 -0.690 4, 4a -4.1 11.9 -4.8 28.3 17.0 -0.681 4T, 4aT 4.13 13.8 -4.3 28.9 26.8 -0.635 3, Sa, 7, 7a 18.6 15.7 20.5 29.8 22.5 -1.017 5'I , Sa T , 7T, 23.0 22.1 22.0 32.6 38.9 -0.968 7aT v - NOTES: (1) Allowable Rein forcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT v CONTAINMENT STRESSES RPV PEDESTAL FIGURE D.1 15 R E V.1, 09/82

l RPV PEDESTAL SECTIONS: 19, 20 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load tombination Ties KSI Equations (4) Radial Hoop Radial Hoon (2)(3) l 1 -2.01 -0.176 -2.95l 0.2?l 0.59 -0.424 l 3, 6 17.9 5.26 11.8 5.4 4.7 -0.483 4, 4a 4.86 3.69 -5.2 7.1 5.68 -0.744 4T, 4aT 5.2 -6.1 -5.39 -4.8 5.68 -0.744

5, Sa, 7, 7a 25.9 7.2 32.5 12.8 15.9 -0.851 ST, Sa T , 7T, 26.2 -5.8 32.3 8.2 15.9 -0.851

    \                         :

NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI A (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "" for Compressive Stress (4) Load Combination Equa tions are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT V CONTAINMENT STRESSES RPV PEDESTAL FIGURE D.1 16 RE V.1, 09/82

o l i DIAPHRAGM SLAB SECTION: 21 Max. . MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radia1 Hoop Radial Hoop (2)(3) 1 - - - - - - 3, 6 8.5 8.6 6.9 7.7 1.01 -0.073 4, 4a 38.8 30.2 28.9 22.7 8.8 -0.374 4T, 4aT 32.8 21.6 35.9 27.6 9.5 -1.R2 ! 5, Sa, 7, 7a 35.6 30.1 29.3 23.3 8.8 -0.365 ST, sat, 7T, 31.7 21,5 34.6 28.0 8.9 -1.83 v) 7aT NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "" for Compressive Stress (4) Load Combination Equa tions are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components.

.i l

LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES DI APHR AGM SLAB FIGURE D.117 R E V.1, 09/82

DIAPHRAGM SLAR SECTIONS: 22, 23 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoon l Dadial Hoop (2)(3) 1 - - - - - - 3, 6 7.9 9.5 10.2 13.0 4.46 -0.370 4, 4a 14.1 21.5 18.4 24.2 16.1 -0.383 4T, 4aT -11.1 12.3 26.2 29.8 7.0 -1.367 5, Sa, 7, 7a 16.4 23.1 23.7 27.9 18.0 -0.623 ST, sat, 7T, -13.1 16.0 25.5 35.9 7.2 -1.727 ['s) 7aT l _ _ _ _ l NOTES: (1) Allowable Rein forcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KS1 (3) "+" for Tensile Stress; "" for Compressive Stress (4) Load Combination Fouations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT lh 1 V CONTAINMENT STRESSES DIAPHRAGM SLAB FIGURE D.1-18 R E V.1, 09/82

O DIAPIIRAGM SLAB SECTION: 24 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 - - - - - - 3, 6 10.2 9.6 9.1 8.0 3.0 -0.272 4, 4a 22.8 22.6 30.5 21.1 5.9 -0.842 4T, 4aT -8.61 -8.29 33.2 29.4 4.2 -1.59 5, Sa, 7, 7a 27.9 25.4 33.4 24.4 6.2 -0.931 ST, Sa T , 7T, -10.3 12.3 35.5 30.8 4.9 -1.738 n 7aT U , NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensilo ctress; " -" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES DIAPHRAGM SLAB FIGURE D.119 R E V.1, 09/82

O DIAPHRAGM SLAB SECTION: 25 Max. MAXIMUM RRBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 - - - - - - 3, 6 13.0 15.5 12.1 14.5 0.66 .157 4, 4a 26.7 28.0 23.5 30.6 9.9 .336 4T, 4aT 12.9 24.0 26.1 35.4 6.5 -2.04 5, Sa, 7, 7a 33.6 38.5 28.6 35.9 10.4 .423 ST, Sa T , 7T, 19.3 31.8 41.1 42.4 9.5 -2.40 7aT NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allo"r.ble Concrete Compressive Stress = 3.4 KSI (3) "+" tor Tensile Stress; "" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, S a 'T' , 7T, ~7aT include thermal components. LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES I DI APHR AGM SLAB FIGURE D.120 R E V.1, 09/82 1

%.J RASE SLAB SECTION: 26 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER l OUTER Transverse Stabs, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 _ _ _ _ _ _ 3, 6 1.7 16.6 5.93 6.22 5.29 -0.318 4, 4a 2.72 1.61 7.10 3.29 0.43 -0.213 4T, 4aT -5.21 -5.63 15.2 34.9 3.4 -1.21 5, Sa, 7, 7a 10.9 20.7 10.4 9.51 4.03 1

                                                                                    -0.443 ST, sat, 7T,           -6.36    -4.95:   18.4     17.3     ,      3.12         -1.34 O) s o          _

7aT l -. c__-__ NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equa-ions are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPOR f v CONTAINMENT STRESSES BASE SLAB FIGURE D.121 REV.1,09/82

D

  \~-]

BASE SLAB SECTION: 27 Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 - - - - - - 3, 6 25.4 26.8 15.5 6.43 24.7 -0.479 4, 4a 10.3 -0.43' 11.1 0.65 23.9 -0.309 4T, 4a7 22.4 -7.3 29.8 13.3 33.1 -1.70 5, Sa, 7, 7a 39.8 34.4 29. 3 13.9 41.0 -0.540 i ST, sat, 7T, 30.0 20.2 29.0 17.1 39.8 -1.79 h 7aT I NOTES: (1) Allowable Reinforcing Steel Stress = 54 VSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. l LIMERICK GENERATING STATION UNITS 1 AND 2 I DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES BASE SLAB i FIGURE D.122 R EV.1,09/82

RASE SLAB SECTION: 28 l Max. MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 - - - - - - 3, 6 34.0 17.R 11.1 10.6 17.5 -0.910 4, 4a 21.7 9.15 16.9 8.7 12.5 -0.304 4T, 4aT -8.25' -8.07 17.7 13.4 6.7 -1.59 5, Sa, 7, 7a 42.1 21.2 18 . 2 16.1 25.2 -0.985 ST, sat, 7T, 25.4 -8.4 l 23.7 19.8 18.7 -1.72 O \ 7aT NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Comprassive Stress = 3.4 KSI (3) "A" for Tensile Stress; " -" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENER ATING STATION UNITS 1 AND 2 f DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES BASE SLAB g FIGURE D.123 R E V.1, 09/82

O BASE SLAB SECTION: 29 Max. l MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete-INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop' (2)(3) 1 - - - - - - 3, 6 12.7 15.6 9.01 8.52 11.1 -0.524 4, 4a 11.8 9.51 17 .3 7.95 11.0 -0.243 4T, 4aT 9.32 -6.40 20.6 12.6 12.0 -1.23 5, Sa, 7, 7a 17.7 18.9 19.0 13.3 18.5 -0.508 ST, sat, 7T, 14.4 -6.22 21.5 16.1 18.9 -1.18 d 7aT l NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI (3) "+" for Tensile Stress; "" for Compressive Stress (4) Load Combination Equations are taken from Table 5.2-1, 4T, 4aT, ST, sat, 7T, 7aT include thermal components. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT CONTAINMENT STRESSES BASE SLAB FIGU RE D.1-24 R EV.1,09/82

(v\ BASE SLAR SECTION: 30 Max. l MAXIMUM REBAR STRESSES, KSI (1)(3) Concrete INNER OUTER Transverse Stress, Load Combination Ties KSI Equations (4) Radial Hoop Radial Hoop (2)(3) 1 - - - - - - 3, 6 9.32 16.9 9.14 8.54 7.0 -0.414 4, 4a 29.9 33.5 34.1 10.2 25.6 -0.430 4T, 4aT 29.2 5.5 38.5 12.9 27.4 -0.902 5, Sa, 7, 7a 23.9 36.5 32.9 16.6 28.4 -0.688 gg ST, Sa%, 7T, i  : 7aT 22.5 -6.36 36.5 19.6 28.9 -0.915

l. l \

NOTES: (1) Allowable Reinforcing Steel Stress = 54 KSI (2) Allowable Concrete Compressive Stress = 3.4 KSI l (3) "+" for Tensile Stress; "" for Compressive Stress l l (4) Load Combination Equa tions are taken from Table 5. 2-1, ! 4T, 4aT, ST, sat, 7T, 7aT include Thermal Components. l LIMERICK GENERATING STATION l UNITS 1 AND 2 DESIGN ASSESSMENT REPORT i l CONTAINMENT STRESSE.S BASE SLAB FIGURE D.125 R E V.1, 09/82

LGS DAR {~} v D.2 SUBMERGED STRUCTURE DESIGN ASSESSMENT The submerged structures in the suppression chamber include the diaphgragm slab support columns, the downcomer bracing system, and the downcomers. The bracing system and the columns are assessed in accordance with Table 5.3-1. In the column assessment, the dynamic loads are combined by the SRSS method and then combined with the static loads using the absolute sum procedure. In the assessment of the downcomer bracing system, all loads are combined using the absolute sum method. For both the downcomer bracing system and the columns, Equation 7 of Table 5.3-1 is the most critical combination. The natural vibration frequencies and shapes of the suppression chamber columns are presented in Figure D.2-1, and the assessment results are summarized in Figure D.2-2. Bolt stresses are not shown in the bottom anchorage because the design is more critical at the connecting flange, which yields a design margin of 10 percent. r The downcomer bracing system mathematical model is shown in Figure D.2-10, and the design margins for the most critical member in each quadrant are summarized in Figure D.2-11. ( O D.2-1 Rev. 1, 09/82

O l l l 1 l I I 1 I i l l I 1 I l\ i i I I l I f o \

                                                      /^,

l i s, i 1 I I I I J

;           MODE 1                   MODE 2              MODE 3 f=20 HZ                  f=53 HZ             f=93 HZ l                           (WITH WATER MASS) 3 i

LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O SUPPRESSION CHAMBER COLUMNS MODE SHAPES 1 FIGURE D.21 R EV.1, 09/82

(N d fN O'd SUPPRESSION CilAMBER COLUMNS MAXIMUM ALLOWABLE MAXIMUM ALLOWABLE COMBINED AXIAL STRESS AXIAL STRESS PLEXURAL PLEXURAL STRESS STRESS COLUMN (KSI) (KSI) STRESS (KSI) STRESS (KSI)' RATIO MARGIN % 42" dia pipe (shell element) 11.7 27.3 8.7 28.0 0.74 26 Top Anchorage 22.6 29.9 - - 0.76 24 Bottom Anchorage - - - - - 10 NOTE: These stress margins are based on load combination 7 of Table 5.3.1 which is the critical load combination. m o O C C = m r o P m = = Y M ~ 3 09 $Q EO DE m u_o O@ Gd* z>E E;; m z zH m D@ oz em ="o m m zoO

• o 4

< r m d L c H o g z ~

O I I I I I I I I I I I I r I I I i I l  ; l j l I l ' I I i i i I l J l I I I I I I I I I I l l 6--%' - I I

                                               ,   i n                   l            l V   k = so. k/t,. . I
                                     !\        !                        i                               .

i Mopai Mops 2 peos 3 Maps 4 G.5 H2 17.0 K 2 4 2. 4 H Z 9o. 8 H T, l l l l LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT DOWNCOMER k MODE SHAPES,K=50 /in FIGURE D.2-3 RE V 1, 09/82

O I I i l l l 1 l l i i l i l I I l l 1 l l l I I l l l i l l l l I 1 I O l

                                                          +----;

[ i l i I l l l h 35e."A.. l l Mope i Mops 2 Mope 3 Moos 4 II.9 H2 22.9 H Z 44 1 HL 9o.9 Hz LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT DOWNCOMER k MODE SHAPES, K=350 / in FIGURE D.2-4 R E V.1, 09/82

I I I I I I I I I l l i I I l I l j I i i l l I I l 1 1 I I I I I I I I l I I l O  ! l V g-- w _ l I LI b isoeo. %. l , I / MaPG i Mepe T. Moos %

13. 5 HE 36.9 H2 99,9 Hs l

l LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT DOWNCOMER MODE SHAPES, K=15000 "/ in FIGURE D.2-5 REV.1,09/82

O DOWNCOMER - STRESS

SUMMARY

AND DESIGN MARGINS LOAD i ALLOWABLE l l iiiiSIGN COMBINATION CONDITION STRESS l STRESS MARGIN (KSI) l (KSI) (%) Equation 1 Upset 28.4 17.5 38.4 Equa tion 2 Emergency 42.5 19.9 53.2 Equation 3 Eme rge ncy 42.5 37.4 12.0 Equation 4 Faulted 56.7 20.0 64.7 Equation 5 Faulted 56.7 37.4 34.0 Equation 6 Faulted 56.7 37.5 33.9 l Equation 7 Faulted 56.7 24.5 56.8 4 O ' NOTE: Equation numbers are based on Table 5.5-1. LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT DOWNCOMER DESIGN MARGIN FIGURE D.2 6 R E V.1. 09/82

w to e O) S rl eP uai smsmU nl G E N A srnaH S O B eeaeC S I D rhrt T PTTS+- + I D * *

• N O

C w A to D B e nl F S rl eF T uai G 1 2 2 2 4 L r smsmU V V E 1 3 8 U o srnaH R R S 6 4 9 A eeaeC S S S 7 0 6 P A rhrt

         /B    PTTS+-           + +      +  0      0        0 Y  I C    * *

• N E

G w R to E e nl S M rl eF R E uai G 1 2 E A smsmU V V M B srnaH R R O S eeaeC S S C rhrt N PTTS+- + + W O **

• D 1 2 9 3 F V 5 2 5 O R 5 4 9 S 7 0 6 Y

R +- 0 0 0 A N M O M I 1 2 8 9 2 U T V V 3 2 4 S I R R 6 4 2 D S S 6 0 6 R N O T O C

               +- +                        0      0        0 C

3J A T 2 G 5 1 9 7 P E V V U 7 2 6 S R R H 4 4 4 E P S S C 5 0 5 G U A / +- 1 +- 0 0 0 S L U A _ M 1 2 0 4 R E V V 2 1 O B R R 0 0 N O S S 0 0 0

              +- + +-                      0               0

~ d l e W t r n g e" e n 0" k0 m 0" i - a- h - R' e c S 3 r 1' a '1 D m2 B3 t2 A r2 2 t2 O L

• fo.

t v m. u uv A ev ae ce pe . ll al il PE VE PE - t t t A A A r 5m3 k o zm:>dzo D " y3oZ czaM*>zoN Oy3Z pgmmME9g mmT@a _L U gZO0Em3 n >dOCm cepcm n mpoHom ~

                                                 -@:mp 0N"                    2m ". ok5

I i STRESS CYCLES FOR FATIGUE EVALUATION OF DOWNCOMERS 1 8 ' ~ LOAD TYPE No. OF CYCLES NSRV1 14100

NSRV2 7700

,                                     NOBE                                 50 NCl1UG                          3000 NSSE                                 10 i

l O NOBE = Cycles associated with OBE 4 N SRV1 = Cycles associated with SRV (Submerged Structure Load) N SRV2 = Cycles associated with SRV2 (Inertia) { NCilUG = Cycles associated with Chugging NSSE = Cycles associated with SSE I 1 j i LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT DOWNCOMER FATIGUE CYCLES

  '                                                             FIGU RE D.2-8                     REV.1,09/82

O O O 1 FATIGUE LOAD HISTOGRAM FOR DOWNCOMERS l NORMAL / UPSET CONDITION I EMERGENCY L FAULTED SBA IBA or SBA DBA l J

                    + ORE )     + SRV1      -
                                             + SRV 1' -+ SRV1

• Pressure Pressure Pressure l-
• Thermal
• Thermal
• Thermal

                   - + SRV1     + SRV2
                                -           + SRV2 Transient       Transient         Transient
                                                                              *

• Stean Flow
• Steam Flow Steam Flow

                     + SRV2      + CHUG                          + CHUG          + CHUG            + CHUG 1
                                                                 + SPV1          + SRV1            f SSE
                                                                 + SRv2          + SRv2 3 SSE 1        3          5        7           9               9                  9
   =

A; A A A, A, A , A a 2 4 6 8 10 10 10 h y

               ~

o E 5 50 3000 4650 6400 Load set pair 9-10 is for one of the

           "                                                    three above events which produce the i 9 A    3   E l cycles'       cyclesl     cycles l cycles d    z Q   l                                              largest combined stress.

c0 pc o co mE kz4z The cycles associated with oscillatory E$ ESE loads combined with SSE are assumed

      $@   !$D                                                  conservatively to be 10.

Om 0z a x $na l  ! 3 m m H m > m E o 6

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e LIMERICK GENERATING STATION M UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O DOWNCOMER BRACING MATHEMATICAL MODEL FIGURE D.2-10 R EV 1,09/82

O DOWNCOMER BRACING SYSTEM - STRESS

SUMMARY

BRACING MEMBER DESIGN MARGINS FOR CRITICAL MEMBERS AND GOVERNING LOAD COMBINATION OUADRANT(2) MEMBER (2) EOUATION(1) MARGIN - % 1 58 7 5% 2 75 7 6% 3 126 7 5% 4 217 7 4% Link between 221 7 3% Ouadrants i .i NOTES: (1) Equation number is based on Table 5. 3-1 (2) Piqure 0. 2-10 gives location reference i I I i i i .i i i LIMERICK GENER ATING STATION j UNITS 1 AND 2 DESIGN ASSESSMENT REPORT 'O DOWNCOMER BRACING SYSTEM DESIGN MARGIN FIGURE D.2-11 R E V.1. 09/82

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i LGS DAR 1 l i APPENDIX E i REACTOR ENCLOSURE AND CONTROL STRUCTURE l ! STRUCTURAL DESIGN ASSESSMENT 1 t l TABLE OF CONTENTS E.1 Reactor Enclosure and Control Structure Structural Design Assessment i f i i ) 1 i i  ? f l I . t ) 1 4 1 1 1 l 1 i i i !O f l l 2  ; I j E-i Rev. 1, 09/82 ,

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l l l l LGS DAR APPENDIX E FIGURES NUMBER TITLE E.1-1 Reactor Enclosure and Control Structure Floor Plan (El. 177') E.1-2 thru E.1-8 Reactor Enclosure Steel Framing Plan (El. 201, 217, 253, 283, 313, 331, and 352 ft) E.1-9 thru E.1-17 Control Structure Steel Framing Plan (El. 200, 217, 239, 254, 269, 289, 304, 332, and 350 ft) E.1-18 thru E.1-21 Control Structure Steel Platforms (El. 313, 322, 340, and 350 ft) E.1-22 Axial Forces - OBE + SRV l E.1-23 Axial Forces - DBE + SRV + LOCA l E.1-24 N-S Shear Forces - OBE + SRV l E.1-25 N-S Shear Forces - DBE + LOCA + SRV l E.1-26 N-S Overturning Moments - OBE + SRV l E.1-27 N-S Overturning Moments - DBE + LOCA + SRV l E.1-28 E-W Shear Forces - OBE + SRV l E.1-29 E-W Shear Forces - DBE + LOCA + SRV l E.1-30 E-W Overturning Moments - OBE + SRV l E.1-31 E-W Overturning Moments - DBE + LOCA + SRV l E.1-32 thru E.1-39 Reactor Enclosure and Control Structure Stresr, Margins O E-ii Rev. 1, 09/82

LGS DAR APPENDIX E I E.1 REACTOR ENCLOSURE AND CONTROL STRUCTURE STRUCTURAL DESIGN l ASSESSMENT Figure E.1-1 presents the reactor enclosure and control structure general floor plan at El. 177 ft to aid in the location of wall marks. Figures E.1-2 through E.1-21 identify and locate the selected critical structural elements where stresses are assessed in the reactor enclosure and control structure. Figures E.1-22 through E.1-31 present diagrams of combined vertical (axial) forces, N-S and E-W shear forces, and N-S and E-W overturning moments, based on the dynamic portion of the load combinations specified in Tables 5.2-1 and 5.3-1. Figures E.1-32 through E.1-39 contain tabulations of predicted stresses, stress allowables, and/or stress margins for the reactor enclosure and control structure floor slabs, floor support steel, and shear walls. O E.1-1 Rev. 1, 09/82

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LIMERICK GENER ATING STATION l l UNITS 1 AND 2 DESIGN ASSESSMENT REPORT REACTOR ENCLOSURE & CONTROL STRUCTURE VERTICAL AXIAL FORCES (X103 KIPS) OBE+SRV (2% DAMPING) l FIGURE E.122 REV 1,09/82

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UMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT REACTOR ENCLOSURE & CONTROL STRUCTURE N-S SHEAR FORCES (X103 KIPS) DBE+LOCA+SRV (5% DAMPING) FIGURE E.125 REV 1,09/82

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LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT REACTOR ENCLOSURE & CONTROL STRUCTURE N-S OVERTURNING 6 MOMENTS (X10 K-FT) OBE+SRV (2% D AMPING) FIGURE E.126 REV 1,09/82

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LIMERICK GdNERATING STATION l UNITS 1 AND 2 l DESIGN ASSESSMENT REPORT l [ REACTOR ENCLOSURE & CONTROL l STRUCTURE N-S OVERTURNING l MOMENTS (X106 K-FT) l DBE+LOCA+SRV (5% DAMPlNG) FIGURE E.127 REV 1,09/82

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l LIMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT REACTOR ENCLOSURE & CONTROL STRUCTURE E-W SHEAR FORCES (X103 KIPS) OBE+SRV (2% DAMPlNG)

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UMERICK GENERATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT / ( REACTOR ENCLOSURE & CONTROL STRUCTURE E-W SHEAR FORCES (X103KIPS) DBE+LOCA+SRV (5% DAMPING) FIGURE E.129 REV 1,09/82

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• UNITS 1 AND 2 DESIGN ASSESSMENT REPORT REACTOR ENCLOSURE &

CONTROL STRUCTURE E-W OVERTURNING MOMENTS (X10* K-FT) j DBE+LOCA+SRV (5% DAMPING) FIGURE E.131 R E V.1, 09/82

Iv) 1 REACTOR ENCLOSURE FLOOR SLABS JELEMENT[ ELEVATION [ SLAR GOVERNING RERARG I STRESS NUMBER (FT) THICKNESS EQUATION (1) STRESS MARGIN (PT) (KSI) (%) 1  ; s1 1.5 1 13.13 75.7 2 2'1 2.5 1 30.55 43.4 3 217 1.5 7a 30.90 42.R 4 217 2.0 7a 27.70 48.7 5 253 1.25 7a 51.26 5.1 6 253 2.0 1 20.40 62.2 7 283 1.25 7a 42.74 20.9 8 283 2.75 1 28.13 47.9 9 313 1.75 1 30.52 43.5 10 313 2.0 7a 23.83 55.9 11 313 3.0 7a 28.16 47.9

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, 13 333 1.67 1 15.47 71.4 14 352 2.0 7a 36.35 32.7 15 352 3.25 7a 11.1R 79.3 l NOTES: (1) Taken from Table 5. 2-1 as follows: Load Combination EON 1 = 1. 4 D + 1. 7 L + 1. 5 SRV Load Combination EON 7a = 1.0D + 1.0L + 1.0 ESS

                                                      + 1.0 SRV + 1.0 LOCA (2) Allowable Reinforcing Steel Stress = 54 KSI LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O                                                  REACTOR ENCLOSURE MARGINS FLOOR SLABS FIGURE E.1-32               R EV.1,09/82

C'T 'Q REACTOR ENCLOSURE FLOOR STEEL BEAM (1) ELEMENT ! ELEVATION GOVE RN ING BENDINGiSTRE5'S NUMBER (PT) STEEL SIZE EOUATION(2) STRESS ' MARGIN (3) (KSI)  % 16 201 W27 x 145 1 23.00 4.2 17 201 W24 x 68 1 20.00 16.7 18 217 W33 x 141 1 21.90 8.6 19 217 W33 x 130 7 27.40 15.5 20 253 W24 x 76 1 22.66 5.6 21 253 W27 x 84 1 20.92 12.A 22 283 72" Girder 1 24.00 0. 23 2R3 W33 x 152 1 19. 27 19.7 24 313 56" Girder 7 30.28 6.5 25 313 W36 x 300 7 29.44 9.1 26 331 W36 x 182 7 23.58 27.2 27 331 W21 x 73 7 20.31 37.3 /; 28 352 W36 x 300 7 18.54 42.7 (_,) 29 352 W24 x 68 7 16.94 47.7 l NOTES: (1) All beams are A-36 steel. (2) Taken from Table 5.3-1 as follows: Load Combination EON 1 = D + L + SRV Load Combination FON 7 = D + L + E + SRV + LOCA 1 (3) Allowable Rending Stresses for Governing Equations 1 and 7 are 24.0 KSI and 32.4 KSI, respectively. LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT V' REACTOR ENCLOSURE MARGINS FLOOR STEEL BEAM FIGURE E.1-33 R EV.1, 09/82

r I i REACTOR ENCLOSURE SUPPORTING COLUMNS ELEVATION INTERACTION STRESS RANGE LOCATION (1) MATERIALS (2) EOUATION MARGIN % 177'-201' 29 & E Steel 0.77 23 177'-201' 30.5 & E Rein fo rced Concrete - 8 201'-217' 29 & E ' Steel 0.78 22 201'-217' 30.5 & E Reinforced Concrete - 1 217'-253' 30.5 & E Steel 1.02 0 253'-283' 30.5 & E Steel 0.88 12 2R 3 '-313 ' 30.5 & E Steel 0.78 22 313'-333' 27.5 & E Steel 0.97 3 313'-333' 30.5 & E Steel 0.91 9 333'-352' 29 & E Steel 0.65 35 O d NOTES: (1) Figure E.1-1 gives location reference (2) For Steel Supports, Load Combination EON (7) of Table 5.3-1 is used: D + L + E + LOCA + SRV + p. For Reinforced Concrete Supports, Load Combination EON (7) of Table 5.2-1 is used: D+L+Eo + LOC A + SRV + P B LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O REACTOR ENCLOSURE MARGINS SUPPORTING COLUMNS FIGURE E.1-34 R EV.1, 09/82

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REACTOR ENCLOSURE SHEAR WALLS

-l COMBINED AXIAL                           SHEAR WALL                          GOVERNING   & BENDING STRESS STRESS ELEVATION            WALL MARKII) EOUATION(2) MARGIN (%) (3)                            MARGIN

(PT) (%) (4) 4 177 Line 14.1 7a 67 1 177 Line 31.9 7a 67 1 177 D 7a 48 12 177 Line 23 7a 24 8 177 Line 21.5 7a 29 9 NOTES: (1) Figure E.1-1 qives location reference (2) Taken from Table 5. 2-1 as follows :

Load Combination EON 7a = D + L + Ess + SRV + LOCA (3) Allowable Reinforcing Steel Stress = 54 XSI i (4) Allowable Reinforcing Steel Stress = 51 KSI i l l LIMERICK GENER ATING STATION UNITS 1 AND 2 l DESIGN ASSESSMENT REPORT iO l REACTOR ENCLOSURE MARGINS SHEAR WALLS FIGURE E.135 REV.1,09/82 i

l O 4 I i CONTROL STRUCTURE FLOOR SLABS i SLAR REBAR STRESS ELEMENT ELEVATION TilICKNESS GOVERNING STRESS (2) MARGIN ] NUMBER (PT) (PT) EOUATION(1) KSI (%) 30 200 1.5 1 14.47 73.2 31 200 6.0 1 37.64 30.3 32 217 1.25 1 14.15 73.8 33 237 1.0 1 31.10 42.4 34 237 1.0 1 30.89 42.8 35 254 1.0 1 27.86 48.4 36 269 1.5 1 12.15 77.5 37 289 1.5 1 10.26 81.0 0 38 39 40 304 332 332 1.0 1.5 2.0 1 1 1 22.95 16.92 41.4 57.5 68.7 23.3 41 350 1.5 1 16.65 69.2 NOTES: (1) Taken from Table 5.2-1 as follows : i Load Comb i n a t io n EON 1 = 1. 4 D + 1. 7 L + 1. 5 S RV (2) Allowable Reinforcing Steel Stress = 54 KSI i i i LIMERICK GENERATING STATION I UNITS 1 AND 2 DESIGN ASSESSMENT REPORT a CONTROL STRUCTURE MARGINS FLOOR SLABS i FIGURE E.1-36 R E V.1, 09/82

CONTROL STRUCTURE FLOOR STEEL BEAM (1} l BENDING [ STRESS I ELEMENT ELEVATION GOVERNING STRESS MARGIN (3) NUMBER' (PT) STEEL SIZE EQUATION (2) (KSI)  % 42 200 W24 x 130 1 23.78 0.9 43 217 W30 x 210 7 29.90 7.7 44 237 W36 x 300 7 27.60 14.8 45 254 W36 x 245 7 28 .80 11.1 46 269 4 2" Girder 7 25.53 21.2 47 289 W36 x 160 7 27 .90 13.9

48 304 W36 x 194 7 30.00 7.4 49 332 38" Girder 7 24.80 23.5 50 350 W18 x 105 7 10.30 68.2 O NOTES

(1) All beams are A-36 steel.

(2) Taken from Table 5.3-1 as follows : Load Combination EON 1 = D + L + SRV j Load Combination EON 7 = D + L + E' + SRV + LOCA . (3) Allowable bending stresses for governing equations 1 and 7 are 24.0 KSI and 32.4 KSI, respectively. LIMERICK GENER ATING STATION l UNITS 1 AND 2 DESIGN ASSESSMENT REPORT iO CONTROL STRUCTURE MARGINS I FLOOR STEEL BEAM FIGURE E.137 R EV.1, 09/82

O CONTROL STRUCTURE SHEAR WALLS WALL COMRINED AXIAL SHEAR ELEVATION GOVERMING & BENDING STRESS STRESS MARGIN (PT) WALL MARK EOUATION(2) MARGIN ( % )(3) ( % )(4 ) 177 Mh 7a 2 12 200 Mh 7a 39 2 269 7 1 44 24 239 J l 1 48 19 177 Line 19.4 7a 23 0 239 Line 19.4 1 3.4 24 177 Line 26.6 l 7a l 28 0 NOTES: (1) Figure E.1-1 gives location reference. U (2) Taken from Table 5. 2-1 as follows : Load Combination EON 1 = 1.4D + 1.7L + 1.5 SRV Load Combination EON 7a = 1.0D + 1.0L + 1.0 Egg

                                               + 1.0 SRV & l.0 LOCA (3)   Allowable Reinforcing Steel Stress = 54 KSI Allowable Concrete Compressive Stress = 2.8 KSI (4)   Allowable Rein forcing Steel Stress = 51 KSI LIMERICK GENER ATING STATION UNITS 1 AND 2 DESIGN ASSESSMENT REPORT O                                                 CONTROL STRUCTURE MARGINS SHEAR WALLS FIGURE E.138                  REV 1,09/82 l

O CONTROL STRUCTURE STEEL PLATFORM (1) ELEMENT ELEVATION STEEL GOVERNING BENDING STRESS , NUMBER (PT) GRADE EQUATION (2) STRESS MARGINS 3) (KSI) (%) 51 1 313 W10x21 2 14.7 38.8 52 313 W12x27 2 18.9 21.3 53 322  : W12x27 2 10.7 55.4 54 340 l W8x24 2 10.0 58.3 55 350 W8x24 2 17.6 26.7 350 56 W10x54 2 10.6 55.8 O NOTES: (1) All beams are A-36 steel. (2) Allowable bending stress = 24 KSI (3) Taken from Table 5.3-1 as follows : Load Combination EON 2 = D + L + To + SRV l l LIMERICK GENER ATING STATION UNITS 1 AND 2 i , DESIGN ASSESSMENT REPORT CONTROL STRUCTURE MARGINS STEEL PLATFORM FIGURE E.139 REV.1,09/82

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