ML22112A051
| ML22112A051 | |
| Person / Time | |
|---|---|
| Site: | LaSalle, 05200002 |
| Issue date: | 01/31/1997 |
| From: | ABB Combustion Engineering |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20148A597 | List:
|
| References | |
| NUDOCS 9705090171 | |
| Download: ML22112A051 (1) | |
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O the System 80+
standardplant 4
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l Design ControlDocument 5
iO volume 5 f
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Combustion Engineering, Inc.
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O Copyright O 1997 Combustion Engineering, Inc.,
All Rights Reserved.
Warning, Legal Notice and Disclaimer of Liability The design, engineering and other information contained in this document have been prepared by or for Combustion Engineering, Inc. in connection with its application to the United States Nuclear Regulatory Commission (US NRC) for design certification of the System 80+5 nuclear plant design pursuant to Title 10, Code of Federal Regulations Part 52. No use of any such information is authorized by Combustion Engineering, Inc.
cxcept for use by the US NRC and its contractors in connection with review and approval of such application. Combustion Engineering, Inc. hereby disclaims all responsibility and liability in connection with unauthorized use of such information.
Neither Combustion Engineering, Inc. nor any other person or entity makes any warranty or representation to any person or entity (other than the US NRC in connection with its review of Con %stion Engineering's application) conceming such information or its use, except to the extent an express warranty is made by Combustion Engineering, Inc. to its customer in a written contract for the sale of the goods or services described in this document. Potential users are hereby wamed that any such information may be unsuitable for use except in connection with the performance of such a written contract by Combustion Engineering, Inc.
Such information or its use are subject to copyright, patent, trademark or other rights of Combustion Engineering 'nc. or of others, and no license is granted with respect to such rights, except that t ,e US NRC is authorized to make such copies as are necessary for the use of the US NRC and its contractors in connection with the Combustion Engineering, Inc. application for design certification.
Publication, distribution or sale of this document does not constitute the performance of engineering or other professional services and does not create or estsblish any duty of care towards any recipient (other than the US NRC in connection with its review of Combustion Engineering's application) or towards any person affected by this document.
For information address: Cornbustion Engineering, Inc., Nuclear Systems Licensing, 2000 Day Hill Road; Windsor, Connecticut 06095 O
i System 80+ Design ControlDocument :
A Introduction LJ Certified Design Material 1.0 Introduction 2.0 System and Structure ITAAC 3.0 Non. System ITAAC 4.0 Interface Requirements 5.0 Site Parameters Approved Design Material- Design & Analysis 1.0 General Plant Description 2.0 Site Characteristics 3.0 Design of Systems, Structures & Components 4.0 Reactor 5.0 RCS and Connected Systems 6.0 Engineered Safety Features 7.0 Instr usentation and Control 8.0 Electric Power 9.0 Auxiliary Systems 4
10.0 Steam and Power Conversion 11.0 Radioactive Waste Management O- 12.0 Radiation Protection
():: 13.0 14.0 Conduct of Operations Initial Test Program 15.0 Accident Analyses 16.0 Technical Specifications
-17.0 Quality Assurance 3 18.0 Human Factors I 19.0 Probabilistic Risk Assessment )
20.0 Unresolved and Generic Safety Issues Approved Design Material - Emergency Operations Guidelines 1.0 Introduction 2.0 Standard Post-Trip Actions 3.0 Diegnostic Actions 4.0 Reactor Trip Recovery 5.0 Lor,s of Coolant Accident Recovery 6.0 Steam Generator Tube Rupture Recovery ;
' 7.0 Excess Steam Demand Event Recovery l
. 8.0 Loss of All Feedwater Recovery l 9.0 _ Loss of Offsite Power Recovery 10.0 _ Station Blackout Recovery 11.0 Functional Recovery Guideline l
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, the System 80+
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- Approved Design Material O Design & Analysis
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Combustion Engineering, Inc.
System 80+ oesign contror Document g
3.8 Design of Category I Structures t]
((The COL applicant referencing the System 80+ Standard Design will provide site and plant-specific structural design information. Information provided will include:
- Site related design parameters,
- Foundation mat construction procedures,
- Allowable loads for designated laydown areas,
- Final electrical layout drawings to verify load allowances for cable trays,
- Procedures for quality control inspections,
- Documentation of radiographic examination of welds,
- Site-specific design specifications, including expansion anchors,
- A steel containment vessel as-built structural analysis report, and
- A Seismic Category I structural analysis report.))1 I 3.8.1 Concrete Containment This section is not applicable to the System 80+ Standard Design. For a description of the containment, see Section 3.8.2. For a description of the containment shield building, see Section 3.8.4.
3.8.2 Steel Containment 3.8.2.1 Descdption of the Containment 3.8.2.1.1- General The contaimnent is a spherical welded steel structure supported by embedding a lower segment between the containment internal structures concrete and the reactor building subsphere concrete. There is no structural connection between the free standing portion of containment and the adjacent structures other than penetrations and their supports. The lateral loads due to seismic and other forces are transferred to the foundation concrete by shear bars, friction and bearing. The diameter of containment is 200 ft. The plate nominal thickness is 1.75 inches. The transition region, where the free standing portion of the steel containment vessel enters the concrete, has a plate thickness of 2 inches. The containment is shown on the plans and elevations of Figures 1.2-2,1.2-3,1.2-5,1.2-6,1.2-7 and 1.2-9.
The arrangement of the Nuclear Island structures, which includes containment and defines critical dimensions, flood barriers, and fire barriers, is shown in Figure 3.8-5.
1 COL information item; see DCD Introduction Section 3.2.
.%i Dee&n nienwW - Du&n of SSC Page 3.8-1
I System 80+ Design controlDocument The spherical shell plate segments will be shop fabricated and field welded. These plates will be approximately 25 feet long and 13 feet wide and can weigh as much as ten tons each; however, these dimensions will vary depending upon the plate location. Two or more plates may be assembled and field welded on the ground and then erected. A vast majority of penetration assemblies will be shop welded to the vessel plates, while others will be attached to the vessel in the field. Vessel plate will be thickened around the penetration to compensate for the openings. Where there is a cluster of penetrations in the same plate segment, the entire segment may be fabricated out of the thicker plate, tapered to 1.75 inches at the edges. The additional thickness will depend upon the nominal size, thickness and location of the penetration sleeve and shall be in accordance with ASME Boiler and Pressure Vessel Code (ASME Code) requirements (Reference 1).
The 2 inch thick portion of the steel containment vessel in the transition region will be shop fabricated and welded. The longitude plate welds will be 2 inch welds and will be postweld heat treated. The top and bottom edges of these 2 inch plates will be tapered to 1.75 inches.
3.8.2J .2 Anchorage Region The containment behaves as an independent, free-standing structure above elevation 91+9. Below elevation 91+9, the vessel is encased between the base slab of the internal stmetures and the shield building foundation. Radially extending shear bars are welded to the interior and exterior faces of the contaimnent vessel in the embedded region to provide restraint against sliding.
In the transition region, a compressible material is provided as shown in Figure 3.8-1 to eliminate excessive bearing loads on the concrete as well as to reduce the secondary stresses in the vessel at this i location. This compressible material extends an arc length of 2 feet into the concrete region, from j elevation 90+3 to elevation 91+9. The range of stiffness for the compressible material is determined i by modeling the system 80+ steel containment vessel as an axisymmetric finite element model using linearly clastic thin shell elements and the compressible material as a uniaxial tension-compression spring j element. Design basis accident pressure and dead weight loads are applied and the spring stiffness varied (
such that the stress intensities at extreme fibers do not vary by more than 5%. Since the seismic forces i are not included in the axisymmetric model, and the maximum stress intensity is nearly doubled when ;
the seismic forces are included in the 3-D model in the ASME Service Load analysis, this is j approximately equal to a variation of 10% for stress intensities at extreme fibers when seismic forces are j considered. The compressible material has an average stiffness in the range of 67.5 psi /in. to 360 psi /in. l The stiffness of the compressible material is verified to be in this range at each containment inservice l inspection interval. If material degradation has occurred, it is replaced, or an evaluation is performed to determine the effect on the steel containment vessel stress analyses. l The containment shell plate has a thickness of 2 inches in the transition region for corrosion allowance. I The 2 inch thick plate extends an additional 2 inches along the shell above and below the compressible material. The vessel plate thickness in the embedded zone is the same as in the free standing region.
3.8.2.1.3 Containment Penetrations Containment penetrations are designed for the Severe Accident assumptions identified in Section 19.11.3 in addition to the requirements given in this section.
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Appvemt Design Material Design of SSC Page 3.8-2 l
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System 80+ : Deslan ControlDocument }
l gO 3.8.2.1.3.1 Equipment Hatch The equipment hatch is composed of a cylindrical sleeve in the containment shell and a dished head 22 feet in diameter with mating bolted flanges. The flanged joint has double seals with an annular space {
for pressurized leak testing in accordance with 10CFR50, Appendix J. ,
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. The equipment hatch is designed and fabricated in accordance with Section III, Subsection NE of the ASME Boiler and Pressure Vessel Code. The equipment hatch is tested with the containment vessel.
Seals are designed to maintain containment integrity for Design Basis Accident conditions, including ;
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pressure, temperature, and radiation.
Details of a typical equipment hatch are shown in Figure 3.8-I.
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3.8.2.1.3.2 Penonnel Locks ;
i Two personnel locks 10 feet in diameter are provided. Each lock has double doors with an interlocking system to prevent both doors being opened simultaneously. Remote indication is provided to indicate the l
. position of each door. Double seals are provided on each door with an annular space for pressurized leak !
testing in accordance with 10CFR50, Appendix J.
i s
The personnel locks are welded steel subassemblies designed, fabricated and tested in accordmce with Section III, Subsection NE of the ASME Code. Seals are designed to maintain containment integrity for l.
- Design Basis Accident conditions, including pressure, temperature, and radiation.
Details of a typical personnel lock are shown in Figure 3.8-1.
3.8.2.1.3.3 Fuel Transfer Penetration ;
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A fuel transfer penetration is provided for transfer of fuel between the fuel pool and the Containment fuel
. transfer canal. The fuel transfer penetration is provided with a double sealed blind flange in the transfer i canal and a gate valve in the fuel pool. An annular space is provided between the double seals on'the- l t blind flange for pressurized leak testing in accordance with 10CFR50, Appendix J. The fuel transfer tube penetration sleeve and flanges are designed, fabricated and tested in accordance with Section III, Subsection NE of the ASME Code. The fuel transfer penetration is shown in Figure 3.8-2.
3.8.2.1.3.4 Mechanical Penatrations t
Mechanical penetrations are treated as fabricated piping assemblies meeting the requirements of the ASME Code,Section III, Subsection NE, and Subsection NC.
The process line and penetration flued head making up the pressure boundary are consistent with the system piping materials; fabrication, inspection, and analysis requirements are as required by the ASME Code,Section III, Subsection NC. _ All welds on the process pipe are accessible for inspection in accordance with the ASME Code,Section XI, i
High energy lines'and selected engineered safety system and auxiliary lines require the typical " Hot !
Penetration" assembly shown on Figure 3.8-2 which features an exterior guard pipe for protection of '!
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other penetrations in the building annular space. Other lines use the typical " Cold Penetration" assembly also shown in Figure 3.8-2 since a leak into the annular space would not cause a personnel hazard or l damage other penetrations in the immediate area. Mechanical bellows are provided where necessary to 4emoed nee 6n anoeuw.conon er ssc rose 2.s-2
System 80+ Design ControlDocument minimize reactions on the steel containment vessel. The bellows are designed to maintain containment integrity under all load conditions.
Mechanical penetrations are leak tested in accordance with 10CFR50 Appendix J.
3,8.2.1.3.5 Containment Electrical Penetration Assemblies Medium voltage containment electrical penetration assemblies for reactor coolant pump power (shown in Figure 3.8-2) use sealed bushings for conductor seals. The assemblies incorporate dual seals along the axis of each conductor. Low voltage power, control, and instrumentation cables enter the containment vessel through containment electrical penetration assemblies which are designed to provide two leak-tight barriers in series with each conductor.
All containment electrical penetration assemblies, including seals, are designed to maintain containment integrity for Design Basis Accident conditions, including pressure, temperature, and radiation. Double barriers permit testing of each asseinbly in accordance with 10CFR50 Appendix J to verify that containment integrity is maintained.
The containment electrical penetration assemblies are designed, fabricated and tested in accordance with IEEE-317. The pressure boundary portion of the assembly is designed, fabricated and tested in accordance with Section III, Subsection NE of the ASME Code.
3.8.2.2 Applicable Codes, Standards, and Specifications
((7hc design))2, materials, (( fabrication, erection]}2, nspection, testing, and inservice surveillance ((of the steel containment andpenetrations is covered by thefollowing codes2, standards, specifications, and regulations: Codes Title ASME Boiler and Pressure Vessel Code, Section II, " Material Specifications" [lASME Boiler and Pressure Vessel Code, Section Ill, Division 1, Subsection NE, " Class MC Components"}}* ASME Boiler and Pressure Vessel Code, Section V, " Nondestructive Examination" ASME Boiler and Pressure Vessel Code, Section IX, " Welding and Brazing Qualifications" ASME Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWE " Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants" 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approwet Design Material- Design of S$C Page 3.8-4
Sv tem 80 + Dasign ControlDocument
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Regulatory Guides Title 1.11 Instrument Lines Penetrating Primary Reactor Containment (Safety Guide 11) 1.50 Control of Preheat Temperature for Welding of Low-Alloy Steel 1.54 Quality Assurance Requirements for Protective Coatings Applied to Water Cooled Nuclear Power Plants 1.57 Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components 1.63 Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants 1.84 Design and Fabrication Code Case Acceptability - ASME Section III, Division 1 1.85 Materials Code Case Acceptability - ASME Section III, Division 1 1.92 Combinations of Modes and Spatial Components in Seismic Response Analysis 1.141 Containment Isolation Provisions for Fluid Systems 1.147 Inservice Inspection Code Case Acceptability - ASME Section XI, Division 1
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O Regulations 10CFR50 Appendix A - General Design Criteria for Nuclear Power Plants 10CFR50 Appendix J - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors Standards IEEE IEEE-317- Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations 3.8.2.3 Loads and Loading Combinations The loads and loading combinations for the analysis and design of the containment vessel are in accordance with Subsection NE, Section III, of the ASME Code and Regulatory Guide 1.57 (Reference 12). In addition, the specified loads and loading combinations are in accordance with The NRC Standard Review Plan, Section 3.8.2, II.3 (Reference 11). The loads and loading combinations are sunuaarized in Tables 3.8-1 and 3.8-2.
* - Dead Load and Construction Loads The dead load includes the weight of the contaimnent vessel and all permanent attachments.
.[ " Construction loads are those loads imposed on the containment vessel only during construction. A typical construction load is the shoring load induced by the formwork for the containment shield building concrete dome. see.mr oeew mww- caew or ssc rare .ts-s
System 80+ Design ControlDocurnent
- Thermal Loads The containment vessel is subject to thermal loads during normal operation of the unit. The maximum operating temperature can reach 110'F.
- Seismic Loads The containment vessel is loaded by simultaneous seismic events in two orthogonal horizontal directions and the vertical direction. Seismic loads are described in Section 3.7.
- External Pressure A vacuum load can be imposed on the Containment Vessel by an inadvertent actuation of the Containment Spray System during normal unit operation. The design vacuum pressure is 2.0 psig.
- Design Basis Accident The Design Basis Accident Loads are based on the peak pressure and temperature developed inside Contaimnent as a result of a rupture in the primary coolant system up to and including a double-ended rupture of the largest pipe (a Loss-of-Coolant-Accident or LOCA) or a main steam line break. The containment vessel design pressure is 53 psig and the design temperature is 290*F. See Chapter 6 for details of the Containment Design Basis Accident.
- Combustible Gas Loads The containment vessel is subject to the consequences of uncontrolled hydrogen-oxygen recombination as specified in the Code of Federal Regulations,10 CFR 50.<4.
- Localized Loads Penetration loads, piping loads and jet impingement loads are all localized loads applied to the containment vessel. Penetration and piping loads are due to the reactions at penetrations, pipe supports / restraints and other attachments welded to the shell. Jet impingement loads are due to fluid jets caused by the rupture of small diameter piping adjacent to the containment vessel.
3.8.2.4 Analyses and Results The steel containment vessel (SCV), including its penetrations, is designed and analyzed to satisfy the requirements of ASME Code Section III, Division 1, Subsection NE, Class MC Components. The containment vessel is analyzed to determine the stress levels and stability factors of safety resulting from the application of specified loads. The vessel is analyzed using thin shell finite element methodology. The ANSYS computer code (Reference 2) is used to generate the geometry of the shell, determine stress levels in the shell, and evaluate shell stability. O Approved Design Material Design of SSC page 3.s-6
4 i System 80+ Design ControlDocument 3.8.2.4.1 Description of Finite Elanent Models 3.8.2.4.1.1 3-D Finite Element Model j
- A pictorial presentation of the containment vessel 3-D finite element model is given in Figure 3.8-3. An eight node isoparametric thin shell element is used. Fixed boundary conditions are applied in the model f at the 90'+3" elevation. 1 I
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; The weight of the personnel airlocks and the equipment hatch penetrations is mcluded in the model by increasing the density of the hell in the region of the penetration. Live load is included in the weight of the penetrations. The penetrations themselves are not modelled explicitly. The thickness of the shell
- in the region of the personnel airlocks and the equipment hatch is increased using the area replacement rules in the ASME Code. The containment spray mass is included in the upper reg *:u of the model by distributing additional mass at the appropriate locations in the dome. The mass of the piping and the-g electrical penetrations in the lower region of the sphere is accounted for by increasing the density of the
!- shell elements in that region. The stiffness of the compressible material at the base of the containment vesse l is modeled as a two-directional spring.
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4 Although the transition region of the SCV is 2 inches thick, the 3-D finite element model has a uniform shell thickness of 1-3/4 inches. The additional 1/4 inch of material in the transition region is for corrosion allowance only and credit for this additional thickness is not included in the 3-D analysis. The l axisymmetric model described in Section 3.8.2.4.1.2 is used to evaluate the effects of the change in material thickness in the transition region. 3.8.2.4.1.2 Axisymmetric Finite Element Model The containment vessel is modelled with thin shell axisymmetric finite elements. The model is fixed at the base, elevation 90'+3", with two-directional spring elements to represent the compressible material at the base of the containment vessel. The meridian rnodeled is the one corresponding to the equipment hatch since it has the largest mass. The mass of the equipment hatch, which includes live load, piping , and electrical penetrations and the containment spray system, are included in the model by adjusting the ' density of the elements in the appropriate regions. > The axisymmetric model is used for the Service Level A stress analysis and to determine the effects of a thickening the transition region of the containment vessel from 1.75 inches to 2 inches. The 2 inch 4 thickness is to provide additional material for corrosion allowance. 3.8.2.4.2 Reduced Load Combinations The load combinations in Table 3.8-2 include loads identified in the NRC Standard Review Plan. Not all loads are applicable to the System 80+ design. The following is a listing of the reduced load combinations with an explanation of why certain loads are not considered in the combinations.
- Test Condition f
I All loads in the combination given in Table 3.8-2 are applicable to the System 80+ design. o l 9 _b - -_'W afe4Bnief
- Destese of SSC Page.T.8-7 l f
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System 80+ Design ControlDocument
- Design Condition The T, load is not included in the combination because thermal loads are considered as secondary stresses. The ASME Code does not require an analysis of secondary stresses for the Design condition.
l The pipe reaction R, and R loads on the steel containment are eliminated as described in the System 80+ design by the use of bellows and/or guard pipes at the containment penetrations where pipe reactions would exist. The reduced load combination is: D+L+P,
- Service Conditions Service Level A:
Pipe reactions R, and Roare eliminated as described in the Design combination. The stresses resulting from the operating temperature and pressure loads, To and Po, are enveloped by the accident temperature and pressure loads and therefore are not analyzed separately. The primary membrane stress evaluation for Service Level A is the same as the Design Condition. When evaluating secondary stress effects, the reduced load combination is: D + L + T, + P, - Service Level C: Pipe reactions R, and R, are eliminated as described in the Design combination. The stresses resulting from the operating pressure loads, Po , are enveloped by the accident pressure loads and therefore are not analyzed separately. The T, and T oloads are not included in the combination because thermal loads are considered as secondary stresses as described in the Design combination. The ASME code does not require an analysis of secondary stresses for Service Level C. The reduced Service Level C loads are the same as the reduced Service Level D loads. The ASME Service Level D allowable stresses are lower than the Service Level C allowable stresses; therefore, the analysis is performed for the reduced Service Level D loading combination and compared with the lower allowable stresses of Service Level D. O Amond Desw ussuw- Desion or ssc mes) roue 3.s-a
. Sy? tem 80+ Design ControlDocument l
Service Level D: Pipe reactions, operating loads, and thermal loads are eliminated as described in the Service Level C combination. ; 1 The pipe rupture loads, Y,, are eliminated in the design by the use of rupture restraints and guard pipes in the System 80+ design. The jet impingement loads, Y, j are eliminated by the use of . guard pipes and, where necessary, jet impingement protection devices. The containnent shell l is protected from the missile loads, Y,,, by the crane wall inside the containment vessel and the head area cable tray system. ' i The reduced load combination is: l D + L + P, + E' I
-* Construction Loads All loads in the combination given in Table 3.8-2 are applicable to the System 80+ design.
- Stability Considerations !
The first stability combination is an evaluation of Service Level C and D conditions as a result of the combined effects of the SSE (E') loads and containment external pressure (P,) due to inadvertent actuation of the containment spray system. This combination produces higher dp compressive stresses than the combination with the design basis accident pressure (P.). J Pipe reaction loads are eliminated as described in the Service Level C combination. The thermal loads associated with the external pressure loads, T,, in the combination with the seismic loads, are not included. The combination is evaluated against Service Level C and D limits. The ASME Code does not require an analysis of secondary stresses for Service Level C l or Service Level D combinations. The reduced load combination including seismic is: D + L + P, + E' The reduced Service Level D loads are the same as the reduced Service Level C loads. The Service Level C combination has a higher therefore, the Service Level C safety factor is ; demonstrated in the analysis. The second stability combination is an evaluation with Service Level A conditions to evaluate the effect of the compressive stresses from accident temperature loads near the base. The reduced load combination with the accident temperature load included is: D + L + T. + P, .
'i Annemer onete nanonier oneen or ssc rose 2.s s
1 System 80+ Design ControlDocument ; I
- Ultimate Capacity Considerations 1
All loads in the combination given in Table 3.8-2 are applicable to the System 80+ design. l 1
- Combustible Gas Load Considerations All loads in the combination given in Table 3.8-2 are applicable to the System 80+ design.
- Overturning and Sliding Considerations All loads in the combination given in Table 3.8-2 are applicable to the System 80+ design. l 3.8.2.4.3 Analysis Description
- Testing Condition (D + L + P, + T,)
The Test Condition combination analysis is performed using the 3-D finite element model. The combined dead and live load analysis (D + L) is performed using the static analysis option of ANSYS by specifying the acceleration due to gravity in the vertical direction. The test pressure analysis (P) is performed using the static analysis option of ANSYS. The only load that is applied to the structure is a 58.3 psi internal pressure that is applied to the inside surface of each element. ; The test temperature analysis (T) is performed using the static analysis option of ANSYS. The only load that is applied to the structure is a uniform temperature of 110 "F. The results from each of the analyses are then combined using the Test Condition combination ) equation. See Table 3.8-3A. l
- Design Condition (D + L + P.) ;
The Design Condition combination analysis is performed using the 3-D finite element model. l The combined dead and live load analysis (D + L) is performed using the static analysis option of ANSYS by specifying the acceleration due to gravity in the vertical direction. The accident pressure load analysis (P,) is performed using the static analysis option of ANSYS. The only load applied to the structure is a 53 psi internal pressure that is applied to the internal surface of each element. The results from each of the analyses are then combined using the Design Condition combination equation. See Table 3.8-3A.
- Service Conditions Level A: (D + L + T, + P,)
Aptwowcf Design Material- Design of SSC Page 3.8-10
i System 80+' Deslan ControlDocument - l The Service Level A analysis is performed using the Axisymmetric finite element model. All j loads are applied simultaneously. The combined dead and live load (D + L) is applied by specifying the acceleration due to gravity in the venical direction. The accident thermal load (T,), 290 'F, is applied as a uniform temperature in all of the elements. The accident pressure (P,),53 psi, is applied as a uniform pressure on the inside face'of each ' element. j 1 The Service Level A stress evaluation is performed for material thicknesses of 1-3/4 inches and ;
. 2 inches in the transition region. .
The Service Level A stress results are presented in Tables 3.8-3A and 3.8-3B. Level D: (D + L + P, + E') .; The Service Level D analysis is performed using the 3-D finite element model, i I The combined dead and live load analysis (D + L) is performed using the static analysis option of ANSYS by specifying the acclenion due to gravity in the venical direction. ; i The accident pressure load analysis (P.) is performed using the static analysis option of ANSYS. !
' The only load applied to the structure is a 53 psi internal pressure that is paplied to the internal ;
surface of each element. l A modal analysis with spectrum input, in the three orthogonal excitation directions, is performed - to determine the eigenvalues (natural frequencies) and eigenvectors (mode shapes) of the steel ! containment vessel. Modal results for each excitation direction are combined according to R.G. l 1.92 (Reference 14) requirements. In addition, the missing mass of the structure (that mass not ! being excited in the dynamic analysis) is accelerated at the Zero Period Acceleration of the input i
< response spectra. The missing mass response for each excitation direction is absolutely summed i with the combined modal results. The combined results from each excitation direction are then l combined using the Square Root of the Sum of the Squares (SRSS) method to give one set of j
- dynamic results that include both stress levels and displacements.
l The results from each of the analyses are then combined using the Service Level D combination equation. See Table 3.8-3A. ! e Construction Loads (D + C) ; i The analysis of the steel containment vessel subjected to construction loads is performed using j i~ a three dimensional finite element model of a partially erected steel contaimnent vessel. Loads are applied statically to determine the maximum membrane stress intensity. Material propenies , at 70'F are used. , Wet concrete and formwork and shoring from construction of the shield building dome are evaluated as resting on top of the steel containment vessel. Wind loads that exist prior to erection ,
, ' of the shield building 'are determined using the methods of References 6 and 7. A wind speed 4
of 110 mph is used. The results in Table 3.8-3A'are the maximum stress intensity from the , combination of the wind and shoring loads. l t b opnmd onow annenw.w or ssc rene 2.s-rr { g y -
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. System 80 + Design ControlDocument leads due to the lifting and supporting of preassembled sections are evaluated using methods from Welding and Research Council Bulletin #107 (Reference 23).
- Stability Considerations Large deflection clastic-plastic analyses of the 3-D finite element model of the steel containment vessel are completed to determine the stability safety factors due to Service Level A and Service Level C and D conditions. The computational procedure factors all the applied norainal loads in incremental load steps until a non-converged solution is obtained. Within each load step, the ANSYS code iterates until convergence criteria are satisfied for both the large deflection solution and the clastic-plastic solution. The total load levels at the last converged load step are divided by the nominal load levels to determine the safety factor.
Imperfections in the containment vessel stability analysis are modeled considering the forming and erection tolerances in the ASME Code, Subsection NE, Article NE-4221 and the size of the individual plates. The ASME Code fabrication tolerance is a maximum radial imperfection with an amplitude of one thickness. This is measured with a true circular template of the outside or inside radius. The ASME erection tolerance is plus or minus 1 percent of the nominal diameter at any cross section. This tolerance allows for global variations in the structure geometry which have less of an effect on buckling than local imperfections. The local imperfections are determined by considering the individual plates which are fabricated with longitudinal dimensions that are 12 to 20 feet near the base. A half wavelength value of 8 feet allows for two radial imperfections in each individual plate, assuming a reasonable plate length of 16 feet; in other words, a full sine wave imperfection in each plate. The maximum amplitude allowed by the forming tolerance using a true circle template would be one thickness o'r 1.75 inches. The nonlinear material properties are determined by applying plasticity reduction factors found in ASME Code Case N-284 (Reference 22), to the stress strain curve. These redu:tions account for the potential residual compressive stresses which would promote the onset of buckling. These factors, found in paragraph -1600 of the Code Case, are applied to the modulus of elasticity to determine a reduced tangent modulus for a given stress level. Service Level A Stability: (D + L + T, + P,) For the Service Level A stability analysis the following nominal Imds are applied:
- Design basis accident pressure (P,) = 53 psi.
- Design basis accident temperature (T ) = 290 'F.
- Dead weight and live loads (D + L)
These loads are applied simultaneously to the model. The combined dead and live loads (D + L) are applied by specifying the acceleration due to gravity in the vertical direction. The accident pressure (P ) is applied on the inside surface of each element. The accident thermal load (T ) is applied as a uniform elemental temperature. Ainproved Desigrs Meterini Design of SSC Page 3.812
System 80+ Deslan contrar Document
. l' The Service Level A stability analysis is completed for a full sine wave with a half wavelength-
= 8 feet and a_ peak to peak amplitude of 1.75 inches. The stability safety factor for Service Level A conditions is 3.0.
I Service Level C and D Stability: (D + L + P, + E') For the Service Level C and D stability analysis the following nominal loads are applied: l o External pressure due to containment spray actuation (P,) = 2 psi. ' e ~ Dead weight and live loads (D + L) i e- . SSE loads (E') These loads are applied simultaneously to the model. The combined dead and live loads (D +
- L) are applied by specifying the acceleration due to gravity in the venical direction. The external j pressure (P,) is applied on the outside surface of each element. 1 The seismic loads (E') are applied statically as nodal forces determined by multiplying the zero !
period accelerations (ZPAs) from the Soil Structure Interaction Analyses by the mass attributed to each node. Inenial forces due to SSE excitation are applied to all nodes of the model as ; horizontal and venical forces. ,
- The Service Level C and D stability analyses are completed for two types of imperfections:
\
e A full sine wave with a half-wavelength = 8 feet and a peak to peak amplitude of 1.75 - inches.
.e A half sine wave with a half-wavelength = 16 feet and a peak amplitude of 1.75 inches. !
U The resulting Service Level C stability safety factors for these two types of imperfections are 2.70 l l and 2.74, respectively. e e Ultimate Load Considerations (D + L + P,) , 1 The Ultimate Capacity is determined using an elastic analysis with the axisymmetric model. All j loads are applied simultaneously in a static manner. The dead and live load is applied as an l increase in the density in the appropriate regions. The internal pressure load, which is applied to the inside face of each element, is increased until the maximum stress intensity reaches the Service Level C allowable membrane stress intensity for the given temperature. The ASME j Service Level C allowable stress intensity value is the nominal yield stress value for the ; temperature given. Temperature values of 150'F,290*F (Design Basis Accident Temperature), l 350'F, and 450*F are evaluated. The material propenies associated with the temperatures are used. The internal pressure value which results in a maximum stress intensity equal to the : Service Level C allowable membrane stress intensity is the ultimate pressure capacity, P, The j results are summarized in Table 3.8-3D. , I ( d Amarowsef Desen Afessnist Desgn er SSC page 3.3 73 , 8
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System 80+ Design ControlDocument
- Combustible Gas Load Considerations (D + L + P, + P,)
The Combustible Gas Loading is evaluated using an clastic analysis with the axisymmetric model. The dead and live load is applied as an increase in the density in the appropriate regions. The peak pressure from hydrogen combustion and the design basis pressure is added together and applied as an internal pressure to the inside face of each element. All loads are applied simultaneously in a static manner to determine the maximum membrane stress intensity. The results are summarized in Table 3.8-3A.
- Containment Ovenurning and Sliding (D + L + E')
The containment is analyzed for sliding and overturning of the interior structures against the steel containment and the interior structures and steel containment against the lower concrete dish structure outside of containment. The interior structures and the steel containment are modelled using a lumped mass stick model. The inertial forces are determined by multiplying nodal masses by the horizontal and vertical floor accelerations from the soil structure interaction analyses. These accelerations already include the effects of structural rocking. Two potential failure planes are considered. One plane is defined by the resultant of the seismic forces and the vertical axis. The other failure plane is perpendicular to the direction of the resultant moment. The imposed seismic and dead weight forces are resisted by normal and tangential reaction forces at the steel containment interface. Stability against overturning is maintained as long as the reaction forces are below the base of the crane wall, a location of substantial stiffness. The tangential reaction force defines the sliding potential. The stability check is performed for each step of the time history for the critical soil cases and control motions. Additionally, three combinations of the time history loading in the vertical direction are considered. The first combination is the vertical time history with the signs retained, and the second is with the signs reversed. Finally, the maximum vertical seismic response throughout the time history is determined and 0.4 times this response is applied in the upward direction. Shear bars are designed to resist the full sliding forces without consideration for frictional resistance. The safety factor against sliding and overturning is 2.4. e Nonaxisymmetric and Localized Loads There are no nonaxisymmetric loads applied to the steel containment vessel during a Design Basis Accident other than seismic loads. Localized loads applied to the containment vessel may be piping support / restraint reactions, reactions from other attachments, jet impingement loads, etc. The ANSYS computer program is used to calculate the local stresses caused by these loads, which are then included in the appropriate loading combination. The System 80+ containment shell discontinuities occur around changes in plate thicknesses and reinforcement inserts. A detailed analysis is performed to ensure the strains in these areas are acceptable. The analysis is a large deflection elastic plastic analysis of a finite element model AppromiDesign Materia!- Destyn of SSC Page 3.814
f L I l System 80 + -' oestan controlDocument . of the affected region. The region is loaded to the ultimate load determined for the containment 'j shell structure. The finite element results are interpreted in terms of a failure criterion based on the ultimate strain, eg of the material based on a uniaxial tensile test result. If actual material test results are ' available at the time of analysis the actual test strain values are used, otherwise nominal ASTM 4 values are used for the ultimate strain. The peak calculated strains, eg, are determined from the
. finite element model loaded to the nominal containment ultimate pressure. This value is
- compared to the material ultimate strain value adjusted by a knockdown factor, K, based on
. relativ: level of sophistication of the finite element analysis, the difference between actual configuration and configuration modeled, and variations in the material property data.
The peak calculated strains are less than or equal to the ultimate strain divided by the knockdown factor: E
'pc'$ ,
The knockdown factor for the finite element analysis is based on mesh refinement studies. The ' factor for the variance in actual and modelled configuration is based on the latest data available 4 (Reference 20). i 4 f
\
The calculated peak strain values with the knockdown factor applied is also compared to the Sandia test results in which strain values of approximately 2.8% were measured (Reference 21). 3.8.2.5 Structural Acet- Criteria
- Allowable Stress Limits for Test, Design, and Service Level Conditions Allowable stresses are established in accordance with the ASME Code, Section III, Division I, Subsection NE and Section 3 . 8 .2 , 11.5 of the NRC Standard Review Plan. Stress intensity l allowable and results are summarized in Table 3.8-3E.
a .
- Stability Safety Factors i
The factors of safety for stability loads is 3.0 for Service Level A loads, and 2.5 for Service Level C loads. These values are established from Article NE-3222 of the ASME Code. The ~ stability analyses results are summarized and compared to these factors of safety in Table 3.8-3C. i i i
- Construction Loads 1 Maximum stress results from construction load combinations are required to be less than 0.9 times the yield stress (Sy ) established in the ASME Code. Construction load stress results are compared to the allowable stress in Table 3.8-3E.
w:ooep unend.ceew a ssc rose 2.s-1s : I
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System 80+ Design ControlDocument
- Ultimate Capacity The ultimate capacity is defined as the pressure which when combined with dead load results in stresses equal to the ASME Service Level C stress intensity limit. Ultimate capacity analyses results are summarized and compared to the allowable stress intensities in Table 3.8-3D.
- Combustible Gas Loads Maximum stress results from the combustible gas load combination are required to satisfy the ASME Service Level C stress intensity limits. Combustible gas load analyses results are compared to the allowable stress intensities in Table 3.8-3E.
- Overturning and Sliding A safety factor of 1.1 must be demonstrated in analyses which considers the sliding of the interior structure against the steel containment vessel and the sliding of the interior structure and steel containment vessel against the containment support structure. Analyses results are included in Section 3.8.2.4.3 (Containment Ovenurning and Sliding). A safety factor of at least 1.1 is demonstrated in all analyses.
- StrLctural Analysis Report A structural analysis report will be prepared for Containment as required by ASME Section III.
This report will document that the containment structure meets the requirements specified in Section 3.8.2 and that design changes and identified construction deviations which could potentially affect the structural capability of the containment structure, have been incorporated into the structural analysis. The following records will be reviewed, as applicable:
- Construction records stating material properties for containment materials
- As-built structure dimensions and arrangements
- Design docurnents for the structure Deviations from the design are acceptable provided the following acceptance criteria are met:
- An evaluation is performed (depending on the extent of the deviations, the evaluation may range from the documenting of an engineering judgement to performance of a revised analysis and design), and
- the structural design meets the requirements specified in Section 3.8.2.
Site-specific information will include an as-built stmetural analysis report for the steel containment vessel. 3.8.2.5.1 Welding and Weld Acceptance Cdteria Welding activities shall be in accordance with the requirements of Section III, Subsection NE of the ASME Code. Asproved Design Material . Design of SSC Page 3.846
_l System 80+ Design Control Document
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Radiographic examinations will be accepted by a nondestructive examination (NDE) Level III examiner ( y prior to final acceptance. See also Section 3.8. 4 Confirmation that facility welding activities are in compliance with the certified design commitments shall include verifications of the following by individuals other than those who performed the activity:
- 1. Facility welding specifications and procedures meet the applicable ASME Code requirements.
- 2. Facility welding activities are performed in accordance with the applicable ASME Code ;
requirements. i
- 3. L Welding activities related records are prepared, evaluated and maintained in accordance with the
- ASME requirements. l
- 4. Welding processes used to weld dissimilar base metal and welding filler metal combinations are compatible for the intended applications, j
- 5. The facility has established procedures for qualifications of welders and welding operators in accordance with the applicable ASME Code requirements.
1
; 6. Approved procedures are available and are used for pre-heating and post-heating of welds, and those procedures meet the applicable requirements of the ASME Code. ;
a
- 7. Completed welds are examined in accordance with the applicable examination method required
- O by the ASME Code.
3.8.2.6 Matedals, Quality Control, and Special Construction Techniques 3.8.2.6.1 Materials The containment vessel materials are in accordance with Anicle NE-2000 of Subsection NE, " Class MC j Components," of the ASME Boiler and Pressure Vessel' Code, Section III, " Nuclear Power Plant Components." l The containment plate material is ASME SAS37 Class 2. This material is exempt from post-weld heat treatment requirements when plate thickness is less than or equal to 1.75 inches in accordance with Table NE-4622.7(b)-1 of the ASME Boiler and Pressure Vessel Code, Section III. When plate thickness exceeds 1.75 inches, post weld heat treatment shall be performed. The material will be impact tested in , accordance with Anicle NE-2300 of Section III of the ASME Code. Fabrication and erection of the containment vessel are in accordance with Anicle NE-4000 of Section III of the ' ASME Code. This includes welding procedures, procedure and operator performance ; qualifications, post weld heat treatment and tolerances. Nondestructive examination of welds and materials is in accordance with Anicle NE-5000 of Section III of the ASME Code. ep r o ws.mnw.o.mir, esse - p.e. 3.s-n
System 80+ Design Control Document 3.8.2.6.2 Quality Control The general provisions of the overall Quality Assurance program are outlined in Chapter 17. These are supplemented by the special provisions of the ASME Code for quality control as applicable to Class MC Components. The containment vessel is ASME Code stamped. Therefore, the ASME Code requirements for quality control have priority over those outlined in Chapter 17 in case of any conflict. 3.8.2.6.3 Special Construction Techniques The steel containment vessel may be assembled in sections in an area of the construction yard and then lifted and moved into place with a walker crane. This procedure allows the assembly of the containment to begin when the plates forming the lower hemisphere are delivered to the site. In this manner the containment assembly can proceed on a parallel path with the construction of the concrete subsphere region. The following options are two of several techniques that can be employed for placing the concrete for the dish pedestal supporting the containment:
- The concrete dish except for the top six inches can be placed. The containment vessel then would be placed on the support heads and pressure grouted.
- The containment vessel can be placed on the support heads and used as formwork for placing the concrete.
With either option care must be taken to prevent the floating of the containment. This is accomplished by filling the containment with water as the grout / concrete is placed. After the lower containment section has been placed, the construction of the interior structure can begin. The assembly of the containment sections will continue in the yard. As the work on the interior structure continues additional sections of the containment can be lifted into place. After the major equipment is placed in the interior structure the top section of the containment vessel can be set. The completed containment will then be used to support the scaffolding for the concrete dome of the shield building. 3.8.2.6.4 Corrosion Prevention The Steel Containment Vessel (SCV) plate nominal thickness is 1.75 inches except in the transition region between the SCV and the lower concrete dish where a 2 inch thickness is used. Between the SCV and concrete at elevation 91+9 is a gap in which a compressible material is placed. The transition area is at a 45 degree slope from the floor elevation which allows inspectability. The following are steps taken against corrosion.
- The non-embedded SCV is coated inside and outside (including the transition region) to minimize corrosion and to facilitate decontamination efforts.
l
- The concrete is sealed to preclude moisture. A visual inspection of coatings is performed. I e Visual inspections of containment base metal and welds are perfonned in accordance with ASME Section XI, Subsection IWE and 10CFR50 Appendix J. These are formal inservice inspection requirements. The portions of containment embedded in the concrete are exempt from these inspection requirements while the welds around the embedded penetrations are required to be i
inspected. Amroved Design Afsteriel- Design of SSC page 18 18 _
Sy~ tem 80+ Design ControlDocument 1%)
- Collection of moisture in the transition region is prevented by use of sloped floors and drains.
- The compressible material which is placed in the transition region between the steel and concrete ,
is removable. Once removed the material and SCV is inspected. I
- No equipment or ductwork is located such that it inhibits a visual inspection at the steel concrete interface for corrosion.
- For further precautionary measures and conservatism, the SCV is 2 inch in thickness in the transition region. This thickness is beyond design requirements and allows for a corrosion allowance of approximately 4 mils per year over a 60 year life. With an inspection program and maintenance of the coatings, corrosion is minimized in this region.
3.8.2.7 Testing and In-service Surveillance Requirements The containment vessel, personnel airlocks and equipment hatch are inspected and tested in accordance with the ASME Boiler and Pressure Vessel Code, Section III, Subsection NE. Penetrations are pressure tested as required for Subsection NC of the ASME Code. Periodic leakage rate tests of the containment are conducted in acccrdance with 10CFR50, Appendix J to verify leak tightness and integrity. These tests and other in-service inspection requirements are described in Section 6.2. Periodic in-service inspections are conducted in accordance with the ASME Boiler and Pressure Vessel Code, Section XI, Subsection IWE. O h 3.8.3 Concrete and Structural Steel Internal Structures ; 3.8.3.1 Description of the Internal Structures The internal structure is a group of reinforced concrete structures that enclose the reactor vessel and primary system. The internal structure provides biological shielding for the containment interior. The internal structure concrete base rests inside the lower portion of the containment vessel sphere. A description of various structures that constitute the internal structure is given in the following paragraphs. The details of the internal structure are shown in Figures 1.2-2,1.2-3,1.2-6,1.2-7 and 1.2-9. The internal stmetures are Seismic Category I structures with the exception of platforms that do not support Seismic Category I equipment and miscellaneous steel. These structures will be Seismic Category 11 structures. Seismic Category II structures are designed for the SSE using Seismic Category I criteria to prevent adverse interaction with other Seismic Category I structures, systems, and components. The arrangement of the Nuclear Island structures, which includes the internal structure and defines critical dimensions, ficod barriers, and fire barriers, is shown in Figure 3.8-5. The primary shield wall encloses the reactor vessel and provides protection for the vessel from internal missiles. The primary shield wall provides biological shielding and is designed to withstand the temperatures and pressures following LOCA. In addition, the primary shield wall provides structural support for the reactor vessel. The primary shield wall is a minimum of six feet thick. (v) s
? :ony neeww- onion or ssc rage 1s-1s
l System 80+ Design ControlDocumem l The secondary shield wall (crane wall) provides supports for the polar crane and protects the steel containment vessel from internal missiles. In addition to providing biological shielding for the coolant j loop and equipment, the crane wall also provides structural support for pipe supports / restraints and platforms at various levels. The crane wall is a right cylinder with an inside diameter of 130 feet and a height of 118 feet from its base. The crane wall is a minimum of four feet thick. The refueling cavity, when filled with borated water, facilitates the fuel handling operation without exceeding the acceptable level of radiation inside the containment. The refueling cavity has the following sub-compartments:
- Storage area for upper guide structure
- Storage area for core support barrel e Refueling cavity The refueling cavity, when filled with borated water, forms a pool above the reactor vessel. The reactor vessel flange is sealed to the bottom of the refueling cavity to prevent leakage of refueling water into the reactor cavity as described in Section 9.1.4.2.2.13. The fuel transfer tube connects the refueling cavity to the refueling canal. The refueling cavity is filled with borated water to a depth that limits the radiation at the surface of the water to acceptable levels during the period when a fuel assembly is being transferred to the Spent Fuel Pool. The shield walls that form the refueling cavity are a minimum of six feet thick.
The In-containment Refueling Water Storage Tank (IRWST) provides storage of refueling water, a single source of water for the safety injection and containment spray pumps and a heat sink for the Safety Depressurization System. The IRWST is dishlike in shape and utilizes the lower section of the Internal Structure as its outer boundary. The IRWST is provided with a stainless steel liner to prevent leakage. Design of the IRWST considers pressurization as a result of the containment systems Design Basis Accident. A full description of the IRWST is provided in Section 6.8. The holdup volume tank (HVT) is a rectangular shaped structural tank located between the primary shield wall and the IRWST inner wall. The HVT provides a collection point for leakage from components and piping not routed to the reactor drain tank or spills inside containment. A screen is provided at the top of the HVT to prevent debris from getting into the tank. The HVT has a sump with pumps to measure the leakage rate and route the liquid to the liquid waste management system. During an accident, the water from breaks and containment spray collect in the HVT and overflow into the IRWST. A description of the HVT is provided in Section 6.8. The operating floor provides access for operating personnel functions and provides biological shielding. Inside the crane wall, the operating floor is a reinforced concrete slab with a covered hatch that is aligned with hatches in the two lower floors. Outside the crane wall, the operating floor consists of steel grating. There are also reinforced concrete floor slabs at elevation 115+6 and elevation 91 +9 that connect the crane wall and the primary shield wall. The support systems for the reactor vessel, steam generators, reactor coolant pumps and primary loop piping are completely described in Section 5.4.14. The locations of the missile shield, hatch covers, and other removable structures are shown in Figures 1.2-2,1.2-3,1.2-6,1.2-7 and 1.2-9. The removable slabs and hatch covers are provided with suitable Approved Design Material Design of SSC Page 3.8-20
System 80+ oesign controlDocument O -Q tiedown devices to eliminate any possibility of these items becoming missiles in case of a seismic event or other loading conditions. 3.8.3.2 Applicable Codes, Standards, and Specifications Category I structures are designed as described in Appendix 3.8A using the codes and criteria listed in Table 3.8-4. 3.8.3.3 Loads and L-ding Combinations The loads and loading combinations used for the internal structures are shown in Section 5.0 of Appendix 3.8A. The internal structures are designed for the following loads:
- Dead load
- Equipment operating loads and other live loads
- Pipe reactions
- Seismic (See Section 3.7 for seismic criteria)
- Internal missiles (The internal structure is designed to withstand internal missiles as defined in x._./ Section 3.5.)
- Pipe rupture jet impingement
- Differential pressures between the reactor vessel cavity, pressurizer enclosure or In-containment Refueling Water Storage Tank and the remainder of the containment free volume.
- The greatest pipe rupture loads from (1) pipe breaks not eliminated by leak-before-break, (2) the largest through wall leakage crack in a high energy line (minimum 10 gpm) whether or not consideration of dynamic effec,ts is eliminated by leak-before-break for the line, or (3) the largest leak from another leak source, such as a valve or pump seal.
In addition, the reactor vessel support corbels have symmetrical reinforcing in the top and bottom to resist the upward forces resulting from a potential severe accident ex-vessel steam explosion in the reactor cavity. 3.8.3.4 Design and Analysis Proceduns The internal structure is designed for the loads and load combinations specified in Section 3.8.3.3 and Appendix 3.8A. The complete internal structure (and supporting substructure) is modeled with three-dimensional solid, plate or shell and beam finite elements using ANSYS or another suitable computer code. The forces and moments resulting from the applied static and dynamic loads are u.;ed to design (g \} the walls, slabs, beams and columns which make up the Internal Structure. The design is performed using either ACI 349 (Reference 3) or ANSI /AISC N690 (Reference 4) as amended by Section 3.8.4.5. Seismic Category Il structures are designed for the SSE using Seismic Category I criteria to prevent adverse interaction with other Seismic Category I structures, systems and components. kneroweet Design nietenfel Desipro of SSC Page 3.8-21
System 80+ Design C*ntrolDocument Design and analysis results for selected areas of the internal structures are presented in Appendix 3.8B. 3.8.3.5 Structural Acceptance Criteria The structural acceptance criteria for the Internal Structures is outlined in Section 3.8.4.5. 3.8.3.6 Materials, Quality Control, and Special Construction Techniques The design addresses the vertical alignment of the Secondary Shield Wali (Crane Wall) with the corresponding structure below the containment. A 5-inch vertical misc.lignment of the centerline of the secondary shield wall above and below the containment is evaluated. The resulting stresses are acceptable for the typical reinforcing pattern determined for the crane wall. With a 5-inch misalignment, the axial resultant remains within the kern of the wall and no moment is transferred to the containment dish. The design also considers potential differential basemat settlement and the effect on the Secondary Shield Wall alignment. Additional materials, quality control, and special construction techniques for the concrete internal structures are outlined in Section 3.8.4.6. 3.8.3.7 Testing and In-service Surveillance Requirements Testing and in-service surveillance requirements are outlined in Section 3.8.4.7. 3.8.4 Other Seismic Category I and Seismic Category II Structures 3.8.4.1 Description of the Structures 3.8.4.1.1 Reactor Building The reactor building is composed of the containment shield building, steel containment vessel including the internal structures, and subsphere. The steel containment vessel is described in Section 3.8.2. The internal structures are described in Section 3.8.3. Details of the reactor building are shown in Figures L2-2 through 1.2-10. Structures are Category I except platforms that do not support Seismic Category I equipment and miscellaneous steel. These structures are Seismic Category II. Seismic Category II structures are designed for the SSE using Seismic Category I criteria. The arrangement of the Nucleer Island structures, which includes the reactor building and defines critical dimensions, flood barriers, and fire barriers, is shown in Figure 3.8-5. The containment shield building is a reinforced concrete structure composed of a right cylinder with a hemispherical dome. The containment shield building shares a common foundation base mat with the nuclear system annex. The containment shield building houses the steel containment vessel and safety-related equipment located in the whsphere, and is designed to provide biological shielding as well as external missile protection for the sted containment shell and safety-related equipment. The containment shield building has an inner radius of 105 feet, a cylinder thickness of 4 feet up to elevation 146 +0. Above elevation 146+0 the shield building thickness is 3 feet including the dome area. The height of the containment shield building is approximately 215 feet. The structural outline of the containment shield building is shown in Figures 1.2-2 and 1.2-3. An annular space is provided between the containment vessel and containment shield building above elevation 91+9 for structural separation r Anwowed Desiges Mater $ Desiger of SSC Page 2.8 22
Sv: tem 80+ D: sign ControlDocument l
- i r
J-j' - and access to penetrations for testing and inspection. The shield building and the nuclear annex are connected to form a monolithic structure. t The subsphere is that portion of the reactor building which is below elevation 91 +9 and external to the containment vessel. The subsphere houses auxiliary sa.fety-related equipment. This area below the ! spherical containment allows efficient use of space for. locating safety equipment adjacent to the l
- containment vessel and eliminating excessive piping while allowing maximum access to the containment l for locating penetrations.
l 1 3.8.4.1.2' Nuclear Gysteen Annex ; i The Nuclear System Annex b composed of the control complex, diesel generator areas, main steam valve i house areas, CVCS and maintenance areas, and spent fuel storage area and unit vent. i The Nuclear System Annex is a Seismic Category I structure with exception of the Unit Vent and , miscellaneous steel that does not support Seismic Category I equipment. These structures are Seismic ! 't Category 11 structures. The Nuclear System Annex, with the exception of the Unit Vent, is a reinforced l concrete structure composed of rectangular walls, columns, beams, and fioor slabs. The Unit Vent is l
; composed of ductwork which runs along the top of the shield building and extends approximately 20 feet above the top of the shield building. The Nuclear System Annex shares common walls and foundatien l
- basemat with and is monoIPhically connected to the containment shield building. In addition to the !
! structural components, there are components designed to provide biological shielding and protection against tornado and turbine missiles. Structural components, as well as members serving as shielding components, vary in thickness from approximately one foot to five feet. Details of the Nuclear System 3 t Annex are shown in Figures 1.2-2 through 1.2-10. The arrangement of the Nuclear Island structures, which includes the nuclear systems annex and defines entical dimensions, flood barriers, and fire barriers, is shown in Figure 3.8-5. 3.8.4.1.3- Station Service Water Pump Structure The Station Service Water Pump Structure is classified as a Seismic Category I structure and is not included in the scope of design certification due to its specific site design requirements. The Station
- Service Water Pump Structure interface requirements are described in Section 9.2.1.1.4.
((The building includes a mat type foundation and a reinforced concrete superstructure with rigid walls.))3 '1he service water pump room and its supporting elements will be protected against flooding. The Station Service Water Pump Structure shall be designed to the same criteria (Section 3.8.4.5) as other [ Seismic Category I structures as well as including consideration of slosh effects using the guidelines of Reference 19. The design shall also consider wave action due to the maximum hurricane flood. 3.8.4.1.4' Diesel Fuel Storage Structures Each of the two Diesel Fuel Storage Structures is classified as a Seismic Category I, reinforced concrete I structure containing two bays. Each bay encloses a diesel fuel oil tank, a ws: em, a sump with a sump pump, and necessary piping. The bays are separated from each other and from the adjacent equipment ' 3 Conceptual Design information; see DCD Introduction Section 3.4. 4preved Deem neeenoder- Deew or ssc Pope 2.s-22
Syatem 80+ Design ControlDocument room by two-hour rated fire barriers. The building arrangement for the Diesel Fuel Storage Stmeture is shown on Figure 1.2.-24. An unloading pad is provided to retain any spills during the fuel oil deliveries. The adjacent equipment room is a Seismic Category 11 steel framed stmeture with insulated metal siding and metal deck roof. The equipment room is designed for the SSE using Seismic Category I criteria. The Diesel Fuel Storage Structures are not connected to the Nuclear Island Structures except by underground fuel transfer piping. 3.8.4.1.5 Component Cooling Water Heat Exchanger Structure The Component Cooling Water (CCW) Heat Exchanger Structure arrangement is shown on Figure 1.2 25. The structure is one of two reinforced concrete buildings classified as Seismic Category I which house the CCW heat exchangers. The two structures provide complete physical separation between the two divisions of the CCW and station service water (SSW) systems. Each structure contains two CCW heat exchangers, CCW and SSW piping and associated valves, sumps and sump pumps. A division of the CCW piping enters the structure through a separate Seismic Category I, reinforced concrete pipe tunnel which runs below grade to the nuclear annex. 3.8.4.1.6 Boric Acid Storage Tank The Boric Acid Storage Tank is a Seismic Category I tank located at grade in the yard within a common Seismic Category 11 reinforced concrete dike structure. The dike structure also surrounds the Holdup and Reactor Water Makeup Tanks and is described in Section 3.8.4.1.11. The tank / dike location is shown on Figure 1.2-1, 3.8.4.1.7 Radwaste Building The Radwaste Building general arrangement is shown on Figure 1.2-23, Sheets 1 through 6. The Radwaste Building is a non-safety related, classified as Seismic Category II, reinforced concrete structure located adjacent to the Nuclear Annex. The Radwaste building is separated from the adjacent Nuclear Annex Structures by a minimum gap of six inches. The Radwaste Building houses the Solid and Liquid Waste Management Systems. Foundations and walls that house the liquid and solid waste management systems are designed such that if a safe-shutdown earthquake (SSE) occurs, the maximum liquid inventory expected to be in the building will be contained. ll7he Radwaste Building is designedfor the SSE using Seismic Category I criteria.))I These design requirements will prevent the collapse of the Radwaste Building on the adjacent Nuclear Annex structure. 3.8.4.1.8 Turbine Building The Turbine Building is a non-safety related Seismic Category 11 structure. The majority of the building is supported by a reinforced concrete foundation which is separated from the adjacent Nuclear Annex Structures by a minimum gap of 6 inches. 8 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approved Design Material. Design of SSC (U97) Page 38-24
Sy:: tem 80+ Design Control Document I (~ In the area of the condensers and turbine pedestal the foundation is locally thickened. The outside wall !
\ from grade is a steel framed superstructure with metal siding. The roof is comprised of prefabricated i trusses with built-up roofing on an insulated metal deck.
There are a main crane and an auxiliary crane, supported off steel columns, that traverse the length of the building. ll1he Turbine Building superstructure lateral resisting steelfrarne is designedfor the SSE using Seismic l Category I criteria.))' These design requirements will prevent the collapse of the Turbine Building on the adjacent Nuclear Annex structure. 3.8.4.1.9 Station Service Building / Auxiliary Boiler Structure The Station Service Building (SSB) and the Auxiliary Boilec Structure (ABS) are shown on Figures 1.2-20, Sheets 1 through 5 and Figure 1.2-21, respectively. The buildings are Seismic Category II structures and located adjacent to the Turbine Building. Since their failure could impact the Turbine Building which in turn could impact the Nuclear Annex structure, ((they are designedfor the SSE using Seismic Category I criteria)) to preclude any effects of collapse on such a system. Each building is a steel framed structure with a steel deck roof and non-combustible roofing. The exterior of each building consists of insulated metal siding. Roof and clean floor drainage are discharged to the storm and waste water system. 3.8.4.1.10 Condensate Storage Tank / Dike h b The Condensate Storage Tank is a non-safety related tak located at grade in the yard within a Seismic Category II, reinforced concrete dike sized to hold the entire contents of the tank. The tank location is shown on Figure 1.2-1. The tank is located at a distance from nuclear safety related systems and structures to preclude any effects of collapse on such a system. The dike structure is designed for the SSE using Seismic Category I criteria. 3.8.4.1.11 IIoldup and Reactor Makeup Water Storage Tanks / Dike The Holdup and Reactor Makeup Water Storage Tanks are non-safety related tanks located at grade in the yard within a common Seismic Category 11 reinforced concrete dike structure sized to hold the entire contents of the tanks located within the dike. The location of the tanks is shown on Figure 1.2-1. The tanks are located at a distance from nuclear safety related systems and structures to preclude any effects of collapse on such a system. The dike structure is designed for the SSE using Seismic Category I criteria. 3.8.4.2 Applicable Codes, Standards, and Specifications Seismic Category I structures are designed as described in Appendix 3.8A using the codes and criteria shown in Table 3.8-4. 1 ( 1
3 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction 3.5.
Anroveet Des &n heateriel- Design of SSC (1/97) Page 3.8 25
Sy-tern 80+ Design Contr'/ Document 3.8.4.3 Loads and Loading Combinations The Seismic Category I and II structures are designed to maintain their function for the following loadings:
- Dead Loads The dead loads include all sustained loads during and after construction.
- Operating Loads Operating loads are those live loads or other variable loads associated with the operation of the plant.
e Design Basis Accident Loads The Design Basis Accident Loads are those associated with the pressure increase in the reactor building annulus due to a temperature rise as a result of the energy release inside the containment vessel due to a loss-of-coolant accident or secondary side pipe break inside containment. Additionally, they include any jet impingement load, missile impact load, and pipe reaction loads due to the postulated pipe break.
- Wind Loads The wind load is based upon ANSI /ASME 7-88 (Reference 5) and ASCE Papers 3269 ar.d 4933 (References 6 and 7) as defined in Section 3.3.1. l 4
The normal and tornado wind loads considered in the design of the containment shield building are nonaxisymmetric loads. The wind loads are analyzed by approximating the wind distribution on the containment shield building as defined in ASCE Paper 4933 by a Fourier Series. The wind distribution curves used in the design are given in Section 3.3.1. Individual harmonics are analyzed and combined to produce the force and moment resultants for the total series. l The wind loads on the Seismic Category I structures other than the shield building are analyzed using the methods defined in ASCE Paper 3269. , l e Tornado Loadmgs I l The tornado loadings are described in Section 3.3.2.
- Rain, Snow and Ice Loads The Seismic Category I structures are designed for rain, snow and ice loads. )
- Soil and Water Pressure The Seismic Category I structures are designed for the earth pressure and groundwater pressure defined in Section 2.4. Dynamic earth pressure loads are described in Appendix 3.7.B for the Nuclear Island. l Approvost Deslyn Material Design of SSC (1/9 71 Page 3.8-26
Sy~ tem 80 + Design ControlDocument
- Seismic Loads (j~')
i. See Section 3.7, " Seismic Design," for the seismic inertia loadings.
- Pressure and Temperature Loads The Seismic Category I structures are designed for global effects of pressure, if any, and temperature gradients, in addition to accident temperature gradients as a result of postulated pipe ruptures described in Section 3.6.1. Exterior walls and roofs above plant finished yard grade are designed for the 1 percent exceedance ambient temperature values in Table 2.0-1.
The potential for global temperature and pressure loads in the Nuclear Annex is minimized by the selected routing of high pressure lines as described in Section 3.6.1.1. Nuclear Annex subcompartments susceptible to potential global pressure and temperature and the design basis for these effects are as follows:
- The Main Steam Valve House (MSVH) is designed for a 10 psi differential pressure across the walls as a result of a Man Steam line break. The MSVH temperature is 300*F. The short duration of this temperature results in a negligible thermal load across the MSVH wall.
- The pipe chase containing the Chemical and Volume Control System (CVCS) 2 inch
,-m letdown line is designed for compartmental pressure and temperature resulting from a l
) postulated rupture of this line.
The Shield Building annulus global temperature and pressure loads from postulated events are described in Section 6.2.1.8. The values identified are 9.3 inches of water and 212*F which are applied for an extended period of time. The Shield Building annulus wall pressure design is governed by a tornado event which results in a differential pressure across the Shield Building l l wall of approximately 2.4 psi. Access doors and penetrations, including seals, are designed for the subcompartment pressures when there is a potential to affect safety-related equipment if the door or penetration fails to retain the pressure boundary. Loading combinations used for the design of Seismic Category I structures are shown in Section 5.0 of Appendix 3.8A. 3.8.4.4 Design and Analysis Procedures Seismic Category I and 11 concrete and steel stmetures, with exception of the steel cortainment vessel, are designed in accordance with the criteria in Appendix 3.8A. The Seismic Category I structures are designed to prevent possible overturning, sliding and flotation. The forces and moments acting on the building which could cause these events are determined for the different loads and load combinations and are then compared to the corresponding forces and moments
,a which resist overturning, sliding or flotation. Safety factors for the possible events are determined for
() comparison with the allowable safety factors listed in Section 5.0 of Appendix 3.8A. Approved Design Material. Design of SSC (1/97) Page 3.8-27 m
Syntem 80+ Design controlDocument All Seismic Category I removable walls and hatch plugs are positively anchored to withstand the design pressures and inertia loads. Design and analysis results for selected areas of the Nuclear Annex are presented in Appendix 3.8B. 3.8.4.5 Structural Ac-eptance Criteria 3.8.4.5.1 Concrete Structures ([ Analysis and design of Seismic Category I concrete structures use the ultimate strength design method in accordance with ACI 349 when supplemented by thefollowingpronsions:
- 1. Special consideration is given to anchorage pull-out capacity (i.e. - reduced concrete failure cone) especially when; e the anchor is near thefree edge of the concrete, and/or, e the anchors are closely spaced, and/or e the anchor (s) are placed in the tension zone of the slab.
- 2. Baseplateflexibility is accountedfor when calculating anchor bolt loads.
- 3. Thefailure cone angle used is consistent with recent test datafor the specipe application.
- 4. The embedment length of ductile anchors is chosen such that the ratio of the anchorpull-out 9
capacity (concrete) to the anchor minimum tensile capacity (steel) is greater than or equal to 1.50.
- 5. Erpansion anchor bolts are designed to have thefollowing minimumfactor ofsafety between the bolt design load and the bolt ultimate capacity deternanedfrom static tests.
e Four (4.0)for wedge and sleeve anchor bolts.
- Three (3.0)for undercut anchors.
The ultimate capacity of the anchor bolt accountsfor shear-tension interaction, minimum edge distance, andproper bolt spacing.
- 6. The energy absorption capability (deformatim capability after yield) is consideredfor the anchor material.
- 7. The efects of cyclic loading are considered in the andor bolt design.}}'
ACI 349 is used to determine the required quantity of shear, tension, and compression reinforcing in Seismic Category I structures, in addition ACI 318, Chapter 21 is used to determine the required 2 NCR Staff approval is required prior to implementing a change in this information; see DCD Introduction Setion 3.5. Astroved Design Meterial Design of SSC t1/97) Page 3.8 28
Sy' tem 80+ D sign ControlDocument F] t/ anchorage and splicing of reinforcing to assure ductility at the structural connections. ACI 318, Chapter 21 is also used to determine the configuration of reinforcing in the structural joints, and regions where reinforcing is spliced, and required placement of stirrups and hoop steel. Typical connection details are provided in Appendix 3.8B, Section 6.0. Masonry block walls are not used in Seismic Category I or 11 structures. 3.8.4.5.2 Steel Structures
\\ Seismic Category I structural steel analysis and design are in accordance with ANSI /AISC N690 when supplemented by thefollowing provisions:
- 1. In Section Q1.0. 2, the depnition ofsecondary stress applies to stresses developed by temperature loading only.
- 2. Thefollowing notes are added to Section Q1.3:
"When any load reduces the effects of other loads, the corresponding coeffcientfor that load shall be taken as 0.9, if it can be demonstrated that the load is always present or occurs simultaneously with other loads. Otherwise, the coefficientfor that load shall be taken as zero. "
"Where the structural effects of differential settlement are present, they should be included with the dead load 'D'. "
,m "For structures or structural components subjected to hydrodynamic loads resultingfrom LOCA (N') and/or SRV actuation, the consideration of such loads should be as indicated in the Appendix to the SRP Section 3.8.1. Anyfluid-structure interaction associated with those hydrodynamic loads and thosefrom postulated earthquake (s) should be taken into account. "
- 3. The stress limit coefficients (SLC)for compression in Table Q1.5.7.1 are asfollows:
1.3 instead of 1.5 [ stated in footnote (c)] in load combinations 2, 5, and 6. 1.4 instead of 1.6 in load combinations 7, 8, and 9. 1.6 instead of 1. 7 in load combination 11.
- 4. Thefollowing note is added to Section Q1.5.8:
"For constrained (rotation and/or displacement) members supporting safety related structures, systenu, or components, the stresses under load combinations 9,10, and 11 should be limited to those allowed in Table Q1.5.7.1 as modiped by provision 3 above. Ductilityfactors of Table Q1.5.8.1 (or provision 5 below) should not be used in these cases. "
- 5. For ductilityfactors 'p' in Sections Q1.5. 7.2 and Q1.5.8, are substitutedprovisions ofAppendix A,11.2 of SRP Section 3.5.3 in lieu of Table Q1.5.8.1.
f 6. In load combination 9 of Section Q2.1, the loadfactor applied to load P is 1.5/1.1 = 1.37, (}) instead of1.25. Approved Design Material- Design of SSC t1/97) Page 3.8 29
Syatem 80+ Design Control Document
- 7. Sections Q1.24 and Q1.25.10 is supplemented with thefollowing requirements regardingpainting of structuralsteel:
- Shop painting shall be in accordance uith Section M3 of Reference 17.
- All exposed areas after installation shall befieldpainted (or coated) in accordance with the applicable portion of Section M3 of Reference 17.
e The quality assurance requirementsfor painting (or coating) of structural steel shall be in accordance uith Reference 18 as endorsed by Regulatory Guide 1.54, " Quality Assurance Requirementsfor Protective Coatings Applied to Water Cooled Nuclear Power Plants ".}}2 Welding activities associated with Seismic Category I structural steel components and their connections shall be accomplished in accordance with written procedures and shall meet the requirements of AWS Dl.1 (Reference 25). The visual acceptance criteria shall be as defined in NClG-01 (Reference 24). 3.8.4.5.3 Concrete and Steel Structures In addition to satisfying the load combinations for structural adequacy against the design loadings, the load combinations to ensure safety factors against overturning, sliding, and flotation are checked to ensure overall stability of Seismic Category I structures. The following events are checked as a minimum:
- The overturning about the toe of the foundation supported on soil.
- The foundation sliding on soil.
O
- Floating of the foundation base mat.
- The containment vessel slipping in the lower concrete support dish.
- The containment vessel overturning about the edge of the lower concrete support dish.
- The interior structure concrete slipping inside the containment vessel.
The safety factors which must be satisfied during any of these events are shown in Appendix 3.8A, Section 5.2.4. Safety factors which meet or exceed these criteria have been demonstrated in all analyses. No increase in allowable stresses under service load conditions due to normal or severe load combinations is permitted due to wind loadings as identified in NUREG-0800, NRC Standard Review Plan, Section 3.8.4, Part 11.5. Welding activities associated with the Holdup Volume Tank, In-Containment Refueling Water Storage Tank (IRWST), Emergency Feedwater Tank, Refueling Cavity and Spent Fuel Pool liners shall be accomplished in accordance with the requirements of the American Welding Society (AWS) Structural Welding Code, D1.1 (Reference 25). The welded seams of the liner plates shall be spot radiographed 2 NCR Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Sved Design Matenial- Design of SSC (1/97) Page 3.8-3o
. . --- -- -- - -- - .- . .- - - - - --~
l l . System 80+ Deslan contrat Document l where accessible, liquid penetrant and vacuum box examined after fabrication to ensure the liners do not leak. The acceptance criteria shall meet the acceptance criteria stated in Article NE-5200, Section III, ! Division 1 of the ASME Code. 3.8.4.5.4 Stnactural Analysis Report for Seismic Category I Structures l v i A structural analysis report will be prepared for Seismic Category I structures. This report will document that the structures meet the acceptance criteria specified in Section 3.8 and design changes and identified construction deviations, which could potentially affect the structural capability of the structure, have been l t ! incorporated into the structural analysis, consistent with the methods and procedures of Section 3.8. s . The following records will be reviewed, as applicable: ; 5
- 1. Construction records stating material properties for concrete, reinforcing steel, and structural ,
steel. 1 2.' As-built structure dimensions and arrangements, including spatial separation of buildings-
' 3. As-built load requirements including those for subcompartment global pressure / temperature 'I
- effects and for anchor and pipe whip restraints; and
- 4. Design documents for the structure.
Deviations from the design are acceptable provided the following acceptance criteria are met: f
- 1. An evaluation consistent with the methods and procedures of Sections 3.7 and 3.8 is performed
- . (depending on the extent of t14 deviations, the evaluation may range from the documenting of an engineering judgement to performance of a revised analysis and design), and
- 2. The structural design meets the acceptance criteria specified in Section 3.8, and
- 3. The seismic floor response spectra of the as-built structure does not exceed the design basis floor response spectra by more than 10%.
The structural analysis report will summarize the results of the reviews, evaluations, and corrective actions, as applicable, and conclude that the as-built structure is in accordance with the design. [ 3.8.4.5.5- Structural Analysis Report for non-Seismic Category I Structures j l 1 . A structural analysis report will be prepared for non-Seismic Category I structures designed in accordance j W.n Section 3.8.4.1. This report will document that the structures meet the acceptance criteria specified
- in Section 3.8.4.1 and design changes and identified construction deviations, which could potentially j affect the structural capability of the structure, have been incorporated into the structural analysis. ,
i The following records will be reviewed, as applicable:
- 1. Construction records stating material properties for concrete, reinforcing steel, and structural ;
steel; I
- 2. As-built structure dimensions and arrangements, including spatial separation of buildings;
, hemed oneon arenenw. ceewn er ssc . tin n rose 2.s-21
Sy~ tem 80+ D~ sign ControlDocument
- 3. As-beilt load requirements; and
- 4. Design documents for the structure.
Deviations from the design are acceptable provided the criteria of Section 3.8.4.1 for non-Seismic Category I structures are met. The structural analysis report will summarize the results of the reviews, evaluations, and corrective actions, as applicable, and conclude that the as-built structure is in accordance with the design. 3.8.4.6 Material, Quality Control, and Special Construction Techniques The Category I structures are poured-in-place reinforced concrete structures. The major materials that will be used in the construction are concrete, reinforcing bars and structural steel. A brief description of these materials is given below. 3.8.4.6.1 Material 3.8.4.6.1.1 Concrete The basic ingredients of concrete are cement, fine aggregates, coarse aggregates, and mixing water. Admixtures will be used if needed. Cement will be Type I or Type II conforming to " Standard Specification for Portland Cement," ASTM C150. For special circumstances, other approved cements will be used. Aggregates will conform to " Standard Specification for Concrete Aggregate," ASTM C33. l Water usert in mixing concrete will be clean and free from injurious amounts of oils, acids, alkalis, salts, organic materials or other substances that may be deleterious to concrete or steel. A comparison of the proposed mixing water prope.rties will be made with distilled water by performing the following tests: ; I
- Soundness, in accordance with " Standard Test Method for Autoclave Expansion of Portland Cement," ASTM C151. The results obtained for the proposed mixing water will not exceed those obtained for distilled water by more than ten percent.
- Time of setting, in accordance with " Standard Test Method for Time of Setting of Hydraulic Cement by Vicat Needle," ASTM C191. The results obtained for the proposed mixing water will be within ten minutes for initial setting time and one hour for final setting time of those obtained i for distilled water.
- Compressive strength, in accordance with " Standard Test Method for Compressive Strength of Hydraulic Cement Mortars (using 2 in. cube specimens)," ASTM C109. The results obtained for the proposed mixing water will not be lower by more than five percent of those obtained for distilled water.
The water used to make ice for concrete pours in hot weather will conform to the requirements for miting water described above. Approwd Design Material Design of SSC (1/97) Page 3.8-32
System 80+ Deskn ControlDocument
'[ . Admixtures, if used and as determined by detailed mix design, will conform with the applicable ASTM
( standard: Le Air-entraining admixtures. " Standard Specification for Air-Entraining Admixtures for Concrete," ^ ASTM C260.
- Water reducing, retarding, and accelerating admixtures. " Standard Specification for Chemical Admixtures for Concrete," ASTM C494.
.* Pozzolanic admixtures. " Standard Specification for Fly Ash and Raw or Calcined Natural Pozzolan for use as a Mineral Admixture in Portland Cement Concrete," ASTM C618.
- Slag cement. " Standard Specification for Blended Hydraulic Cements," ASTM C595.
- Plasticizing admixtures. " Standard Specification for Chemical Admixtures for Use in Producing Flowing Concrete," ASTM C1017.
i The combined chloride content of the admixtures and mixing water will not exceed 250 ppm.. . The ingredient materials will be stored in accordance with the detailed recommendations presented in ACI , _304 (Reference 10). , 1 Concrete mixes will be designed in accordance with ACI 301 (Reference 9). The batching, mixing and transporting of concrete will conform to ACI 301. The placement of concrete, consisting of preparation before placing, conveying, depositing, protection and bonding will be in accordance with ACI 301. 3.8.4.6.1.2 Reinforcing Steel Reinforcing steel will consist of deformed reinforcing bars conforming to " Standard Specification for Deformed and Plain Billet - Steel Bars for Concrete Reinforcement," ASTM A615, Grade 60 or
" Specifications for Low-alloy Steel Deformed Bars for Concrete Reinforcing," ASTM A706, Grade 60.
The fabrication of reinforcing bars, including fabrication tolerances, will be in accordance with CRSI
" Manual of Standard Practice" MSP-1. The placing of reinforcing bars, including spacing of bars, '
concrete protection of reinforcement, splicing of bars and field tolerances will be in accordance with ACI . 349. Epoxy coated reinforcing steel is used for areas where a corrosive environment is encountered. 3.8.4.6.1.3 Structural Steel The structural steel will essentially consist of low carbon steel shapes, plates and bars conforming to ,
" Standard Specification for Suuctural Steel," ASTM A36. Other structural steels listed in ANSI /AISC N690 may also be used.
Fabrication and erection of structural steel in Seismic Category I structures will be in accordance with the' requirements of ANSI /AISC N690. The structural connections will be either welded or bolted. 7 Welding activities associated with Seismic Category I structural steel components and their connections shall meet the requirements in Section 3.8.4.5.2. f a L .::Dee4pn a0eamel- Design of SSC M9H Page 3.8-33
. ,~ _
Design C*ntrolDocumnt Sg. 80 + All bolted connections will be made with high strength bolts conforming to one of the following specifications:
* " Specification for High-Strength Bolts for Structural Steel Joints," ASTM A325. * " Specification for Heat Treated Steel Structural Bolts,150 ksi Tensile Strength," ASTM A490.
Other bolts listed in ANSI /AISC N690 may also be used. 3.8.4.6.2 Quality Control The quality of materials will be controlled by requiring the suppliers to furnish appropriate mill test reports as required under relevant ASTM Specifications as described in Subsection 3.8.4.6.1. These mill test reports will be reviewed and approved in accordance with the general provisions of the overall Quality Assurance Program outlined in Chapter 17 and supplemented by the special provisions of the appropriate codes and specifications for design listed in Table 3.8-4. Erection tolerances, in general, will be in accordance with the referenced design code. Where special tolerances that influence the erection of equipment, etc., are required, they will be indicated on the drawings by the Engineer. 3.8.4.6.3 Special Construction Techniques No unique or untried construction techniques are contemplated. Both the cylindrical and the dome portions of the shield building will be consuucted using standard construction techniques. 3.8.4.7 Testing and In-service Surveillance Requirements There will be no testing or in-service surveillance beyond those quality control tests performed during construction, which will be in accordance with ACI 349, ACI 301, ANSI /AISC N690 or ANSI N45.2.5 (Reference 8) as applicable. 3.8.5 Foundations 3.8.5.1 Description of the Foundations The foundations of the Category I structures are reinforced concrete mats. The foundation of the Nuclear Island is approximately 10 feet thick, has a flat bottom and rests on soil or rock. The top of the Nuclear Island basemat is located 40.75 feet i I foot below the finished grade elevation. The minimum foundation mat thi':knesses for the Diesel Generator Fuel Oil structure and Component Cooling Water lleat Exchanger structure are approximately 2 feet and 4 feet, respectively. Site-specific foundation mat construction procedures will be submitted in accordance with SRP 3.8.5. 3.8.5.2 Applicable Codes, Standards, and Specifications Reinforced concrete foundations and supports of Category I structures are designed as described in Appendix 3.8A using the codes and criteria shown in Table 3.8-4. Approved Design Material- Design of SSC (1/97) Page 3.3-34
__ _ . _ _ _ . ~ . . _ ._._._ _ _ . . _ _ _ _ - - -
- ! Syrtem 80+ oestan ControlDocument [
. t 3.8.5.3- Loads and Loading Combinations
- The design loads and loading combinations are described in Section 3.8.4.3 and Appendix 3.8A.
-3.8.5.4 Design and Analysis Procedures i
The reinforced concrete foundations of Category I structures are analyzed and designed for the reactions due to static, seismic and all other significant loads at the base 'of the superstructures supported by the foundation in accordance with the criteria in Appendix 3.8A. The foundation mat is modeled as a three dimensional finite element structure as an integral part of the Nuclear Island finite element model and l accounts for variation of soil properties. This model is described in Appendix 3.8A. The forces and T moments determined in the analysis are input to the structural design. Design and analysis results are ; L presented in Appendix 3.8B.
' The analysis and design of the foundations considers the effects of varying soil properties beneath a specific foundation, the effects of potential mat uplift, and the effects of constructidn sequence, with ;
. particular emphasis on differential settlements of the basemat.
Reinforcing steel is designed for forces and moments determined by analysis. This area of steel is duplicated in the opposite face of the basemat. Thus, the basemat is reinforced symmetrically to address ; the potential for cracking due to differential settlements. ; A settlement monitoring program is required for all Seismic Category I structures. Settlement monuments are provided at appropriate locations to track total and differential settlements. Monitoring is begun as ) each monument is installed. Actual versus predicted settlement is tracked and evaluated for each Seismic i Category I structure. 3.8.5.5 Structural Acceptance Criteria -
- These are outlined in Section 3.8.4.5.
3.8.54 Material, Quality Control, and Special Cuw uction Techniques l Epoxy coated reinforcing steel is used for areas where a corrosive environment is encountered. Other criteria are outlined in Section 3.8.4.6 and Appendix 3.8A. 3.8.5.7 Testing and In-service Surveillmace Requiresnents These are outlined in Section 3.8.4.7. 4 References for Section 3.8 ' o li ASME Boiler and Pressure Vessel Code. U 2.' Gabriel L DeSalvo and John A'. Swanson, ANSYS Fnoineerine Analysis System User's Manual, , Swanson Analysis Systems,Inc. 3i ~ ACI 349,~ " Code Requirements for Nuclear Safety Related Concrete Structures." > [%) L idee 4pn neeeeriet
- Deedon of ssc . - ( 1/ 9 71 Pope 3.8-35
xp Syntem 80+ Design CsntrolDocument
- 4. ANSI /AISC N690, " Nuclear Facilities - Steel Safety-Related Structures for Design Fabrication and Erection."
- 5. " Minimum Design Loads for Buildings and Other Structures," ANSI /ASCE 7.
l 6. " Wind Forces on Structures," ASCE Paper No. 3269, Transactions, ASCE, Vol.126, Part II, 1961, p.1124.
- 7. " Wind Loads on Dome-Cylinder and Dome-Cone Shapes", ASCE Paper No. 4933, Journal of the Structural Division, ASCE, Vol. 92, No. ST5, October 1966, p.79.
- 8. " Supplementary Quality Assurance Requirements for Installation, Inspecuon, and Testing of Structural Concrete, Structural Steel, Soils and Foundations During the Construction Phase of Nuclear Power Plants," ANSI N45.2.5.
- 9. ACI 301, " Specifications for Structural Concrete for Buildings."
- 10. ACI 304, " Recommended Practice for Measuring, Mixing, Transporting and Placing Concrete."
- 11. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants."
- 12. Regulatory Guide 1.57, " Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components."
- 13. Regulatory Guide 1.61, " Damping Valves for Seismic Design of Nuclear Power Plants."
- 14. Regulatory Guide 1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis."
- 15. Code of Federal Regulations, Title 10, Part 50.
- 16. Regulatory Guide 1.84, Design and Fabrication Code Case Acceptability ASME Section ll!
Division 1.
- 17. " Allowable Stress Design," American Institute of Steel Construction, Chicago, Ill.,1989.
- 18. ANSI N101.4, " Quality Assurance for Protective Coatings Applied to Nuclear Facilities,"
American Institute for Chemical Engineers, New York, New York.
- 19. American Society of Civil Engineers, " Seismic Analysis of Safety Related Nuclear Structures and Commentary on Standard for Seismic Analysis of Safety Related Nuclear Structures," Publication No. ASCE 4-86, September 1986.
- 20. Miller, J.D. and Clauss, D.B., " SAND 88-1631C Evaluation of Performance of the Sequoyah Unit 1 Containment Under Conditions of Severe Accident Loading, Founh Workshop on Containment Integrity," NUREG/CP-0095, SAND 88-1836, June 1988, pages 571-588.
- 21. NUREG/CR-4216, SAND 85-0790, Experimental Results for a 1:8 Scale Steel Model Nuclear Power Plant Containment Pressurized to Failure.
Approved Des]per Material Des 19er of SSC (1/97) Page 3.8-36a
t Sy:, tem 80+ Deskn centrolDocument
/ 22. ASME Code Case N 284, Metal Containment Shell Buckling Design Methods Section III,
-' Division 1, Class MC.
- 23. Welding Research Council Bulletin 107, Local Stresses in Spherical and Cylirdrical Shells due to External Loadings, March 1979.
- 24. NCIG-01, " Visual Weld Acceptance Criteria for Structural Welding of Nuclear Power Plants," ;
Revision 2, EPRI NP-5380.
- 25. AWS DI.1, The American Welding Society, Structural Weldingfode 1990.
l 0 4 4 l ! 1 4 (f l Aporenet Dee> Meterial- Dee# of SSC (1/pyj peye ygy
2< . System l'O + Oesion ControlDocument :
! f"N -
1 Q Table 3.8-1 Design Loadings for Steel Containment , 4 D --- Dead loads C - Construction loads l- L? ' Live loads , P, .- - Test pressure. , T, - Test temperature-T.o
- Thermal loads during startup, normal operating or shutdown conditions R. - Pipe reactions during startup, normal operating or shutdown conditions P. - Pressure loads during startup, normal operating or shutdown conditions P, - External pressure loads resulting from pressure variation either inside or outside containment ,
T,: - Thermal effects associated with external pressure loads ] R,. - Pipe reactions associated with external pressure loads ! Ei - Loads generated by the safe shutdown earthquake Pi - Pressure load generated by the postulated pipe break accident l P, - Pressure load generated by an uncontrolled hydrogen-oxygen gas recombination P, - Pressure load associated with the ultimate capacity stress on the containment T. - Thermal loads generated by the postulated pipe break accident n R.
' - Pipe reactions due to thermal conditions generated by the postulated pipe break accident Y, -- Equivalent static load on the structure generated by the reaction on the broken pipe during the design basis accident Yj - Jet impingement equivaient static load on the structure generated by the broken pipe during the design basis accident Y, .- Missile impact equivalent static load on the structure generated by or during the design basis accident, such as pipe whipping Fo - Loads generated by the post LOCA flooding of the containment M:
i Y . 1L
./
, w w aneeenw- neekn er ssc rose 2.s-27
Design controlDocument , System 80+ Table 3.8-2 Loading Combinations for Steel Containment The containment vessel may be subject to the combined effect of two or more of the design loadings listed in Table 3.8-1. The stresses computed by analysis for each design loading shall be combined in the following manner:
- Testing Condition:
This includes the testing condition of containment to verify its leak tight integrity. D + L + P,+ T,
- Design Conditions:
These include all design loadings to which the containment vessel might be subjected during the expected life of the plant D + L + T, + R, + P,
- Service Conditions:
The loads corresponding to the service limits may be combined by their actual time history of occurrence while considering their dynamic effect upon the structure. Level A Service Limits: D + L + To+ Ro + P. D + L + T, + R, + P. Level C Service Limits: D + L + T, + R, + P + E' D + L + To + R, + P + E' Level D Service Limits: D + L + T, + R, + P, + Y, + Yj + Y, + E'
- Construction Conditions:
D+C O Appromt Desse n unterw- Design of SSC Page 3.8-38
System 80+ Design C ntrolDocument ' MN) Table 3.8-2 Loading Combinations for Steel Containment (Cont'd)
- Stability Considerations:
The following combination is considred for stability with the containment subject to external pressure. The peak external pressure is conservatively combined with the seismic load without regard to time phasing of the events. The load combination meets the ; requirements of Paragraph 3222.2(b), Subsection NE, Section III, Division 1, of the ASME Code. The Service Limits are defined in the ASME Code. D + L + T, + % + P, + E' The following combination is considered for stability due to the thermal conditions which l induce compressive stresses at the base of the steel containment vessel in the transition region. The load combination meets the requirements of Paragraph 3222.2(a), Subsection NE, Section III, Division 1 of the ASME Code. D + L + T, + R, + P,
- Ultimate Capacity Conditions D+L+P,
- Combustible Gas Load Conditions D + L + P, + P, Overturning and Sliding Conditions D + L + E' O
^L/
c.= onw uerew- outen or ssc rage s.a 3s
System 80+ Design controlDocument Table 3.8-3A Stress Intensity Results for the Steel Containment Vessel g mmu alculated Reduced Load 8 " '*"SIII Load Categories Combination Equation Type 3) Test Condition D + L + P, +T, P, 22.4 Design Condition D+L+P, P, 20.1 D+L+P, P. 20.1 Level A D + L +T, + P, Pi + P,+ Q See Table 3.8-3B Level D D + L + P,+ E' P. 31.4 l Construction D+L+C P, 12.8 l Ultimate Capacity D+L+P, P, See Table 3.8-3D Combustible Gas D + L + P, + P, P, 48.1 l i l l 9 Approved Design Meterial- Design of SSC page 3.g4g
System 80+ Design ControlDocument 4 ) Table 3.8-3B Service Level A Stress Analysis, Simplified Elastic-Plastic Analysis (Para. NE-3228.3) 1%" Base 2" Base Allowable Pi + P +Q% %,i wing) , 76.8 ksi 74.4 ksi 80.1 ksi (NE-3228.3(a))
- 1. 1. 3 Stress Conc. Fact. (NE-3228.3(b);
Carbon steel: m=3, n=0.2) (S,=102 ksi) (S,=101 ksi) Usage Factor
- 0.53 0.50 1 (NE-3228.3(c))
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LEGEND a . . DooRvAv WEN m VERTICAL NI EAM VALVE HOUSE O am , (\\] FLODD BARRIER l:{ j:{j 3-w FIRE BARRIER 3-HR r!RE AND ILUUO I#IER SHIELD BUILDING CONTAINMENT FOR NOTES SEE FIGURE 3tf-5 SK 12 ; SECONDARY < SHIE D WALL
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- 1. FLOOD DOORS ARE PROVIDED IN FLOOD BARRIERS, AND PENETRATIONS ARE SEALED UP TO THF EXTERNAL AND INTERNAL FLODD LEVELS.
SENSORS APE PROVIDED ON FLODD DOORS V!TH OPEN AND CLOSE ST ATUS INDIC ATIONS AT A H3NITORED LOCATION.
- 2. 3-HOUP FIRE RATED DOORS AND ELECTRICAL AND MECHANICAL PENETRATION SF ALi ARE PF<0VIDED FOR OPENINGS IN THE 3-HOUR FIRE RATED DARRIERS.
- 3. THE FOLLOVING STRUCTURES, SYSTEMS. AND COMPONENTS DEPICTED ON THESE FIGURES ARE NOT SEISMIC CATEGORY 1r DOORWAY GrENINGS VERTICAL ACCESS OPENINGS STA RS ELE VATORS
- 4. THIS DIMENSION IS MEASURED AT THE TOP ELEVATION OF THE LEVEL 4 REINFORCED CONCRETE FLOOR (10'-6*) IN A DIRECTION PARALLEL TO THE RESPECTIVE PLANT ORIENTATION AZIMUTH, 0* - 180* (12*) OR 90* - 2 70* (12*),
BETVEEN THE EXTERIOR SURF ACES OF THE REINFORCED CONCRETE AT THE CORNERS SHOVN S. THIS DIMENSION IS THE DIFFERENCE BETVEEN THE PLANT GRADE EL EVATION AND THE TOP ELEVATION OF THE LEVEL 1 REINFORCED LONCRETE FLOOR AT THE LOCATIONS INDICATED ON FIGURE 3 8-5, SH. 3. THE PLANT GRADE ELEVATION IS DETERMINED AT THE EXTERIOR CORNER OF THE REINFORCED CONCRETE VALL ADJACENT TO THE l LOCATIONS INDICATED ON FIGURE 3.8-5, SH. 3. 1 LEGEND u.- ABPREVIATIONS
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t'. 200'-0' (12'-0') IS THE INSIDE DIAMETER OF THE STEEL s CONTAINHENT SPHERE, THE INSIDE RADIUS Or THE SPHERE IS 100'-O' (* 1'-0'). THE INSIDE RADIUS IS HEASURED AT THE ELEVATION OF THE CENTER OF THE SPHERE (f 0'-6') IN FOUR DIRECTIDNS, PL ANT ORIENTATION A71MUTHS 0* (15*), '30' (1 5*), 180* (15*), 270* (15*). DME ADDITIONAL INSIDE RADIUS IS ME ASURED FROM THE CENTER GF THE SPHERE, VERTICALLY (15*). TO THE TOP Dr THE CONTAINMENT,
- 7. THE MINIHUM DISTANCE IS ME ASURED IN A HORIZONTAL DIRECTICN AT THE ELEVATION OF THE CENTERLINE OF THE CONTROL ROOH AIR INTAKE (A0'-6')
BETVEEN THE IDENTIFIED LOCATlDNS. THE LOCATIONS ARE FURTHER DESCRIBED AS FOLLOVSs
@ AND h THE INTERSECTICN OF TF CENTERLINE OF THE CONTROL RCOM AIR INTAKE VITH i i PLANE OF THE EXTERIOR REINFORCED CONCRETE NUCLEAR ANNEX VALL.
h AND h THE POINT ALONG THE L]NE FORHED BY THE INTERSECTION Or THE REINFORCED CONCRETE SHIELD BUILDING EXTERIDR VALL AND THE EXTERIOR SURFACE OF THE REINICRCED CONCRETE NUCLEAR ANNEX RDOF THAT IS CLOSEST TD THE CONTROL RDOM AIR INTAKE, h AND h - THE INTERSECTION STEAM VALVE OF THEEXTERIOR HOUSE REINFORCED VALL, THE CONCRETE REINFORCED HAIN CONCRETE SHIELD BUILDING EXTERIDR VALL, AND THE EXTERIOR SURFACE OF THE REINFORCED CONCRETE NUCLEAR ANNEX ROOF. f. , h - THE INTERSECTION OF THE CENTERLINE Or THE UNIT VENT VITH THE PLANE PASSING THROUGH THE TOP SURF ACE - # {fp C 0F THE UN1T VENT, g. ' 4 < f jy h AND h THE INTERSECTION OF THE CENTERLINE OF TH~ DIESEL GENERATOR EXHAUST VITH THE PLANE THE TOP SURFACE DF THE DIESEL GENERATOR EXHAUST. 4p #7ttrIt//y,PASSING T o gj; h AND h THE POINT ALONG THE LINE FORMED BY THE INTERSECTION OF THE REINFORCED CONCRETE HAIN STEAM VALVE HOUSE EXTERICR VALL AND THE EXTERIOR SURFACE OF THE REINFORCED CONCRETE NUCLEAR ANNEX ROOF THAT IS CLDSEST TO THE CONTROL LOCH AIR INT AKE. 9705090171 - Nuclear Island Structures; Notes, Legend, and Abbreviatiom Figure 3.8-5 Sheet 12 of 12 Approved Design Matedal- Design of SSC Pege 3.819
System 80+ Deskn Control Document i
..f) Effective Page Listing Appendix 3.8A-Pages Date ,
i,li 1/97 lii - y Original vi . 2/95 vii . Original 3.8A 1 through 3.8A-35 Original 3.8A-36 2/95
' 3.8A-37 through 3.8A-40 Original )
3.8A-41 2/95 I l 3.8A-42, 3.8A-43 Original l
- 3.8A-44 2/95 !
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system 80+ oesten cetrot occummt . Q,n Appendix 3.8A Structural Design Criteria Contents Page 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8A-1 2.0 Definitions and Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-2 2.1 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 3.8A-2 2.2 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8 A-3 i 3.0 Plant Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-5 3.1 Nuclear lsland Category l Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-5 . 3.1.1 %1 ear lsland Foundation Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-5 U.O 3.1.2 Reactor Building .........................................3.8A-5 3.1.3 Nuclear Annex .............. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-6 3.2 Non-Nuclear Island Seismic Category I and II Structures . . . . . . . . . . . . . . . . 3 . 8A-6 i 4.0 - Quality Assurance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A 4.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-6 4.2 Quality Assurance Classifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-6 4.3 Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8 A-7 4.4 Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-7 4.5 Construction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-7 4.5.1 Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-7 i 5.0 Structural Design Loads and Load Combinations . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-7 5.1 Design Loads ...........................................3.8A-8 5.1.1 Normal Loads . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8 A-8 7 5.1.2 Severe Environmental 1 mads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-13 5.1.3 Extreme Environmental Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-14 , 5.1.4 Abnormal Loads ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A- 16 5.2 Design lead Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-16
-5.2.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8A- 16 5.2.2 Loading Combinations for Seismic Category I Concrete Structures .......... 3.8A-17 5.2.3 Loading Combinations for Seismic Category I Steel Strv.ctures . . . . . . . . . . . . 3.8A-18 n 5.2.4 Loading Combinations for Sliding, Overturning, and Elotation . . . . . . . . . . . . 3.8A-19 5.2.5 Construction Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-20
-(") - 5.2.6 Applicability of Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-20 Newed Deefpn Mowiel Deefpn of $$C *~en u
System 80+ Design ControlDocument 1 1 Contents (Cont'd.) 6.0 Structural Analysis and Design, Requirements and Procedures . ........... 3.8A-20 ! 6.1 Analys is . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 3.8A-20 l 6.1.1 General . . . . . . . . . . . . . . . . . . . . . ................. .... .. 3.8A-20 6.1.2 Seismic Analysis .............................. .......... 3.8A-21 6.1.3 Thermal Analysis . . . ..................... ..... ......... 3.8A-22 6.1.4 Other Analyses ........ .................... ............ 3.8A-22 6.2 Structural Design . . . . . . . . . . . . . . . . .. .... ..... .. ...... 3.8A-22 6.2.1 General Requirements . . . ... .. ....... ............ .... 3.8A-22 6.2.2 Special Design Criteria .... ................... ... ....... 3.8A-26 7.0 Construction; Forming, Fabrication, and Erection . . . . . . . .... .. ... 3.8A-28 7.1 Concrete .................................... . .. .... 3.8A-28 7.1.1 Concrete Mix Design .............. ........ . .... ...... 3.8A-28 7.1.2 Concrete Placement . . . . . . . . . . . . . . . . ........ ...... ....... 3.8A-28 7.1.3 Reinforcing . . . . . . . . . . . . .. ........... ..... .......... 3.8A-28 7.1.4 Construction Sequencing ............. . ... ... ......... 3.8A-29 7.2 Structural Steel ............ .... ..... ............... 3.8A-29 7.2.1 Structural Steel: Fabrication and Erection . . . ... ... . . . . . . . . . . . 3.8A-29 7.2.2 High Strength Bolted Connections . . . . . .. . ... .. . .. .... ... 3.8A-29 7.2.3 Welded Connections . . . . . . . . . . . . . . . . . . . . . . . .. .... .. . .. 3.8A-29 8.0 Structural Acceptance Criteria . . . . . . ........ . ...... ..... 3.8A-29 9.0 Materials . . . . . . . . . . . . . . . . . . . ... ....... .. .... .... 3.8A-30 9.1 General . . . . . . . . . . . . . . . . . ...................... ..... 3.8A-30 9.2 Specifications .... .... . . ... ........ .. . ....... 3.8A-30 9.2.1 Concrete ...................... .... ... .......... . 3.8A-30 9.2.2 Steel . . . . . .... ..... .... .. .. ...... ....... . .. 3.8A-31 9.3 Restricted Material . . . . . . .... ..... ..... .. .. ... . 3.8A-32 10.0 Supplemental Design Criteria for Nuclear Island, Category I and II Structures . . 3.8A-32 10.1 Structural Foundation /Basemat . .. ...... ....... .... ..... .. 3.8A-32 10.1.1 Description . . ..... . ... ...... .. ..... ..... ..... .. 3.8A-32 10.1.2 Design Requirements . . . . ............ .. . ..... 3.8A-32 10.1.3 Design Loads ........ . ................. ... ...... 3.8A-33 10.2 Containment Shield Building ........... ..... . ... ... ... 3.8A-33 10.2.1 Description ............ ...... ... ..... ... . .... 3.8A-33 10.2.2 Design Requirements ...... ........... .... . ......... . 3.8A-33 10.2.3 Design Loads ....... ...... . ... . .. .. . .. ..... 3.8A-34 10.3 Reactor Building Subsphere . . . . . . ................. .. .. 3.8A-34 10.3.1 Containment Support Pedestal . . . . . . . .... ...... . ... . .. 3.8A-34 10.4 Containment Internal Structures . . . . ... ..... . ... .... . 3.8A-35 10.4.1 Reactor Vessel Primary Shield Wall . . .... . ..... . .... 3.8A-35 10.4.2 Crane Wall . . ..... ... . .. . .... .. . . ... 3.8A-35 10.4.3 Refueling Cavity .... ............. ... . ... ...... .. . 3.8A-36 10.4.4 Operating Floor . . . . . . . .. .. ... .... .. ... .... ...... 3.8A-36 10.4.5 In-containment Refueling Water Storage Tank . . . .. ........... 3.8A-37 10.4.6 Lower Concrete Dish ... ........ .......... ......... 3.8A-37 Altwowd Design hinterial . Deslyn of SSC Pageiv
~__ __ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _
l L Sv, tem C0+ oestan contmioocument : e
, Caata=*e (Cont'd.) l . i 10.5 Nuclear Annex .........................................3.8A-37 .
t 10.5.1: General Design Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-38 ; 10.5.2 Diesel Generator Areas . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A +
; 10.5.3 Control Complex . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-3 8 .
10.5.4 Mnn Steam Valve House . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . 3.8A-39 l 10.5.5 ' Fuel Handling Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-39 l
- i. 10.5.6 EFW Tank Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8 A-40 i
t 10.5.7 CVCS and Maintenance Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-40 j i 11.0 Supplemental Design Criteria for Non-nuclear Island,' Seismic Category 1 and II j Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8 A-40 11.1 Diesel Fuel Storage Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-40 _ ! I 11.1.1 Building Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-40
; 11.1.2 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8A-41 !
1 11.1.3 Elevations .............................................3.8A-41 11.1.4 Codes and Standards . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-41 11.1.5 Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8 A-42 11.1.6 Loading Combinations and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 3. 8 A-42 i- 11.1.7 - Other Requirements ......................................3.8A-42
- 11.2 Component Cooling Water Heat Exchanger Structure . . . . . . . . . . . . . . . . . . 3.8A-43 .
11.2.1 Building Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-43 3 11.2.2 ' Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8A-43 ! 11.2.3 Elevations ............................................3.8A ! 11.2.4 Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-44 ! i ~
-11.2.5 Loads ................................................ 3.8A-44 .
_11.2.6 Loading Combinations and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 3.8A-45 : 11.2.7 Other Requirements ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-4 5 11.3 Radwaste Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-45 11.3.1 Building Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-45 , - 11.3.2 Desc ription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-45 ;
- 11.3.3 Elevat ions ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-46 ,
i 11.3.4 Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-46 1 11.3.5 Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8A-46 .! 4 11.3.6 Loading Combinations and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 3. 8 A-47 ) l 11.3.7 Other Requirements ......................................3.8A-47 j e 11.4 Service Water Pumphouse and Intake Structure . . . . . . . . . . . . . . . . . . . . . 3. 8 A-47 ; 11.4.1 Building Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-47
- - 11.4.2 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8A-47 11.4.3 Elevations .............................................3.8A-48
'11.4.4 Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8 A-48 11.4.5 Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8A-48 11.4.6 Loading Combinations and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . 3. 8A-49 11.4.7 Other Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 A-4 9 11.5 Turbine BuiMing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-49
.11.5.1 Building Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-49 11.5.2- Descrip tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 3. 8A-49 11.5.3 Elevatims ......................................... ... 3.8A-50 11.5.4 Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-50 Anvens onow anonener- onew or ssc emne v i
Sr tem 80 + Design ControlDocument 1 1 Contents (Cont'd.) 11.5.5 Loads ................................ ............... 3.8A-50 11.5.6 Loading Combinations and Acceptance Criteria ......... ... ........ 3.8A-51 l 11.5.7 Other Requirements ....................................... 3.8A-52 l 11.6 Dike for outdoor Tanks . . .. ... .... . .. . . . ......... ...... 3.8A-52 11.6.1 Building Classification . . . . . . . . . . . . . . .... .............. ... 3.8A-52 11.6.2 Description . . . . . . . ............. ..................... 3.8A-52 11.6.4 Codes and Standards . . . .......... ....... . .. ..... . ... 3.8A-52 11.6.5 Loads ............................. ................. 3.8A-52 11.6.6 Loading Combinations and Acceptance Criteria ........... ... .... . 3.8A-53 11.6.7 Other Requirements .. .. ................................. 3.8A-53 11.7 Component Cooling Water Tunnel ........... ... .... ......... 3.8A-53 11.7.1 Building Classification . ..... ...... .. . ............. ... 3.8A-53 11.7.2 Description . . . . . . . . . . . . . . ..... ..... ... ....... .. .. 3,8A-53 11.7.3 Elevations ..... ........ ... ... ........... . .... . 3.8A-54 11.7.4 Codes and Standards . . . . . . .... .... ...... . .... . . ?,.8A-54 11.7.5 Loads .. ..... ............ ........ ...... ... . . 3.8A-54 11.7.6 Loading Combinations and Acceptance Criteria .. .. . . . . . . . . . . . . . 3. 8 A-5 5 11.7.7 Other Requirements ........... ..... . ... . .. . ...... 3.8A-55 11.8 Buried Cable Tunnels, and Conduit Banks . . . . . . ... .. . .... . 3.8A-55 11.8.1 Conduit Classification . . . . . . . . . ..... . .. ...... .......... 3.8A-55 11.8.2 Description . . . . . . . . . . . . . . . . . ........ .... ... .. . .. 3.8A-55 11.8.3 Codes and Standards . . ....... ... ... ........ ..... 3.8A-55 11.8.4 Loads ....... .... .... ....... .. .. . . .. . 3.8A-55 11.8.5 Loading Combinations and Acceptance Criteria .. . ......... .... 3.8A-56 11.8.6 Other Requirements ............. ........ . ... . .... 3.8A-56 Tables 3.8A-1 Design Loads for Nuclear Island Category I Structures . .. .. . ..... 3.8A-58 3.8A-2 Diesel Fuel Storage Structure, SSE Accelerations in Gs . . . . ..... .. .. 3.8A-63 3.8A-3 Component Cooling Water Heat Exchanger Structure, SSE Accelerations in Gs . 3.8A-63 3.8A 4 Dead Weight Loads for Major Radwaste Facility Equipment . .... ..... 3.8A-63 3.8A-5 Live Loads for Radwaste Facility . . . . . . . . .... . ... .. . . 3.8A-64 3.8A-6 Radwaste Facility, SSE Accelerations in Gs . ... . .. . . .... 3.8A-65 3.8A-7 Dead Weight Loads for Major Turbine Building Equipment ....... .. 3.8A-65 3.8A-8 Live Loads for Turbine Building . ..... . .... . . . . . .... 3.8A-66 3.8A-9 Turbine Building, SSE Accelerations in Gs . . .... . . .. 3.8A-67 3.8A-10 Dikes for Outdoor Tanks, SSE Accelerations in Gs . ......... .. . 3.8A-67 3.8A-11 Component Cooling Water Tunnel, SSE Accelerations in Gs . .. .. .. 3.8A-67 0 Asproved Design Material Design of SSC (2/95) Page vi
, t Sv' tem 80+ Design ControlDocument .i i 1- c O "'"- 3.8A-1 Nuclear Island Structures - Plan at Level 2 . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-68
- 3.8 A-2 Schematic Representation of 3-D Finite Element Model of the Nuclear Island and the Nuclear Annex Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-70 3.8A-3 Envelope ZPA Profile from SSI Analyses (All Motions, All Soil Cases, All Sticks) .
for the N-S Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-71 i 3.8A-4 Envelope ZPA Profile from SSI Analyses (All Motions, All Soil Cases, All Sticks) ! 4
+
for the E-W Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-72 ; 3.8A-5 Envelope ZPA Profile from SSI Analyses (All Motions, All Soil Cases, All Sticks) for the Vertical Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8A-73 l t w I 1' , 4 4 4 ( i I I
- . l i
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System 80+ Design ControlDocument (]v! 1.0 Introduction This appendix provides the criteria for the analysis and design of structures that comprise the System 80 + Standard Plant. The information presented in this appendix shall be used in the analysis and design of Seismic Category I and Il structural components comprising the System 80+ Standard Plant structures included in the certified design. Design requirements for individual structures are based upon their seismic category and safety classifications listed in Table 3.2-1. The criteria for the Steel Contairanent Vessel are provided in Section 3.8.2 and are excluded from this appendix. The dike surrounding the Station Service Water Pond is site specific and is not addressed within this appendix. All structures required to shut down and maintain the reactor in a safe and orderly condition or prevent the uncontrolled release of excessive amounts of radioactivity following a Safe Shutdown Earthquake have a classification of Seismic Category I. These structures shall be designed to withstand, without loss of function, the most severe postulated plant accident or natural phenomena for the site. Safety classifications are defined in Section 3.2.2. Structural components required as part of the primary containment pressure boundary or for its support and under the scope of the ASME Boiler and Pressure Vessel Code are Safety Class 2. All other structural components required to perform safety related ; functions are Safety Class 3. Sai.ty Class 1 applies to the NSSS primary system components. Safety Classes 1 & 2 are not applicable within this appendix. O ' Those non-Seismic Category I structures capable of impairing the functioning of any Seismic Category I structures or component in the event of failure are classified as Seismic Category II. Seismic Category I 11 structures are designed to prevent failure in the direction of a Seismic Category I structure or component under extreme environmental or accident conditions. The seismic design requirements for i Category II structures under these conditions is equivalent to that of Seismic Category I structures. j Seismic Category I and II, Non-Nuclear Island structures covered by this appendix include the Turbine Building, Diesel Fuel Storage Structure, Component Cooling Water (CCW) Heat Exchanger Structure, CCW Pipe Tunnel, Radwaste Facility, Service Water Pumphouse & Intake Structure and buried cable tunnels and conduit banks. Also included is the concrete dike surrounding the outside CVCS Boric Acid Storage Tank, Ifoldup Tank, and Reactor Makeup Tank. Primary structural components consist of concrete floors, roof slabs, foundation basemats, walls, beams, and columns. Steel beams and columns will be included within this appendix if their primary function is to provide support to walls, floors, or roof slabs. Component support building structures will meet the code requirements of this appendix. Specific load and functional requirements are addressed in , Section 3.9.3.4 and Appenaix 3.9A and under specific design criteria / specifications. Information presented in this appendix is sufficiently comprehensive in nature to:
- Provide the criteria necessary to perform an analysis and translate that analysis into a final design, and n
- Provide a correlation of analysis, design, and constructian requirements with those in Sections
() 3.8.3, 3.8.4, and 3.8.5. , l l l 55nd Design Atatorial- Design of SSC Page3.3A.1
System 80+ Design ControlDocument Miscellaneous components, while not primary structural components, must be considered in the design of primary components as to their loads and method of attachment. Design of these components is based i upon the allowable loads and design requirements found in the ACI, ANSI, ASME and/or other i specialized codes. i Design parameters or information indicated "(by COL)" are delegated to the Combined Operating Liceec 1 Applicant for completion as part of the site specific final design. Design parameters in Table 3xA-1 marked "(TBD)" are also delegated to the COL applicant for completion. See also Section 3.8. 2.0 Definitions and Abbreviations 2.1 Definitions Combined Operating License Combined Constmetion Permit and Operating License with conditions for a nuclear power facility issued in accordance 10CFR part 52 Subpart C. Design Engineer For this criteria, the person given responsibility by the Plant Designer to provide final approval for any structural design activity. Exceedance Value A value for a design parameter based upon a selected probability that the identified value will not be exceeded. Plant Designer A team of Architect Engineers and NSSS vendors who have the responsibility to develop and complete the System 80+ Standard Plant design. Quality Class QA program classifications as identified by ABB-CE and included in Table 3.2-1. Safety related Category I & Il stmetures will be Quality Class 1. Safety Class Relative importance of fluid system components and related equipment as classified in ANSI ANS 51.1 (reference Section 3.2.2) Safety Classes 1, 2, 3, and NNS. Seismic Category Classification of structores (Category I, II, or NS) with respect to requirement to withstand effects of SSE without loss of functional requirements. (Reg Guide 1.29). Zero Period Acceleration The ZPA is the response spectrum acceleration associated with the
" rigid" range frequencies of the response spectrum for a structure or component. The ZPA is the acceleration value corresponding to the peak acceleration of the seismic input upon which the response spectra is based. The " rigid" or ZPA frequency is the lowest frequency at which the acceleration due to the structural response approaches and is approximately equal to 1.0 times the ZPA.
O Approved Design Materie!- Des}gn of SSC Page 3.8A 2
Srtem 80+ Design ControlDocument 2.2 Abbreviations {m} AASHO American Association of State Highway Officials ABB-CE Asea Brown Boveri-Combustion Engineering ABS Auxiliary Boiler Structure ACI American Concrete Institute ADB Administration Building AICE American Institute of Chemical Engineers AISC American Institute of Steel Constmetion ALWR Advanced Light Water Reactor ANSI American National Standards Institute ASCE American Society of Civil Engineers ASME (BPVC) American Society of Mechanical Engineers (Boiler & Pressure Vessel Code) ASTM American Society for Testing and Materials AWS American Welding Society BGSA Bulk Gas Storage Area CC Control Complex , CCW Component Cooling Water CESSAR-DC Combustion Engineering Standard Safety Analysis heport-Design Certification CFR Code of Federal Regulations COL Combined Construction & Operating License (10CFR 52 Subpart C) CPA Condensate Polishing Area CS Containment Spray System O CT Cooling Tower b CTF CTFS Combustion Turbine Facility Combustion Turbine Fuel Storage CVCS Chemical Volume Control System CWPS Circulation Water Pump Storage DBA Design Basis Accident DF Diesel Fuel DGA Diesel Generator Area DS Discharge Structure EFW Emergency Feedwater EPRI Electric Power Research Institute FHA Fuel Handling Area FPH Fire Pump House FSAR Final Safety Analysis Report GDC General Design Criteria / Criterion HIC High Integrity Container HVAC Heating Ventilation and Air Conditioning HVT Holdup Volume Tank I&C(s) Instrumentation & Control (s) ICI In-core Instrumentation IRWST In-Containment Refueling Water Storage Tank IS Intake Structure ITAAC Inspection , Tests, Analyses and Acceptance Criteria
,3 1.BB Leak-Before-Break
( ) MB CVCS & Maintenance Area MS Main Steam Valve House MX Miscellaneous Buildings Approved Design Motorial- Design of SSC Page 3.8A-3
System 80+ Design ControlDocument NA Nuclear Annex NFPA National Fire Protection Association NI Nuclear Island NRC Nuclear Regulatery Commission Nuclear Steam Supply System ! NSSS NUREG NRC Technical Report Designation PAP Personnel Access Portal PMF Probable Maximum Flood PMP Probable Maximum Precipitation PRT Pressurizer Relief Tank PRZ Pressurizer QA Quality Assurance , RA Reactor Shield Building Annulus RB/RXB Reactor Building RC Reactor Building Steel Containment Vessel RCP Reactor Coolant Pump RDT Reactor Drain Tank RFAI Relay House RS Reactor Building Subsphere RW Radwaste Facility SAR Safety Analysis Report SB Station Service Building SCS Shutdown Cooling System SD Station Service Water Discharge Structure SER Safety Evaluation Report (NUREG-1462) SF Spent Fuel Storage Area SG Switch Gear Building SI Station Service Water Pump Structure SIS Safety injection System SP Station Service Water Pump Structure SR Station Service Water Reservoir SRP Standard Review Plan (NUREG-0800) SRSS Square Root of the Sum of the Squares SSC Structure, System, and Component SSE Safety Shutdown Earthquake SSI Soil Structure Interaction ST Sewage Treatment Facilities SY Switchyard TB Turbine Building TBD To Be Determined UBC Uniform Building Code VA Vehicle Access Portal WH Warehouse WT Water Treatment Area XY Transformer Yard YA Yard, above ground ZPA Zero Period Acceleration 10CFR50 Chapter 10, Code of Federal Regulations, Part 50 Anorowd Desiger Meterial. Desips of SSC Page 3.8A-4
System 80+ Design controlDocument ( 3.0 Plant Description 3.1 Nuclear Island Category 1 Structures The term " Nuclear Island" refers to the Basemat, the Reactor Building, and the surrounding Nuclear Annex. The Reactor Building consists of the Containment Shield Building, the Reactor Building Subsphere, Containment Vessel, and Containment Internal Structures. The Nuclear Annex is comprised of all other structures on the Nuclear Island basemat and surrounding the Containment Shield Building. The Nuclear Island, except for electrical and mechanical system tie-ins, is structurally isolated from adjacent structures. Refer to Section 1.2 for details of these structures. 3.1.1 Nuclear Island Foundation Basemat The Nuclear Island foundation is a 10 foot thick reinforced concrete basemat (see Section 10.1 of this 1 appendix) that provides a common foundation for all of the Nuclear Island structures. The top elevation of the basemat is at El. 50'+0". The basemat provides a barrier apinst release of plant fluids to the soil underlying the basemat. Drr. ins are located at the top of the baseme and piped to strategically located sumps. Recessed pump pits are provided in the Containment Subspare area for the CS, SCS, and SIS pumps that are connected to adjacent sumps. 3.1.2 Reactor Building The Reactor Building consists of the Containment Shield Building, Reactor Building Subsphere, l Containment Internal Structures, and the Containment Vessel. 3.1.2.1 Containment Shield Building The Containment Shield Building (see Section 10.2 of this appendix) is the concrete structure that l surrounds the steel Containment Vessel and Reactor Building Subsphere and provides protection from j postulated external missiles and other environmental effects. The Containment Shield Building provides { an additional barrier against the release of fission products. l The Shield Building has a 105' inside radius,4 feet thick, cylindrical reinforced concrete shell extending from the foundation basemat at El. 50'-0" to El.146'-0" The cylindrical wall extends upward from El. l 146' with a 3 ft thickness to the spring line at El.157'-0". The Shield Building is topped by a 3 feet thick reinforced concrete hemispherical roof. The outside apex of the dome is at elevation 265'-0". l 3.1.2.2 Reactor Building Subsphere The Reactor Building Subsphere (see Section 10.3 of this appendix) is located inside the Shield Building and external :o the Containment Vessel. The Subsphere consists of reinforced concrete walls and slabs , and the Containment Support Pedestal. The purpose of the subsphere structures is to support the l containment vessel and the Internal Structures and isolate safety related equipment.
,m i I L)
Approved Desigre Material. Design of SSC Page.3.BA.S i l
System 80+ Design ControlDocument 3.1.2.3 Containment Internal Structures The Containment Internal Structures (see Section 10.4 of this appendix) are located inside the spherical steel containment vessel. The purpose of these internal structures is to provide structural support, radiation and missile shielding, and space for the IRWST. These structures are constructed of reinforced concrete and structural steel. These structures are described in Section 3.8.3.1. 3.1.3 Nuclear Annex The Nuclear Annex (see Section 10.5 of this appendix) is a multi-level reinforced concrete structure surrounding the Reactor Building. The Nuclear Annex is integral with the Containment Shield Building and provides lateral bracing while providing partial tornado wind and missile protection. The Nuclear Annex provides protected areas (Control Complex, Diesel Generator Area, Fuel Handling Area, CVCS Area, and Main Steam Valve House) for safety related equipment. Structural components provide biological shielding required as a result of handling nuclear fuel or processing radioactive wastes. 3.2 Non-Nuclear Island Seistnic Category I and II Structures Refer to Section 11.0 for detailed descriptions of the following:
- Diesel Fuel Storage Structure - Category I,
- Component Cooling Water Heat Exchanger Structure - Category I,
- Radwaste Facility - Category II,
- Service Water Pumphouse ana Intake Structure - Category I,
- Turbine Building - Category II,
- Dike for Outdoor Tanks - Category II,
- Component Cooling Water Tunnel - Category I,
- Buried Cable Tunnels, and Conduit Banks - Category I.
4.0 Quality Assurance Requirements 4.1 General The requirements for a QA program are established in 10CFR50, Appendix B. These requirements are identified in Reg Guides 1.28, and 1.94 by reference to ANSI Standard N45.2.5. The ANSI Standard has been replaced by the ASME QA Program NQA-1. The QA Program is based upon three quality designations as identified in Table 3.2-1 and the two seismic classifications, Category I and II, identified in Section 3.2. The QA Program fulfills the requirements of ACI-349 Section 1.5. 4.2 Quality Assurance Classifications The following Quality Class (QC) designations are applicable to the QA program;
- QC-1 is the highest level quality class and embodies all necessary controls for items and/or services which are required to meet 10 CFR 50 Appendix requirements.
- QC-2 is an intermediate level quality class which is used for items or services which require a moderate level of control of activities affecting quality, but which are neither Nuclear Safety-Related nor required to meet the requirements of 10 CFR 50 Appendix B. Circumstances Anwowd Design Materini . Design of SSC Page 3.BA-6
Sy tem 80+ Design C ntrolDocument , ( appropriate for QC-2 designation include non-standard complex items, or those which must perform reliably, in a harsh environment or with less than normal operator attention or maintenance.
- QC-3 is the quality class which applies to all items or services which are not assigned to another quality class. Quality requirements may be specified in quality plans, procurement ;
documents and/or special procedures if deemed necessary. 4.3 Documentation All structures and components addressed by this appendix are subject to the documentation requirements defined in an approved Quality Assurance Program. 4.4 Materials QA requirements for materials assure that those materials, specified, received, and used meet the requirements in Section 9.0 of this appendix or the applicable design document. The quality of materials is assured by requiring suppliers to furnish certification as required by applicable codes or specifications. : i For fabricated materials, design / procurement specifications shall include acceptance criteria that assure, with the proper QA inspections, materials received match the requirements considered in the design , qualification as well as those shown on design drawings. 4.5 Construction l
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4.5.1 Inspections 4 Quality Control inspections for Seismic Category I structures are addressed in tM ACI 349 concrete and the ANSI N690 steel codes. Procedures shall be prepared to assure that the in3pections are conducted. See also Section 3.8. The Plant Designer shall be responsible for having procedures prepared to assure i that the requirements of ACI 349, Section 1.3 are met. Inspections procedures should include the more detailed provisions of ACI 318 Section 1.3. Inspections of stmetural steel shall be conducted to assure compliance with Sections Ql.23 thru Ql.27 of ANSI N690 and AWS Dl.1 Chapters 3 & 4. ! 5.0 Structural Design Loads and Load Combinations ; Design loads on Category I structural components for System 80+ are identified in this section. The l loads used for the System 80+ Standard Plant envelop expected loads over a broad range of site j conditions. These loads are separated into four (4) categories: normal loads, severe environmental loads, ' extreme environmental loads, and abnormal loads. For each site location, specific loads must be shown i to lie within the standard envelope or additional analyses must be performed to verify structural adequacy. j l The loads identified below are applicable to all structures. The specific loads for which each structure, or part thereof, should be designed shall depend on the conditions to which that particular structure could l be subjected. v[s).
- wmr onew uneww. onion or ssc page s.a u !
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System 80+ Design ControlDocument 5.1 Design Loads General design loads applicable to all structures are identified below. Design loads applicable to individual Seismic Category I and II structures are identified in sections 10.0 and 11.0 of this appendix. Design loads may be either local or global in application. 5.1.1 Normal Loads Normal loads are those loads encountered during normal plant operation and shutdown. They include: Dead loads (D), Live loads (L), Hydrostatic fluid pressure loads (F), Soil pressure loads (H), Thermal loads (T ), and Pipe reactions (Ro). 5.1.1.1 D - Dead Load Dead load refers to loads which are constant in magnitude and point of application. Dead loads are the mass of the structure plus any permanent equipment loads. Equipment designated as a permanent dead load need not be physically attached provided its size and location are expected to remain constant. "D" may also refer to the internal forces and moments due to dead loads. The effects of differential settlement shall be considered with dead loads. Hydrostatic loads from constant fluid levels shall be considered with dead loads. Uniform dead loads represent the structural mass, miscellaneous equipment, and distribution system (electrical cable trays and mechanical piping or HVAC) loads. Specific loads for designated equipment are represented by concentrated loads at the point of application. 5.1.1.2 L - Live Loads Live load, also referred to as operating load, refers to any normal load that may vary with intensity and/or location of occurrence. Variable loads include movable equipment or equipment that is likely to be moved. "L" may also refer to the internal forces and moments due to live loads. Live loads are applied to the structure as either concentrated or uniform loads. For equipment supports, live loads should also consider contributory loads due to the effects of vibration and any support movement. Design drawings should show allowable loads for the designated laydown areas. See also Section 3.8. 5.1.1.2.1 Precipitation The minimum design live load due to precipitation (rain, snow, or ice) for Seismic Category I buildings shall be taken as 50 psf. This live load, equivalent to approximately 9%" of water, will be sufficient for the design peak rainfall of 19.4 in/hr or 6.2 in/5 min given in Table 2.0-1. The design load for rain shall also include the additionalload that may result from ponding due to the deflection of the supporting roof or the blockage of the primary roof drains. 5.1.1.2.2 Compartmental Pressure Loads Compartments shall be evaluated for the potential for internal pressurization. Pressure loads associated with tornadoes. LOCAs, or other explosive type loads shall be classified as extreme environmental or abnormal loads. See Sections 5.1.3.2.1 and 5.1.4.1. Approveet Design Material . Design of SSC Page 3.8A.8
l ! l System 80+ Design ControlDocument 1 (v ) 5.1.1.2.3 Truck Loads Loads due to vehicular traffic in designated truck bays is in accordance with standard AASHO truck loading or identified special loads. Special loads may consist of construction or maintenance loads or routine shipments of fuel casks or other high level radioactive waste. 5.1.1.2.4 Rail Loads Design of the rail / truck bays is controlled by anticipated shipping weights. 5.1.1.2.5 Cranes, Elevators, and Other Hohts This criteria is applicable to structural members and components required for the support of permanently installed cranes and hoists required for station operation and maintenance as well as structural members and components required for the support of temporary construction cranes and hoists. The structural design shall consider the placement of construction hoists on floors, walls, and columns. Design loads shall include the full rated capacity of the hoists plus impact loads as well as test load requirements. Test loads shall be evaluated as 125% of the crane rated capacity. The test loads shall be increased by an additional 25% to account for impact. Test loads shall be checked in Service Load combinations with a factor of 1.1 applied instead of the 1.7 factor normally applied to live loads. The factor is reduced because the test loads are known and the tests are performed under controlled conditions. i p) ( For construction cranes located adjacent to the structure, the structural design shall include soil surcharge loads produced by the full load of the crane. Cranes permanently mounted to structures shall be identified on general arrangement drawings. Pendant operated traveling cranes and trolley hoists shall be designed for 110% of the rated load capacity, to account for impact as required by ANSI N690 Section Q1.3.2. Design loads for motor operated trolleys and cab operated traveling cranes shall be increased by 25 % of the rated load capacity to account for impact in Service and Factored load combinations. Minimum lateral design loads on crane runways shall be 20% of the sum of the rated hoist capacity plus the weight of the crane trolley to account for the effects of the moving trolley. Load shall be applied at l the top of the rail in either direction and distributed according to the relative stiffnesses of the end ! supports. Minimum longitudinal load on each crane rail shall be 10% of the maximum crane wheel loads. Elevators live loads shall be increased by 100% for design of supports. 5.1.1.2.6 Load Allowances for Cable Trays Loads to be applied in areas where multiple cable tray runs are identified include: J
- 7 kips at mid-span on steel beams and columns.
/'%
( '}
- 7 kips at a spacing of 8 ft on center for slabs.
v Approved Design Material Design of SSC Page 3.8A 9 1
Srtem 80+ Design ControlDocument Acceptability of these design loads will be determined through review of the final electrical layout drawings. See also Section 3.8-5.1.1.2.7 Miscellaneous Equipment and Large Bore Piping The following load allowances shall be considered where multiple large bore piping runs are located or where large temporary loads are identified.
- In addition to major equipment located on general arrangement drawings, a point load of 20 kips g
should be applied at the midpoint of each concrete Q : floor slab and concrete beams (Case A). 3ag cose c
- A point load of 40 kips shall be applied at the $ 20K case D2 midpoint of steel collector beams providing primary 3og ces, c yg framing (Case B). m ox case s 30K cose D1
- A point load of 30 kips shall be applied to the midpoint of other steel collector beams or beams 7 provided for support framing (Case C). @[ 0 30kcesec 1
- A point load of 30 kips at midspan on prW steel filler beams framing into steel collector beams (Case DI) and 20 kips on other steel filler beams or stringers (Case D2). (Note: These loads are for added design margin on the beams and slabs and are not to be carried beyond the beam support connection to the supporting beam or column.)
- A contingency load of 80 kips should be applied to the top of each steel column.
Acceptability of these design loads will be verified through review of the final plant configuration. 5.1.1.2.8 Miscellaneous Equipment, Small Bore Piping, Cable Tray, and HVAC Ductwork The following load allowances should be included for areas with multiple runs of small bore piping, cable tray, or HVAC ducts.
- A load of 15 kips on steel collector beams
- A load of 5 kips on other steel beams i l
I
- A load of 50 kips on steel columns Acceptability of these design loads will be verified through review of the final plant configuration. !
l l 5.1.1.2.9 Alternate Load Allowances for Piping, Cable Trays, Conduit, HVAC Ductwork and Miscellaneous Equipment The following alternate load criteria may be used in lieu of Sections 5.1.1.2.6,5.1.1.2.7 and 5.1.1.2.8. , 'Appronef Desips Notedet Desigre of SSC Page 3.8A-10 I
System 80+ Design ControlDocument
.a
(' ) - For piping, cable trays, conduits and HVAC ducts, conservative estimated loads shall be used
~
with a minimum value of 50 psf.
- For major equipment, actual loads shall be used.
- In addition to the above loads, 5 kips concentrated load on beams, girders and slabs shall be used to maximize moment and shear. This load is not carried beyond the beam support connection to the supporting beam, girder or column.
Actual loads shall be tracked during the design process, reconciled with the load allowances established and documented in the structural analysis report described in the structural acceptance criteria, Section 3.8.4.5.4. 5.1.1.3 II - Soil Load Lateral soil pressure shall be br ed upon the soil density and shall include the effects of ground water in accordance .vith Section 5.1.'. 4 of this appendix. Normal soil loads shall consider a ground water level up to El. t>8'-9" 2'-0" bebw plant finished yard grade elevation (El. 90'-9").
- Soil Density (y): Saturated Soil = 145 pcf (pounds / cubic foot)
Moist Soil = 125 pcf ; Dry Soil = 80 pcf A
- Angle of Inttrnal Friction: 4 = 30' N ,)
- Coefficient of friction: = tan 4 (i.e., 0.57) assuming the concrete is poured directly on competent structural backfill without any intervening material, such as waterproofing; p = % tan & if concrete is not poured directly onto the soil.
The friction coefficient shall be funher reduced when intervening materials are used.
- At-rest lateral soil pressure coefficient: Ko = 0.5 (Used in Service Load Combinations) 2
- Active lateral soil pressure coefficient: K4 = tan (45' - 4/2)
- Passive lateral soil pressure coefficient: Kp = tan2 (45* + 4/2)
- Active lateral earthquake soil pressure coefficient: Kxg sins (&+p-0)
Ka= cos (0) sin2 (p) s n (p-O'-6) [1+ p ($+8) sin (&-O'-a) 32
% rdn (D-8-0) sin (a4 p)
- Passive lateral earthquake soil pressure coefficient: Kpg l
Nl\ Afywooed Desiges Material. Desigre of SSC Pope 3.8A-11
System 80+ oesign controtoccument sins (p+0'-4) g, cos (O') sin s (p) cos (6+p+0'-90) [1- sin (4+6) sin (4+a-09 2 3 h sin (p +5+0') sin (a+p) where: k, 6' = tan-, (1 - k,) kn
= averace horizontal earthauake acceleratiori component acceleration due to gravity, g k, = averace vertical earthauake acceleration component acceleration due to gravity, g a = The slope or angle of the backfill surface as measured from the horizontal
# = (1) The angle formed by the exterior face of the wall and the horizontal. (2) The angle shall be measured at 180' minus the angle formed by the exterior wall surface and the horizontal di:e. tion extending out under the backfill. (3) This value will be 90' for vertical walls.
6 = The angle of wall friction is a quantitative value, expressed in degrees, used to define the level of friction between soil backfill and the retaining structure The total lateral carth pressure is calculated as:
- At-rest lateral soil pressure: Po
= %KoyH 2
- Active lateral soil pressure: Pr = %KyH 3
2
- Passive lateral soil pressure: Pp = %KryH 2
- Active lateral earthquake soil pressure: Pn = %KayH2 (1ik,)
2
- Passive lateral earthquake soil pressure: Ppg = %KpsyH (1ik,)
where: y= soil density (pcf) H = height of soil-wall interface (ft) The effects of buildings, vehicles, cranes, material stockpiles, etc. acting as surcharge loads on the soil l adjacent to exterior building walls shall also be considered. l
- For factored load combinations, the lateral soil load shall be based upon saturated soil associated with flooding and a ground water level l'-0" below the finished plant yard. ;
I w,mr oosy wrenant. oesy or ssC Page 18A-r2 l i
t f Srtem 80+ Design ControlDocument (G
)
Reference:
Das, B.M., Principles of Foundation Eneineering, second ed., PWS-Kent Publishing Co., Boston,1990. 5.1.1.4 F - Ilydrostatic Loads flydrostatic loads are due to ground water, exterior flood waters, or fluid with fluctuating levels in internal compartments, including internal flooding. Maximum flood level is specified to be l'-0" below finished plant grade. Site specific flood elevations greater than this will be addressed in site-specific SAR. See also Section 2.0. 5.1.1.5 T, - Thermal Loads Thermal effects consist of thermally induced forces and moments resulting from plant operation or environmental conditions. Thermal loads and their effects are based on the critical transient or steady state condition. Thermal expansion loads due to axial restraint as well as loads resulting from thermal gradients shall be considered. The following ambient temperature values during normal conditions shall be used as a basis for design. Site specific provisions may be taken to minimize the effects of the structural temperature gradients produced by these conditions. External ambient conditions, reference Table 2.0-1. , A () Outside air temperatures - 100*F max.
-10'F min.
Ground Temperature - 50*F . Internal ambient conditions, reference Appendix 3.II A and Sections 10 and 11 of this appendix. Thermal analysis may be perfonned to determine concrete surface temperatures. 5.1.1.6 R,- Pipe Reactions Pipe reactions are those loads applied by piping distribution system supports during normal operating or shutdown conditions based on the critical transient or steady state condition. The dead weight of the 1 piping and its contents are included. Appropriate dynamic load factors shall be used when applying transient loads, such as water hammers. 5.1.2 Severe Environmental leads Severe environmentalloads are those loads that could infrequently be encountered during the life of the plant. Included in this category are Wind loads (W). 5.1.2.1 W - Wind Loads i [\; Wind loads are forces generated by the " Design Wind" of 110 mph at 33 ft above nominal ground elevation (Section 3.3.1.1). " Wind" does not include tornado force winds. Wind Loads are determined in accordance with ANSI /ASCE 7-88 or ASCE Papers 3269 and 4933 as specified in Section 3.8.4.3. 4promi Desiger Materiet - Design of SSC Page 3.8A-13 , I i
System 80+ Design ControlDocument l Loads calculated using ANSI /ASCE 7-88 shall be based upon a 0.01 % annual probability of occurrence in an assumed 100 year recurrence period for Category I and II structures. Other non-safety related structures use the same wind speed but with a 0.02 % annual probability of occurrence during an assumed recurrence period of 50 years. For safety related structures use an " imp (rtance factor" (I) of 1.11 with an exposure category of "C" as defined in ANSI /ASCE 7-88, Section 6.5..'. An "importance factor" of 1.0 should apply to r on-safety related structures. " Recurrence interval" and "importance factors" are discussed in ANSI /ASCE 7-88 Commentary Section 6.5.2. Design wind pressure, p (psf), shall be calculated by the formula p = q Cin C p, (Reference ANSI /ASCE 7-88, Sections 6.4 and 6.5) where: q = velocity pressure, (= 0.00256K,(IV)2. for V = 110 mph design wind speed), Gn
= gust response factor, ANSI /ASCE 7-88, Table 8 (dependent upon height, z, above ground),
Cp = external pressure coefficient, dependent on shape of the structure, negative pressure is < suction, K, = velocity pressure exposure coefficient from ANSI /ASCE 7-88 Table 6. 5.1.3 Extreme Environmental Loads . Extreme environmental loads are those which are credible but are highly improbable. 5.1.3.1 E' - Safe Shutdown Eanhquake (SSE) SSE loads are loads generated by an canhquake with a peak horizontal ground acceleration of 0.30g. Refer to Section 2.5.2.5.1. Total loads for E'shall consider simultaneous seismic accelerations acting in three orthogonal directions (two horizontal and one vertical). Each of the three directional components of the earthquakes will produce responses in all three directions. Colinear responses due to each of the 3 individual canhquakes may be combined using the " Square Root of the Sum of the Squares" (SRSS) method. The resultant nodal loads are applied simultaneously to the structure. The scismic forces and moments may also be combined simultaneously using directional combination participation factors of 100%/40%/40% applied to the individual loads produced as a result of each earthquake to produce the design SSE loads. The critical load combination would use 100% of the loads due to one earthquake and 40% due to the other 2 earthquakes, i.e., E of F, due to i100%E', i40%E'y 40 % E',. SSE loads are obtained by multiplying the dead load and 25 % of the design live load by the structural acceleration obtained from the seismic analysis of the structure. The full potential live load shall be used for local analyses of structural members. Amplification of these accelerations due to flexibility of structural members should be considered. Construction loads are not required to be included when determining seismic loads. Other temporary loads must be evaluated for applicability on a case by case basis. Seismic toil loads shall be used in combination with E'. SSE damping values used in design (Reference NRC Reg Guide 1.61 and Table 3.7-1) shall be as follows: i Atyweved Desiers Material . Designs of SSC page 3.gA.r4
- - . . - - ~ . _
l I System 80+ Design ControlDocument i V Structure Type % of Critier.1 Damping Welded Steel 4 Bolted Steel 7 Reinforced Concrete 7 Prestressed Concrete 5 Equipment (steel assembly) 3 i Fluid sloshing loads in the IRWST, Spent Fuel Pool, and all other fluid reservoirs due to the SSE shall be considered in accordance with ASCE 4-86. ; - 5.1.3.2 W - Tornado Loads i Loads generated by the design tornado are as identified in Section 3.3.2. Tornado loads include loads due to the tornado wind pressure (W,), the tornado created differential pressure (Wp), and .
. tornado-generated missiles (W m ). Twenty-five percent of the design live load shall be considered with tornado load combinations. The full potential live load shall be used for local analyses of structural :
members. The following parameters from Section 3.3.2.1 and Table 2.0-1 shall be used for the design basis tornado:
- Maximum wind speed = 330 mph I t
V
- Maximum rotational speed = 260 mph
- Maximum translational speed = 70 mph
- Radius of maximum rotational speed = 150 ft
- Maximum pressure drop = 2.4 psi
- Rate of pressure drop = 1.7 psi /sec Tornado winds loads shall be converted a wind pressure loads in accordance with ASCE 7-88. The tornado velocity pressure is to be considered constant with height, therefore Kz = 1.0. In determining tornado wind loadings, the importance factor (I) and the gust factor (Gn) for tornadoes shall be taken as being unity. Therefore, tornado wind pressure loads shall be computed by the formula; i
2 p = 0.00256K,(VI)2G3 Cp = 0.00256V C p l The external pressure shape coefficient is determined in accordance with ASCE 7-88, Figure 2. j i The tornado induced pressure for interior wall design shall be taken as - 0.5 psi unless a lower value can be justified, p) ( Tornado missiles shall be considered in accordance with SRP Section 3.5.1.4 Spectrum II, Region I, and Table 3.5 ' j w .a o w unw w.oes+ orssc roue 3.as.1s l l
System 80+ Design ControlDocument Design for missile impacts shall be in accordance with Section 3.5.0 snd ACI 349, Appendix C. Minimum concrete wall and roof thicknesses shall be in accordance with Standard Review Plan 3.5.3 Table 1. Non-Category I structures shall not be assumed to shield seismic Category 1 structures from tornado wind, differential pressure, or missile loads. 5.1.4 Aboormal Loads Abnormal loads are those loads generated by a postulated high-energy pipe break accident. This event is classified as a " Design Basis Accident". Included in this category are: Pressure loads (P,), Thermal loads (T,), Pipe reactions (R,), Load on the structure generated by the reaction on the pipe (Y,), Jet impingement loads (Y)), and Missile impact loads (Y,). These loads are defined by:
- P, - Pressure equivalent static load within or across a compartment and/or building, generated by the postulated break, and including an appropriate dynsmic load factor to account for the dynamic nature of the load.
- T, - Thermal loads generated by the postulated break and including To .
- R,- Pipe reactions generated by the postulated break and including R o.
- Y, - Equivalent static load on the stmeture generated by reaction of the broken high-energy pipe during the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.
- Y j- Jet impingement equivalent static load on a structure generated by the postulated break, and including an appropriate dynamic load factor to account for the dynamic nature of the load.
- Y, - Missile impact equivalent static load on the structure generated by or during the postulated break, such as pipe whipping, and including an appropriate dynamic load factor to account for the dynamic nature of the load.
5.2 Design Load Combinations 5.2.1 General The following loading combinations shall be used for analysis and design of Category I structures and their components. Live loads shall be applied (fully or partially), removed, or shifted in location and pattern as necessary to obtain the worst case loading conditions for maximizing internal moments and forces for all load combinations, impact forces due to moving loads shall be applied where appropriate. Where any load is determined to have a mitigating effect on the overall loading for a steel or concrete structural member, a load coefficient of 0.9 should be applied to that load component. The reducing coefficient should be used only for that load which can be demonstrated to be always present or occurring simultaneously with other loads. For loads which cannot be shown to be always present, the coefficient for the counteracting load is set to zero. The 0.9 coefficient should be used in lieu of the calculated concrete and steel coefficients; Approved Design MeterW ~ Desipre of SSC Page .1.8A.16
I l System 80+ Design ceneralDocument l
/
(% v l
- for concrete replace 0.75*1.4D or 0.75*l.7L with 0.9D or 0.9L, ACI 349 Section 9.2.3, and
- for steel 1.7*1.0D with 0.9*D.
No increase in allowable loads due to wind in service load combinations is permitted for steel or concrete j - components. 5.2.2 Loading Combinations for Seismic Category I Concrete Structures ! The following set of load combinations define deWn limits for all Seismic Category I concrete structures: 5.2.2.1 Service Load Conditions l l The following load combinations represent normal operating conditions or combinations of normal loads with severe environmental conditions.
- U = 1.4D + 1.7L i l
- U = 1.2D + 1.7W
- U = 1.4D + 1.7F + 1.7L + 1.7H + 1.7W ,
l If the thermal stresses due to R and Toare present, the following combinations shall be satisfied.
- U = (.75)(1.4D + 1.7F + 1.7L + 1.7H + 1.7To + 1.7Ro)
- U = (.75)(1.4D + 1.7F + 1.7L + 1.7H + 1.7To + 1.7Ro + 1.7W) t For concrete structures, U is the section strength required to resist design loads based upon the ultimate strength design methods described in ACI 349.
5.2.2.2 Factored Load Conditions The following load combinations represent combinations of normal operating loads with (either or both) extreme environmental loads or abnormal loads.
- U = D + F + L + H + To + % + E'
- U = D + F + L + H + T, + R o+ W, i for W, use: W,, W , or W, individually and in combination, (W, + 0.5Wp), , + W,), or (W, + 0.5W, + W,)
- U = D + F + L + H + T, + R, + 1.5P,
- U = D + F + L + H + T + R, + 1.0P, + 1.0(Y,+Y)+Y ) + E' {
For load combinations above, the maximum values of P., T., R , Y;, Y,, and Y., including an appropriate dynamic load factor, are used unless a time-history analysis is performed tojustify otherwise.
?;v -.; Desigen neeseniel- Desiger of SSC Page 3.8A-17
System 80+ Design Control Document Ductility ratios determined from ACI 349 Appendix C should be used. Deflections shall be evaluated for potential loss of function for safety related systems. The second load combination in 5.2.2.2 shall first be satisfied without the tornado misti'e load. The fourth load combination in 5.2.2.2 shall first be satisfied without the Y loads. When including these loads .awever, local section strength capacities may be exceeded under the effect of these concentrated loads, provided there will be no loss of function of any safety related system. Structural effects of differential settlement, creep, or shrinkage shall be included with the dead load. 5.2.3 Loading Combinations for Seismic Category I Steel Structures The following set of load combinations define design requirements used for all Seismic Category I steel structures. 5.2.3.1 Service Load Conditions ,
\
5.2.3.1.1 If clastic allowable strength design methods are used: l
- S=D+F+L+H
- S=D+F+L+H+W If thermal stresses due to oT and R o are present, the following combinatir re also satisfied:
- 1.5 S = D + F + L + H + R +o T (tension o members) 1.3 S = D + F + L + H + R +oT (compression o members)
- 1.5 S = D + F + L + H + R +o T + oW (tension members) 1.3 S = D + F + L + H + R, + T + W o (compression members)
For steel members, S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of ANSI /AISC N690 5.2.3.1.2 If plastic design methods are used: 1
- Y = 1.7 (D + F + L + H)
- Y = 1.7 (D + F + L + H + W)
- Y = 1.3 (D + F + L + H + To + R o)
- Y = 1.3 (D + F + L + H + To + Ro + W)
! For steel members Y is the section strength required to resist design loads based on the plastic design I methods described in Part 2 of ANSI /AISC N690. O Approwd Design Material Design of SSC Page 3.8A-18
e
' System 80+ Deakn controlDocurnorrt
~5.2.3.2 Factored Load Conditions ,
i 5.2.3.2.1 . If clastic allowable strength design methods are used:
- e. 1.6 S = D + F + L + H + R, + T o+ E' (tension tmbers) 1.4 S = D .+ F + L + H + R, + T + oE' (compression members) j_ e 1.6 S = D + F + L + H + R, + T +o W,(tension members) 1.4 S = D + F + L + H + R, + T + W,(compression o members) . j e for W, use: W,, W,, or W, individually and in combination,
- (W,' + 0.5Wp ), (W, + W,), or (W + 0.5 Wp +.W,) .
e 1.6 S '= D + F + L + H + R, + T, + P (tension members) -l 1.4 S = D + F + L + H + R, + T + P (compression members)' l e 1.7 S = D + F + L + H + R, + T, + (Y,+Y +Y,) j + E' + P (tension members) l 1.6 S = D + F + L + H + R, + T, + (Y,+Y)+Y ) + E' + P, (compression members) i
.(The plastic section modulus for steel shapes may be used for this load combination.) i 5.2.3.2.2 If plastic design methods are used: l l
e Y* = 1.0 (D + F + L + H + Ro + To + E') !
- O e. Y' = 1.0 (D . + F + L + H + R, + To + W,)
$$ for W, use: W,, Wp, or W, individually and in combination, (W, + 0.5W,), (W, + W ), or (W, + 0.5 Wp + W,) : t o y' = 1.0 (D + F + L + H + R, + T + 1.5 P.) l o Y' = 1.0 (D + F + L + H + R, + T, + Y, + Yj + Y, + E' + P )
' .*use 0.9Y for Internal Structures and 1.0 for all other Category I structures. -(Reference SRP 3.8.3.II.5)
I ! 5.2.4 Loading Combinations for Sliding, Overturning, and Flotation i Minimum Factors of Safety . Load Combination Overturning Sliding Flotation { t D+H+W 1.5 1.5 - l D + H + W,. 1.1 1.1 -
-f
i D + H + E'- '1.1 1.1 -
- D '+ F ' - - 1.1 .
i q , Anwewur Deew neeeuw. cee> er ssc - rene 2.sa.ts
Sy~ tem 80+ Design Control Document 5.2.5 Construction Load Combinations Service load combinations shall be used to evaluate construction methods and sequence and determine structural integrity of the partially erected structures. 5.2.6 Applicability of Loads
- Lateral loads due to soil bearing pressure shall apply to all exterior walls up to El. 90*-9".
Tornado loads shall be applied to roofs and all exterior walls of all safety-related Category I Structures. Where required tornado pressure boundaries are not established at the exterior walls, appropriate interior walls shall be designed as tornado pressure boundaries. External hydrostatic forces are applicable to the basemat and to all exterior walls up to elevation 89'-9". 6.0 Structural Analysis and Design, Requirements and Procedures 6.1 Analysis The Seismic Category I and II structures are analyzed to account for both global and local effects of design basis loads described in Sections 3.8.3 through 3.8.5. 6.1.1 General The major structures that make tq, the Nuclear Island are described in Section 3.1 of this appendix. O Othen structures are described in Section 10.0. The supplemental design criteria for the Nuclear Island and the Non-Nuclear Island strucures are identified in Sections 10.0 and 11.0. A detailed three dimensional finita element model of the Nuclear Island is developed to distribute the global loads to all structural compments. This model is shown in Figure 3.8A-2. The model includes all walls, floor slabs, and major stn.ctural beams and columns as well as the foundation mat and a spring representation of the underlying soil media. Shell elements are used to model all walls and slabs. Beam elements are used to model all structural beams and columns. Three dimensional solid elements are used to model the basemat and thick walls. Two different versions of the model are created, one corresponding to the rock site conditions and on.c for the soft soil conditions, in order to adequately capture the bounding spectmm of the load dictribtttion within the superstructure from different soil stiffnesses. The superstructure is identical in both models. The only difference in the two models is the stiffness of the spring elements representing the soil. Equivalent static global loading conditions are applied to the models and results are combined using the loading combinations identified in Section 5.2 of this appendix. The global results from the three dimensional finite element model are added to local loading results to develop input for the design of the walls, columns and slabs. The resuhs are used to design for in-plane ferces and moments only. l Design out-of-plane forces and moments are determined by hand calculations or local area models. Local j model end conditions such as displacements and forces are compared to equivalent in-plane results from j the three dimensional global static model to ensure compatibility with global loading results. l l i Approved Design Meterial- Designs of SSC Page 3.8A-20 l
i i l Svitam 80+ Desfan c:ntrolDocument l c i
, 5
~
- 6.1.2 Seimnic Analysis !
1 The seismic inenia loads are determined from the three dimensional model results in the following , . manner. The Zero Period of Acceleration (ZPA) values are extracted for each area from the Soil
, Structure Interaction (SSI) analyses described in Section 3.7B. For each elevation these ZPA values are f enveloped from the values of each stick in the SSI model at the corresponding elevation. This enveloping l of ZPA values at each elevation is repeated for all control motions and all soil cases, and a final envelope !
of ZPA values for each elevation is determined. The ZPA values are further amplified where necessary ~ to account for floor slab fluibili:y. Figures 3.8A-3 through 3.8A-5 show this envelope profile for the ! NS, EW and venical directions respectively. This envelope of ZPA values is applied as a uniform factor i
- to the floor mass and contributing ponions of the wall masses at each elevation within the structure as j the applied sei*,mic loading. l For the so t soil model this envelope loading is obtained from an envelope of soil cases B-2, C-2, C-3 d
I and C 1.5. These soil cases represent the soft soil site category. Applying the envelope of the soft soit j ZPA values is compatible with the soil stiffness modeled for the soft soils. { t j The enveloping ZPAs are used in the local analyses to determine the forces and moments from the inenia loads.- The masses in the local models are accelerated by the appropriate ZPA value for the elevation ; I
- being analyzed and the forces are applied as static point loads, static body forces, or static uniformly distributed loads.
I For each load the response from all three directional canhquakes are combined simultaneously. The j .. independent directional responses are combined using the square root of the sum of the squares (SRSS) ! method or the 100-40-40 Percent Rule described in ASCE 4-86. The 100-40-40 Rule is based on the ; - observation that the maximum increase in the resultant for two onhogonal forces occurs when these forces j - are equal. The maximum value is 1.4 times one component. All possible combinations of the three { onhogonal responses are considered. The 100-40-40 combination is expressed mathematically as. , L R=( l.0Rx i 0.4Ry i 0.4Rz) ! or, l R= (i0.4Rx i 1.0Ry ! 0.4Rz) or, R= (i0.4Rx 0.4Ry i 1,0Rz) l The 100-40-40 Percent Rule may also be applied for combining responses in the same direction due to j l
- different components of motion. l Additional seismic loads due to accidental torsion is accounted for as required by SRP _
Section 3.7.2.11.1 1. This accounts for variations in material densities, member sizes, architectural- ! variations, equipment loads, etc., from design assumptions. Due to these potential variations, an ! additional eccentricity of the mass at each floor equivalent to 5% of the maximum building dimension f is included. The accidental torsion load is an additional shear force at each floor elevation determined based on a percentage of total accumulated shear at each elevation. The dynamic increment for horizontal soil loads on the exterior walls of the Nuclear Island, CCW heat exchanger structure and diesel fuel storege structure is determined from the 2D SSI analyses as described , in Section 3.7J For other structures, the elastic solution method in ASCE 4-86 is used. [ t r I
"~
4weemt neeon naeserw.oeeon er ssc roue 2.sa.2 r N
.- - - - . = _ . .a--. . . . - . - - - -. - . . . , _ . . . - , - , - - . - , . ....-
I I Sy: tem 80+ Design ControlDocument l i The dynamic increment for the surcharge loads on the exterior walls is determined by multiplying the surcharge static load by the maximum vertical ZPA at the ground surface determined in the seismic ) analyses in Section 3.7. The basemat forces and moments are obtained from static analyses using the three dimensional finite element model of the Nuclear Island as shown in Figure 3.8A-2. Two bounding analyses are performed corresponding to the rock and soft scil conditions. Where potential uplift in the basemat corners due to superposition of seismic loads, dead loads, and potential maximum buoyancy force is indicated in the two dimensional dynamic SSI analyses dest.ribed in Section 3.7, the soil springs in the affected area in the three dimensional fir *e element model are released and the static analysis is repeated. The envelope of basemat moments and forces from all analysis cases are used in the design. Inertial loads from sloshing fluids are determined by the method identified in standard ASCE 4-86 or ASCE Manual No. 58. 6.1.3 Thermal Analysis ACI 349 ApperAix A, ACI report ACI 349.1R or thermal analysis computer programs are used to evaluate thermally induced forces and moments. Thermal analyses may be performed to determine actual concrete surface temperatures. Ambient temperature values are provided in Section 5.1.1.5. 6.1.4 Other Analyses All other loads described in Section 5.1 are analyzed as static point loads, static body forces, or static uniformly distributed loads. 6.2- Structural Design 6.2.1 General Requirements 6.2.1.1 Concrete The requirements for the design and construction of Seismic Category I concrete structures shall conform to all requirements of ACI 349 and NRC Regulatory Guide 1.142, except as modified by this appendix. The Seismic Category I concrete members are designed as if they were parts of ordinary moment frames. These frames shall be designed for strength based on the Strength Design Method as defined in ACI 349. Deformations are assumed to be restricted with energy dissipation occurring due to elastic deformation. ACI 318 Chapter 21 8is incorporated into the design configuration of Seismic Category I member connections as if they were parts of special moment frames. Special moment frames assume larger deformations which could lead to the formation of hings in areas of maximum moments. ACI 318, Chapter 21 shall be used for Seismic Category I structures to determine the required anchorage and splicing of connections, and to determine the configuration of reinforcing in the structural joints, and regions where reinforcing is spliced, and required placemen of stirrups and hoop steel. I Reference of Appendix A of 318 referenced in Reg. Guide 1.142 and ACI 349-85 are now included in O ACI 318-89 with limited revisions. Approved Design Materis! Design of SSC Page 3.8A-22 m
System 80+ oestan controlDocument . 1 Exceptions to use of ACI 349 Appendix B for the design of embedments and expansion anchors are ; defined in Section 6.2.1.1.2. Design provisions for impulsive and impactive effects in Seismic Category I stmetures shall be in ) accordance with ACI 349-90, Appendix C. i ACI Specification ACI 349 Appendix A, ACI Report ACI 349.1R or computer analysis programs are ; used to evaluate thermally induced forces and moments in Seismic Category I structural members. l Masonry block walls shall not be used in Seismic Category I or II structures. 6.2.1.1.1 Reinforcing Required reinforcing for Seismic Category I concrete members shall be determined in accordance with ACI 349. , i 4 Concrete joints in Seismic Category I structures shall be detailed for ductility in accordance with Chapter 21 of ACI 318,1989 edition,1991 printing. Supplemental reinforcing requirements based upon ACI 318 are: i i
- Mechanical or welded splices are permitted subject to a 50% limit at a given location with splices on remaining bars staggered at least 24" between centerline of adjacent splices, Paragraph 21.3.2.4.
- Lap splices in beam and columns shall have hoop reinforcement over the length of the splice. l Hoop reinforcement will be sized per Paragraphs 21.3.2.3 and 7.10.5.1.
l
- Hoop reinforcing for beams and columns shall be installed as required by l Paragraphs 21.3.3.l(1) and 21.3.3.2 and Section 7.10.5.
- Spacing (<4" or <d/4) of hoop reinforcing will be according to Paragraph 21.4.4.2 with crossties or legs spaced no more 14" per Paragraph 21.4.4.3 distributed over a length specified in Paragraph 21.4.4.4. !
i
)
' '* Reinforcing at terminating ends of beam, walls and columns shall be in accordance with ! Paragraphs 21.4.4.5 and 21.5.3.5, as appropriate. ] l
- Transverse reinforcing at the edges of wall panels shall be anchored in accordance with Paragraphs 21.5.3.5 and 21.5.3.6. .
- Longitudinal reinforcing for beams shall be anchored according to Paragraph 21.6.1.3 with hoop reinforcement per Paragraph 21.6.2.1.
- Development lengths for reinforcing will be according to Paragraph 21.6.4.
' i l
l 1 L .z De& hineenial. Deelen of SSC Pope 3.8A.23 <
i I I System 80+ Design ControlDocument Unless the ground water level is below the foundation level, due to either natural site conditions or ' provision of a permanent dewatering system to accommodate site-specific conditions, epoxy coated reinforcing shall be used for exterior walls and slabs when the existing groundwater is determined to be sufficiently corrosive so as to adversely affect the long term durability of the concrete structure. When epoxy coated reinforcing is used, the required splice length given in ACI 349 Section 12.2.2 shall be increased using factors provided in ACI 318 Section 12.2.4.3. When feasible, uniform reinforcement patterns she>uld be used for sections with similar requirements, thickness and loading. 6.2.1.1.2 Concrete Expansion Anchors Concrete expansion anchors shall meet the requirements of Section 3.8.4.5.1. A specification for the design, installation, and use of expansion anchors should be developed and include:
- expansion anchor allowable loads, e expansion anchor minimum spacing,
- spacing requirements for expansion anchors, e procedures for addressing baseplate flexibility's in calculating design loads on expansion anchors, e procedures for addressing shear tension interaction, and e required load reductions for cyclic loadings.
See also Section 3.8. When high capacity concrete anchors are specified, they should be of the direct bearing or " undercut" type. Load transfer for these anchors is achieved by bearing of the expanded embedded tip against the undercut coxtete hole produced by a special flaring tool. Undercutting of the concrete is required for the anchor to provide the concrete shear capacity to match the high strength bolts. For smaller safety related or non-safety related applications expansion anchors referred to as " Sleeves" or " Wedges" may be used, subject to the safety factors given in Section 3.8.4.5.1. 6.2.1.2 Steel The design of Category I steel structures and/or components shall use Allowable Strength Design methods in accordance with ANSI /AISC N690 as amended by Section 3.8.4.5.2. Uniform depths of steel beams and connections should be maintained. Bolted connections should be used for field erection of structural steel beams and columns. Load indicator bolts are recommended. The design of bolted connections shall be in accordance with ANSI N690 Section Ql.16 and the " Specification for Structural Joints Using ASTM A325 or A490 Bolts". Astroved Design Material- Design of SSC Page 3.8A-24
System 80+ Design ControlDocument Bolted connections shall be designed to be " slip critical" unless justified otherwise. O) The requirements for welded connections are defined in Section 7.2.3 of this appendix. Maximum utilization of shop fabricated connections should be considered to avoid welding in hazardous j environments. Transverse welds across the flanges of rolled Sections of Seismic Category I or II steel members are prohibited without approval of the design engineer. All transverse welds on Category I or 11 members shall be shown on approved drawings. Structural members with restrained end conditions and thermal loads shall be evaluated for potential buckling. 6.2.1.3 Missile Protection Exterior walls and roof slabs of Seismic Category I structures are required to function as missile barriers for tornado generated missiles. Design of missile barriers shall assure that the structure will not collapse under the missile load nor will there be penetration through the barrier. Safety related stmetures, systems and components shall be protected from secondary missiles as a result of backface scabbing. Where it is evaluated to be necessary, interior walls and floors shall be designed to function as missile barriers. p 6.2.1.4 Fire Protection G Fire protection is provided in the form of fire rated walls and barriers as identified in Figure 3.8-5. In ! addition to passive fire protection offered by fire rated structural barriers, the structural design shall offer protection to the active fire suppression system to assure that they will not be made inoperable due to the failure of any structural member. 6.2.1.5 Flooding Flooding is addressed in Section 3.4.4. Flood barriers are identified in Figure 3.8-5. Protection of the Seismic Category I structures against flooding shall be ensured by:
- allowing no access openings in the exterior walls lower than I foot above plant grade, 4
- having no unsealed exterior wall or floor penetrations below plant flood level (El. 89'-9",
I foot below finished yard grade),
- having water stops in all below grade exterior construction joints, and
- providing floor drainage.
- _ .,aac -e
Syntem 80+ Design ControlDocument 6.2.1.6 Coastruction Support Cost saving may be achieved by reducing the duration of the construction schedule. Durations may be reduced by standardizing details and using modular designs that will allow offsite fabrication and assembly. Modu!ar designs must consider transpottability to the point ofinstallation. Connections / fit-ups with previously erected components must be considered. 6.2.1.7 Security 10CFR Chapter I Part 73 provides the regulatory requirements for physical protection of the plant against sabotage as a result of unauthorized access. Plant designs shall prevent use of unauthorized access routes. In accordance with Part 73 Section 45(f)(1)(i), barriers shall be provided to channel access through protected area entry control points or delay any unauthorized penetration attempt sufficiently to allow detection by security personnel. 6.2.2 Special Design Criteria 6.2.2.1 Radiation / Contamination Control The design of structural elements shall provide surface features to prevent the spread of contamination and facilitate plant cleanup. Sumps for drain lines that may collect potentially contaminated liquids will be lined with stainless steel over the potentially wetted surface. Concrete surfaces should be protected by a smooth surface epoxy coating where the potential exists for contamination. Walls or curbs shall be included around locations of potential leaks of contaminated fluids. Penetrations , in walls or floors shall be fitted with appropriate seals to prevent the spread of contaminated fluids. I 6.2.2.2 Grout Grout shall be selected based upon required bearing strength and exposure conditions. A field specification should be prepared to provide instructions in selecting site approved grouts. The grout specification should also include instructions for concrete repairs. See also Section 3.8. l 6.2.2.3 Roof Drains Primary and secondary roof drains shall be provided on all structures with parapets to assure that the load resulting from rainfall will be less than the design load of 50 psf. Roof drains shall be located to eliminate ponding where the potential for excessive roof deflections may exist. A minimum roof slope of %" per foot is recommended to further reduce the potential for ponding. Scuppers may be used as , secondary roof drains. l 6.2.2.4 Adjacent Structures , Non-Seismic Category 1 Structures shall be designed or located to prevent any adverse interaction with Seismic Category I Structures. . 6.2.2.5 Wall / Floor Penetrations Requirements All openings in walls and slabs of Seismic Category I and Il structures shall be shown on construction ; drawings. Openings shall be acceptable without analysis if they meet the criteria identified in ACI 349 l Approwd Design Atatorial- Design of SSC Page 3.8A 26
Srtem 80+ Design ControlDocument
) Section 13.5.2. Penetrations shall not be added to an erected safety related Seismic Category I or II '" concrete wall or slab without prior evaluation by the Design Engineer.
Round pipe sleeves shall be used in lieu of rectangular penetrations except where required by other design criteria. Each corner of rectangular openings in walls or slabs should be provided with diagonal reinforcing to reduce cracking due to stress concentration at these locations .a accordance with ACI 349 Section 14.3.4. 6.2.2.6 Miscellaneous Components 6.2.2.6.1 Plationns, Ladders, and Manways Seismic Category I safety related platforms shall be designed and installed in accordance with Sections 6.2.1.2 and 7.2.1 of this appendix. Access structures not supporting Safety Class equipment shall be designed as Seismic Category II. 6.2.2.6.2 Electrical Cable Tray and HVAC Ductwork Design of building structures for support of cable trays and HVAC ductwork shall meet the requirements of Appendix 3.9A. rm 6.2.2.6.3 Support / Restraints for Piping and Its Components b Design of building structures for support of piping and its components shall meet the requirements of Appendix 3.9A. 6.2.2.6.4 Fabricated Embedments The walls and floors of Seismic Category I Structures shall be provided with embedments for the mounting or attachment of structures and components. Additional typical embedments should be provided for welding structural attachments which will reduce the number of attachments utilizing expansion anchors. Tolerances for fabrication and installation of embedments shall be provided on design drawings or in specifications. See also Section 3.8. The anchorage for structural embedments shall be designed based upon ACI 349, Appendix B with the following exception. The assumed concrete failure cone projects out at an angle of 35' instead of 45*. The angle shall be measured from the plane normal to the axis of the embedment. This exception applies to structural embedments and headed anchors, such as " NELSON Studs", and expansion anchors. The exception is to prevent an overlapping of the concrete shear cones when anchors are spaced at a "2d" macing (reference Section 3.8.4.5) and to avoid a less than required minimum edge distance. A reduction in load capacity for embedments shall be applied for placement of anchors in the tension zone of concrete members. n f )
/sproved Design Material- Design of SSC Pope 3.8A-27
System 80+ De:ign ControlDocument 7.0 Construction; Forming, Fabrication, and Erection h 7.1 Concrete Concrete work for Seismic Category I structures shall conform to all requirements of ACI 349 and ACI 301 except as modified by this appendix. 7.1.1 Concrete Mix Design Concrete mix design for Seismic Category I structures, see Section 9.2 of this appendix, shall be determined based upon field testing of trial mixtures with the materials to be used. Testing shall evaluate:
- ultimate concrete strength as well as early strength in support of an aggressive construction schedule, e concrete workability and consistency,
- required concrete admixtures,
- heat of hydration and required temperature control for large or thick concrete pours, and e special exposure requirements when identified on design drawings.
7.1.2 Concrete Placement Requirements and/or limitations on concrete placement will be determined in conjunction with the construction schedule. A site specific construction specification should be prepared to address requirements and procedures for concrete placement. See also Section 3.8. The concrete specification should address: o desired volume of concrete pours and rate of deposition,
- special forming requiremeras, o maximum height of pours.
- temperature Insations; weather conditions and concrete mix, including approved methods for temperature wntrol, and
- curing requirements and procedures.
7.1.3 Reinforcing Fabrication and placing of reinforcing bars for concrete in Seismic Category I structures shall confctm , to the rquirements and tolerances specified in ACI 349 Section 7.5 and in ACI 301 Sections 5.5. 5.6, and 5.7. AMroved Design MaterW- Design of SSC Page 3.8A 28
Sv~ tem 80+ oesten controloccament , P Consideration shall be given for modular assemblies of reinforcing. Such assemblies shall be designed i to be moved without changing their alignment.
]
- Lap splices shall be prohibited for locations with tension stresses normal to the plane for the splice and for bar sizes greater than #11, except as provide by ACI 349 Section 12.14.2.1. ,
Welding of reinforcing shall be prohibited except as provided for in approved splice details. Welding } al con form to t he requ rements
. shl i of AWS D1.4, " Structural Welding Code - Reinforcing Steel." Welded ;
reinforcing shall be shown on reinforcing drawing details. ! 7.1.4 Construction Sequencing Construction sequence will be included in site-specific information. Additional design requirements due - to the construction sequence will be determined during the final design. See also Section 3.8. i I
- Advanced construction methods, such as modular construction or forming concrete slabs using metal l
- -decit, steel beams and columns which may be used to facilitate the construction sequence, which will !
affect design details must be justified by as-built analyses and results documented in the structural analysis ; d~ report described in the structural acceptance criteria, Section 3.8.4.5.4. ! i 7.2 Structural Steel t 7.2.1 Structural Steel; Fabrication and Erection
- Fabrication and erection of safety related steel members shall be in accordance with AISC N690, Sections i Q1.23 and Ql.25. Additional r quirements are applicable as provided for in this appendix. j 7.2.2 High Strength Bolted Connections !
l Bolts shall be installed and tightened in accordance with Section 8(d) of " Specification for Structural ; Joints Using ASTM A325 or A490 Bolts." The use of " load indicator" bolts or washers should be used ; where possible. " Snug tight" installation of bolts in " slip critical" connections shall not be permitted. 4 , 7.2.3 Welded Connections Welding activities associated with Seismic Category I structural steel and their connections shall be !
] accomplished in accordance with written procedures and shall meet the requirements of the AWS DI.1 .
Structural Welding Code. The visual acceptance criteria shall be as defined in NCIG-01, " Visual l Acceptance Criteria for Structural Welding of Nuclear Power Plants," Revision 2, EPRI NP-5380. :
- 8.0 ' Structural Acceptance Criteria ;
! . Structural Acceptance Criteria are specified in Section 3.8.4.5.
i Separation Criteria for Seismic Category I and non-Seismic Category structures and components shall be i verified. j
- t et ;
t Approwed Deeker neeenrant- Deehrt of SSC Pnge 3.8A 29 ,
System 80+ Design ControlDocument 9.0 Materials h 9.1 General Material shall confonn to requirements for Section 3.8.4.6.1 and this appendix. Materials used should be selected based upon a proven record of service in other nuclear facilities. Materials shall be specified based upon approved codes and standards. Additional material restrictions or requirements may be added by the design engineer to meet anticipated design or field conditions. With suitable qualification and no applicable material restrictions, substitute materials may be used. Materials used shall be qualified to withstand environmental conditions for normal and accident conditions. Site specific design specifications should identify required qualifying environmental conditions. See also Section 3.8. 9.2 Specifications The materials identified below and in Section 3.8.4.6.1 shall be considered acceptable for the analysis and design of System 80+ Standard Plant structures. Additional materials may be added to this criteria when qualified by appropriate codes and standards. 9.2.1 Concrete Normal weight with a density of 135 to 160 pef. Compressive strength = 4000 psi Nuclear Island basemat - Non-Nuclear Island structures Compressive strength = 5000 psi Nuclear Island superstructure (A concrete strength of 4000 psi may be used when justified by as-built analyses and design details with results documented in the structural analysis report described in the structural acceptance criteria, Section 3.8.4.5.4.) Cement - material shall conform to ASTM C 150 per ACI 349 par. 3.2. Cement shall conform to Type 1 or Type 11 designations except where additional qualifications are conducted for special applications. Aggregates - material shall conform to ASTM C 33 per ACI 349 par. 3.3. ASTM specification C 637 may apply where deemed necessary for radiation shielding. Limestone based aggregates should be ; considered for use in the floor of the reactor cavity for core concrete interaction concerns. j Admixtures - Admixtures conforming to applicable ASTM standards are acceptable when qualified by testing to verify required mix design. l
)
Approved Design nCaterial Desips of SSC Page 3.8A-30 l l
Sy-tem 80+ oesign controlDocument f. Water shall conform to requirements of ACI 349 Section 3.4 and Section 3.8.4.6.1.1. Use of ( non-potable water shall be restricted in accordance with ACI 349 Section 3.4.3. Reinforcing Steel- ASTM A615 Grade 60, Fy = 60,000 psi or - ASTM A706 Fy = 60,000 psi The use of welded splices and mechanical connections is addressed under Paragraph 12.14.3 of ACI 349. , Mechanical reinforcing coupler devices may be used. Epoxy coating of reinforcing shall be in accordance with ASTM A775 (ACI 318 paragraph 3.5.3.7). 9.2.2 Steel ] 9.2.2.1 Structural Steel Structural Shapes - ASTM-A36, Fy = 36,000 psi additional material per ANSI /AISC N690 Section Q1.4.1 (excluding round & tubular shapes) Structural Tubing - ASTM-A500 Grade B, Fy = 42,000 psi Steel Plates - ASTM A240 Type 304L Stainless Steel ASTM A36 I V 9.2.2.2 Structural Bolts Structural Bolts shall comply with ASTM material specifications identified in Section Q1.4.3 of the ANSI /AISC Standard N690 or other materials identified in the " Specification for Structural Bolting Using ASTM A325 or A490 Bolts" Bolts shall have nuts and washers as identified below:
- Bolts - A193, A320, A325, A490, A354, or A449
- Nuts, for A325 and A449 - A194 Grade 2 or 2H nuts or A563 Grade C, C3, D, DH, or DH3
- Nuts for A193, A320, A354, and A490 - A194 Grade 24 or A563 Grade DH or DH3
- Washers - F436 hardened steel washers.
High strength threaded rods such as A193 Grade B7 or A320 Grade L43 may be used in lieu of A325 bolts with qualifying documentation identifying the installation. 9.2.2.3 Welding Welding materials shall conform to the requirements of the Structural Welding Code (AWS-DI.1). AWS D1,1 Table 4.1.1 shows the compatibility of filler metal with base metal. ANSI /AISC N690 provides supplemental information on weld materials for stainless steel. fn Amew onien www outon or ssc roue 3.sA.31
System 80+ Design ControlDocument 9.3 Restricted Material The use of the restricted materials should be based upon a proven need and avoided where possible. Materials that are restricted include:
- Use of teflon based low friction sliding bearing plates such as "Flurogold" or neoprene based gaskets, seals, or bearings shall be kept to a minimum due to presence of fluoride or chloride ions and the increased potential for stress corrosion cracking.
- I.ow melting point metals (lead, zinc, etc.) have been identified for their deleterious effect on corrosion resistance and ductility of metallic components. Restrictions on zine will also mean a restriction on galvanized materials. This, restriction is particularly applicable inside Containment where the zine in the galvanized coating can result in chemical reactions producing additional hydrogen.
10.0 Supplemental Design Criteria for Nuclear Island, Category I and II Structures All structures located on the Nuclear Island are Seismic Category I, Safety Class 3, and Quality Class 1. Refer to Figure 3.8A-1 of this appendix for location of structures addressed in this section. 10.1 Structural Foundation /Basemat 10.1.1 Description The Basemat is a 10 foot thick reinforcui concrete slab that supports the Nuclear Island structures. The Basemat measures 334 feet by 442 feet, which includes an extension of four feet beyond the Nuclear Island perimeter along all four sides to allow for one method for detailing of reinforcing at the edge of the basemat. The four foot extension of the basemat is not credited in any analyses. Alternate design details that meet the ACI Code requirements may be used provided that the as-built design details are documented in the structural analysis report described in the structural acceptance criteria, Section 3.8.4.5.4. Typical reinforcing details for alternate designs are shown in Appendix 3.8B, Figures 3.8B-3 and 3.8B4. 10.1.2 Design Requirements The basemat is designed for the envelope of reactions considering all soil cases. The basemat analysis provides support reactions assuining a homogeneous foundation subgrade. These reactions are used to determine an effective soil bearing pressure under the basemat. Reactions are represented by vertical soil springs. Spring constants are calculated based upon contributory areas and the underlying soil stiffness. The basemat shall use a symmetrical reinforcing configuration based on the maximum required reinforcing, either top or bottom of the basemat to account for differential settlement. Design of the basema shall consider stresses due to pouring sequence of the mat as well as the erection sequence for components located above the mat. Pour layouts should minimize skewed intersection of constructionjoints with walls due to conflicts in placement of wall dowels. Approwd Desigrr Material. Desiger of SSC Page 3.8A-32
+
i ^ Sv tem 80 +' Desian caneraloocanent 7 ); . Concrete pours shall reg. _re engineered constructionjoints detailed on concrete and reinforcing design !' 2 drawings. Details shal' Allow for proper spacing and stagger of individual rebar splices and shear
; ' reinforcing required by .,CI 349 Sections 12.14 through 12.17. Design of the construction joint shall .
j consider the requirements for additional shear reinforcing identified in ACI 349 Section 11.7. l Shrinkage cracks in the exposed vertical faces of concrete. pours shall be controlled by minimum > I
- 1. reinforcing as specified in ACI 349 Section 7.12. This reinforcing shall apply to temporarily exposed i faces of interior construction joints. ;
t
- Design of the basemat includes blockouts needed for equipment sumps. At these sump locations, basemat thickness is reduced. Additional horizontal reinforcing shall be added in the sump sidewalls to j' eccamma4e basemat design moments. ;
I The basemat shall be founded on competent structural backfill. The backfill material shall meet the , l " requirements of the Unified Soil Classification S / stem for SP, SM, GP, or GW soils, except that the , maximum percentage of soil passing through No. 200 sieve shall be no greater than 12 percent. The soil ! shall be compacted to be a minimum of 95 percent as determined by ASTM D1557. l l
'10.1.3 Design Loads I Refer to Table 3.8 A-1 of this appendix for additional Basemat design loads.
10.2 Containment Shield Building ; ( 10.2.1 Description l l The Containment Shield Building is the concrete structure that surrou.xis the steel Containment Vessel l I and Containment Subsphere and provides protection from postulated external missiles and environmental effects. The Containment Shield Building provides an additional barrier against the release of fission l products. ; l The Shield Building consists of a cylindrical 4 feet thick reinforced concrete shell wall with a 105' inside j
- radius extending from the foundation basemat at El. 50'-0" to El.146'-0". The cylindrical wall extends :
l upward from El.146'-0" with a 3 ft thickness to the spring line at El.157'-0". The Shield Building is ! , topped by a 3 ft thick reinforced concrete hemispherical roof. The outside apex of the dome is at i . elevation 265'-0". j
; 1 The Unit Vent is a Seismic Category Il structure attached to the exterior of the Containment Shield l Building. The Unit Vent is designed for the SSE using Seismic Category I criteria.
I ) 10.2.2 Design Requireements The Containment Shield Building penetrations shall be sealed to maintain the annulus ventilation boundary. 1 I Mwoud Denipn nonauw.Dee4pn er ssc page 2.sa.22 1
i l System 80+ Design Con %of Document 10.2.3 Design Loads (Reference Section 3.8.4.3) ASCE Paper 4933 applies to wind loads on the Containment Shield Building. For tornado winds, the external pressure shape coefficient (Cp ) used in the formula in Section 5.1.3.2.2 of this appendix is taken from the tables in ASCE 7-88. The wind load distribution curves for the Containment Shield Building are in Section 3.3. Refer to Table 3.8A-1 for additional design loads applicable to the Containment Shield Building. 10.3 Reactor Building Subsphere The Reactor Building Subsphere, located inside the Shield Building and external to the Containment Vessel, consists of reinforced concrete walls and slabs and the Containment Support Pedestal. The purpose of the subsphere structures is to support the Containment Vessel and the Internal Structures and isolate safety related equipment. Refer to Table 3.8A-1 for Reactor Building Subsphere general design loads. 10.3.1 Containment Support Pedestal 10.3.1.1 Description The Containment Suppon Pedestal is the intermediate concrete suppon between the Containment Vessel and the Nuclear Island Foundation Basemat. The containment suppon pedestal has no pressure retaining function and therefore is not designed per ASME Code. The pedestal is comprised of a circular columnar suppon 66 feet in diameter, with an additional area extending out 8'-7" under the Upper Guide Structure Laydown Area, and a 3 ft thick curved dish pedestal. The center column extends from El. 50'-0" to the Containment Vessel inven at El. 57'-0". The Dish Pedestal extends around the containment vessel from the center column upward to El. 91'-9" 10.3.1.2 Design Requirements Resistance for the Containment Vessel against sliding and ovenurning on the Containment suppon pedestal is provided by shear connectors welded to the Containment Vessel. Compressible material is provided around the upper edge of the dish pedestal dish to reduce bearing stresses between the dish and the Containment Vessel at El. 91'-9". Design details shall allow for insertion of the compressible material and containment inspectability. Preventive measures are required in this bearing area to reduce or prevent containment corrosion. These measures include:
- Scaling of the concrete to keep out moisture
- Use of sloped floors and drains to prevent collection of surface water in the transition area Containment penetrations that pass through the suppon pedestal concrete must allow for inspection and testing at the Containment Vessel in compliance with General Design Criterion 53. Provisions for inspection and testing must be included in the design.
AMmmf Design htwia!- Design of SSC Pope 3.8A-34
J System 80+ _ Deslan controlDocument 10.3.1.3 Design Inada (Reference Section 3.8.4.3)
- Refer to Table 3.8A-1 for additional design loads applicable to the Containment Support Pedestal.
l 10.4 Conf alanwnt Internal Structures i The Containment Internal Structures are located inside the spherical steel containment vessel. The , purpose of these internal structures is to provide structural support, radiation and missile shielding, and : space for the IRWST. Thece structures are constructed of reinforced concrete and structural steel. These structures are described in Section 3.8.3.1. Refer to Table 3.8A-1 for general design loads for the Containment Internal %tructures. l
; 10.4.1 Pear *ar Vessel nW Shield Wall
.10.4.1.1 Description _
J , The Primary Shield Wall (PSW) is a reinforced concrete enclosure that surrounds the Reactor Vessel. ; $ ' The Primary Shield Wall is a minimum of six feet thick. [ 10.4.1.2 Design Requir==an*= ) The Primary Shield Wall (PSW) provides protection for the vessel from internal missiles. The Prunary ) j f Shield Wall provides biological shielding and is designed to withstand the temperatures and pressures following LOCA. In addition, the primary shield wall provides structural support for the Reactor Vessel ; j (Reference Section 3.8.3.1). i 10.4.1.3 Design Loads (Reference Section 3.8.3.3) i . The Primary Shield Wall shall be designed for normal dead loads as well as equipment live loads and l telated seismic forces. In addition the PSW shall resist the dynamic loads due to the NSSS components. , l The inner face of the lower Prunary Shield Wall will be provided with projecting reinforced concrete corbels to be used as the support bases for the Reactor Vessel steel support Columns. Corbels shall i have symmetrical reinforcing in the top and bottom to resist the upward loads resulting from a potential ex-vessel steam explosion (Section 3.8.3.3). ~ Refer to Table 3.8A-1 for additional design loads that are applicable to the Pnmary Shield Wall. I' 10.4.2 Crane Wall (Secondary Shield Wall) i 10.4.2.1 Description
~ The Crane Wall is a reinforced concrete right cylinder with an inside diameter of 130 feet and height of 118'-3" from its base. The top elevation is at El. 210*-0". The Crane Wall is a minimum of four feet thick.
4 1 O 4pmeer Demon annearnet Denten er ssc rege 2.sa-25 v
System 80+ Design Control Document 10.4.2.1 Design Requirements The Crane Wall provides supports for the polar crane and protects the steel containment vessel from internal missiles. In addition to providing biological shielding for the coolant loop and equipment, the Crane Wall also provides structural support for pipe supports / restraints and platforms at various levels. The design shall address the vertical alignment of the Crane Wall with the corresponding structure below the Containment Vessel and provides special construction tolerances, as necessary, to ensure potential misalignment is appropriately considered. The design also considers potential differential basemat settlement and the effect on the Crane Wall alignment. 10.4.2.3 Design Loads (Reference Section 3.8.3.3) l Refer to Table 3.8A-1 for additional loads that are applicable to the Crane Wall. 10.4.3 Refueling Cavity 10.4.3.1 Description The Refueling Cavity is the reinforced concrete enclosure that provides a pool filled with borated water above the reactor vessel to facilitate the fuel handling operation without exceeding the acceptable level of radiation inside the Containment Vessel. The Refueling Cavity has the following sub-compartments. e Storage Area for Upper Guide Structure e Storage area for Core Support Barrel e Refueling Canal The Reactor Vessel flange is sealed to the bottom of the Refueling Cavity to prevent leakage of refueling water into the reactor cavity. The Fuel Transfer Tube connects the Refueling Cavity to the Spent Fuel Pool. The shield walls that form the Refueling Cavity are a minimum of six feet thick. 10.4.3.2 Design Requirements I The Refueling Cavity walls and floor shall be covered with stainless : teel plate for leak tightness and for contamination and corrosion control. 10.4.3.3 Design Loads (Reference Section 3.8.3.3) ) Refer to Table 3.8A-1 for additional design loads that are applicable to the Refueling Cavity. 10.4.4 Operating Floor 10.4.4.1 Description The Operating Floor at El.146'-0" provides access for operating personnel functions and provides biological shielding. Inside the Crane Wall, the operating floor is a reinforced concrete slab with a covered hatch that is aligned with hatches in the two lower floors. Outside the Crane Wall, the Operating Floor consists of steel grating. Asnproved Dessgre Motonial- Desigru of SSC (2/95) Page 3.8A 36
____ . __ _ ~. __ _ _ _ _ _ _. _ . _ _ _ _ _ _ _ . . _ . __ _ . _ . _ . _ . _ ._ _ i i a Sv tem 80+ oesion controlDocanent ;
'10.4.4.2- Design Imula (Refenace Section 3.8.3.3)
Refer to Table 3.8A-1 for additional design loads that are applicable to the Operating Floor. !
' 10.4.5 In-containneent Refuehng Water Storage Tank j P
, 10.4.5.1 Description , The IRWST provides storage of refueling water, a single source of water for the Safety Injection and i Containment Spray pumps and a heat sink for the Safety Depressunzation System. The IRWST is dishlike in shape and utilizes the lower section of the Internal Structures as its outer boundary. 10.4.5.2 Design Requirements The IRWST is provided with a stainless steel liner to prevent leakage. . Design of the IRWST considers pressurization as a result of the containment systems Design Basis Accident. 10.4.5.3 Design Loads (Reference Section 3.8.3.3) l Refer to Table 3.8A-1 for additional design loads applicable to the IRWST. 10.4.6 ' Lower Conente Dish 10.4.6.1 Description The lower concrete dish, or the Containment concrete base is the base support for all of the Reactor Building internal structures and NSSS components. The dish is comprised of a segment of a sphere from the containment invert up to the reactor cavity floor at elevation 62'-0", and a three feet thick inner liner to the Containment up to El. 91'-9". The reactor cavity floor is provided with a sump within the lowe- dish concrete. The lower concrete dish transfers loads via direct bearing and shear connectors to the Containment Vessel through to the containment support pedestal. 10.4.6.2 Design Requirements Resistance for the lower concrete dish against sliding and overturning is provided by shear connectors welded to the Containment Vessel. 10.4.6.3 Design Loads (Reference Section 3.8.3.3) Refer to Table 3.8A-1 for additional design loads applicable to the Lower Concrete Dish. 10.5 Nuclear Annex
~ The Nuclear Annex is composed of the Control Complex, Diesel Generator Areas. Main Steam Valve House Areas. CVCS and Maintenance Areas, and Fuel Handling Area.
The Nuclear Annex is a reinforced concrete structure composed of rectangular walls, columns, beams,
- and floor slabs. The. Nuclear Annex shares common walls and foundation basemat with and is 4prevar Deewr neesenw
- Deegn et SSC Page 3.8A-37
l Design ControlDocument yem 80+ monolithically connected to the Containment Shield Building. In addition to these structural components, there are components designed to provide biological shielding and protection against tornado and turbine ! missiles. Structural components, as well as members serving as shielding components, vary in thickness j from approximately one foot to five feet. l 10.5.1 General Design Requirements Exterior walls shall be designed to withstand the soil loads due to the normal finished yard grade. In ; addition the walls shall be designed for surcharge loads due to adjacent structures and/or temporary j construction or maintenance loads. Dynamic loads due to seismic soil structure interaction shall be included with applicable factored load combinations. The exterior walls and roof slabs provide protection for the interior of the Nuclear Annex against environmental loads. Refer to Table 3.8A-1 for general design loads applicable to the Nuclear Annex. 10.5.2 Diesel Generator Areas 10.5.2.1 Description The Diesel Generator Areas provide protection to two diesel generators installed in separate compartments ' located on opposite sides of the Nuclear Annex. 10.5.2.2 Design Loads (Reference Section 3.8.4.3) These components shall be designed for the general requirements given in Section 10.5.1 of this appendix and the additional loads given Table 3.8A-1. 10.5.3 Control Complex 10.5.3.1 Description The Control Complex consists of the Vital Instrument & Equipment Rooms at El. 50* +0" and those areas located above them. The Control Complex provides two physically separate divisions for electrical distribution. control, and instrumentation systems leading to the Control Room. 10.5.3.2 Design Requirements The upper floor of the Control Complex contains the Control Room which shall be designed to provide secu;h he, and environmental protection to the control equipment and the Control Room operators. 10CFR Chapter i Part 73 Section 55(c)(6) specifies that walls, doors, ceiling, and floor of the Control Room shall be bullet-resisting. 10.5.3.3 Design Loads (Reference Section 3.8.4.3) Refer to Table 3.8A-1 for additional design loads applicable to the Control Complex. O Approved Design Meterial . Deupre of SSC Pope 3.8A.38
Syntem 80+ Design ControlDocument p t ,g .10.5.4 - Main Steam Valve House 10.5.4.1 Description The Main Steam Valve House is a compartment located above the EFW Tank Areas on the north and south sides of the Nuclear Annex. The compartment floor elevation is El.106'-0". The Nuclear Annex roof at El.156*-0" is the top of the compartment. 10.5.4.2 Design Requirements ; The Main Steam Valve House shall be designed to provide environmental protection, primarily missile l protection, for the Main Steam and Feedwater Line safety related valves and piping. The Valve House , also provides protection to the Nuclear Annex penetrations through the inside walls of the Valve House. 10.5.4.3 Design Loads (Reference Section 3.8.4.3)
..In addition to the applicable design loads given in Table 3.8A-1, the walls of the Valve House shall be design to resist loads due to potential pipe rupture loads from the Main Steam and Feedwater Lines.
10.5.5 Fuel Handhng Ana 10.5.5.1 Description
./G The Fuel Handling Area includes the Spent Fuel Pool, Refueling Canal, Cask Laydown and Washdown b Areas, truck / rail shipping bay, and New Fuel Storage Area. l The spent fuel pool is an open stainless steel lined reinforced concrete vessel used for submerged storage of radioactive spent fuel assemblies. The pool is approximately 32'-6" by 43' with a depth of 42'. The walls and floor of the spent fuel pool are a minimum of 6' thick.
Fuel assemblies are transferred from the Fuel Pool to the Refueling Cavity via the Refueling Canal at the end of the Fuel Pool and the Fuel Transfer Tube through the Shield Building and the Steel Containment. The Refueling Canal measures 6 feet wide by 49'-5" long. The minimum wall thickness, on the fuel pool side, is 6'. An opening in the fuel pool wall allows for passage of fuel between the Fuel Pool and the Refueling Canal. A steel divider is provided for the opening. 10.5.5.2 Design Requirements Seals are incorporated to allow draining of the refueling canal while maintaining the water level in the spent fuel pool. The fuel pool liner plate is designed for impact loads due to dropped fuel assemblies. 10.5.5.3 Design Loads (Reference Section 3.8.4.3) An overhead bridge crane with a capacity of 150 tons must be provided over the shipping bay and i extending over the fuel pool and refueling canal. L :: Design neennial- Design of SSC Page 3.8A-39 j l
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SWtem 80+ Design Contro! Document The maximum water depth in the Spent Fuel Pool is 40'. Refer to Table 3.8A-1 for additional design loads applicable to the Fuel Handling Area. 10.5.6 EFW Tank Areas 10.5.6.1 Description The two EFW tanks consist of paired stainless steel lined reinforced concrete rooms. Each pair of rooms comprise a single tank adjacent to each Diesel Generator Area. The tanks extend from EL. 70'.0" to the underside of the floor slab at El.106'-0". . 10.5.6.2 Design Loads Refer to Table 3.8A-1 for additional design loads applicable to the EFW Tank Areas. 10.5.7 CVCS and Maintenance Areas 10.5.7.1 Description The CVCS Area consists of a number of smaller rooms used to isolate components for water treatment required by operating systems. Individual rooms are required for radiation shielding. An underground pipe chase through the wall along Column Line "W" is provided to tie in components in the CVCS Area to the Radwaste Building. Other areas at El. 91'-9" are designated for equipment decontamination and El 146'-0" for personnel decontamination. A rail / truck shipping bay is provided for material deliveries for the CVCS area and shipments involving access to the Reactor Building Equipment Hatch. 10.5.7.2 Design Loads (Reference Section 3.8.4.3) A 225 ton overhead bridge crane must be provided over the shipping bay. Refer to Table 3.8A-1 for additional design loads applicable to the CVCS Area. 11.0 Supplemental Design Criteria for Non-nuclear Island, Seismic Category I and II Structures 11.1 Diesel Fuel Storage Structure 11.1.1 Building Classification
- Quality Class 1 e Safety Class 3 e Seismic Category i O
Approved Design Material- Design of SSC Pope 3.8A40
i i ! System 80 + - Denian Coneof Doewnent l 11.1.2 Description j There are two Diesel Fuel Storage Structures; one on each side of the Nuclear Island. 1 The main reinforced concrete structure is approximately 25 ft high,63 ft long and 44 ft wide founded J on a 2' - 3" thick reinforced concrete mat located 12'-6" below the grade elevation of 90'-9". The walls and the roof are 2' - 3" thick. There is a two foot thick center reinforced concrete wall that divides the ; i structure into two separate bays. Each bay encloses a diesel fuel oil tank, a tank vent, a sump with a , sump pump, and a*ca-ry piping. The bays are separated from each other and from the equipment room , 4
- by three-hour rated fire barriers (i.e.,2 ft thick walls). A steel platform at elevation 89'-3" surrounos ;
- each of the fuel tanks. 'Ine outside doors are protected against tornado missiles by a concrete missile j barrier. .
i There is also an attached outside Seismic Category II equipment room that is approximately 10 ft high, ; 12 ft long and 28 ft wide fotended on a 15" reinforced concrete mat. The equipment room is a steel , framed structure with insulatec' metal siding and a metal deck roof. The Diesel Fuel Storage Structure shall be located a minimum of 50 feet from any hydrogen storage area j to preclude loading to the structure from a potential hydrogen burn. .i 11.1.3 Elevations l
- El. 78'-3" Bottom of base mat for the main structure l t
e El. 91'-9" Top of base mat for the equipment room structure . e- El. 91*-9" Top of steel platform i e El.103'-3" Top of roof ! 11.1.4 Codes and Senadards l The codes and standards applicable to Seismic Category I buildings shall be met for the Diesel Fuel ! l Storage Structure including the equipment room. ! t . 1 11.1.5 1nada
- in addition to the minimum design loads requirements of Section 5.1 of this appendix, the following i additional specific load requirements shall be met. Should conflicting values occur between this section .!
and Section 5.1 of this appendix, the values specified in this section apply. ) l 11.1.5.1 Dead Imd (D) The foundation slab shall be designed to include the reactions imparted by the steel fuel tank support frames. The weight of each tank and oil is approximately 402 kips. (The site specific SAR shall verify
; the tank volume is adequate for the diesel generators purchased, such that they meet their design criteria.)
The tank support frame is not covered by this criteria and shall be designed in accordance with the rules of Reference ASME Section III, Division 1, Subsection NF. l i Areemer aeno,, aneee,w anoon er ssc - tmst rene 2.su1 4
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System 80+ Design ControlDocument 11.1.5.2 Live Load (L) The Diesel Fuel Storage Structure shall be designed for the following floor live load. 11g31 Live lead
- Basemat Floor 250 psf
- Steel Platform 150 psf
- Roof 100 psf 11.1.5.3 Temperature Loads (T.)
The normal concrete surface operating temperature within the building ranges from 60'F to 90'F. The ambient temperature range outside of the building shall be -10'F to 100*F (See Section 5.1.1.5 of this appendix). Site specific provisions may be taken to minimize the effects of the structural temperature gradient produced by these conditions. 11.1.5.4 Seismic Loads (E') The seismic accelerations shall be as specified in Table 3.8A-2. 11.1.5.5 Oil Leakage All building walls shall be designed to contain the contents of the 45,000 gallon oil tanks in the event one tank fails. 11.1.5.6 Other Loads All abnormal loads (i.e., P., T., R., Y;. Y, and Y,) are zero. 11.1.6 Loading Combinations and Acceptance Criteria 11.1.6.1 Concrete The requirements of Section 5.2.2 of this appendix shall be met. 11.1.6.2 Stability l The requirements of Section 5.2.4 of this appendix shall be met. ) 11.1.7 Other Requirements The building is to be founded on competent structural backfill as defined in Sect.,,a 10.1 of this appendix. The bearing pressure shall not exceed the allowable value given in Table 2.0-1. O , howmd onion wtww. onion or ssc rage 3.sw
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. System 80 + Declan controlDocument ~
i 11.2 Component Cooling Water Heat Ewh==er Structure .
')
;11.2.1 Building Classification l
'* Quality Class 1
*7 Safety. Class 3 1
.* Seismic Category 1 ]
11.2.2 Description i There are two Component Cooling Water (CCW) Heat Exchanger Structures, each structure houses two . heat exchangers. The CCW system is a redundant system with only two heat exchangers required for d plant operation. The first floor houses the heat exchanger, while the basemat levels contains piping and i _ equipment. I
' Each structure is a' two story reinforced concrete structure approximately 34 ft high, from the top of the j mat,110 ft long, and 44 ft wide founded on a four foot thick reinforced concrete mat located 17' - 0" L below grade. The walls are 2' - 3" thick and the roof is two feet thick. The first floor of the structure <
i is three feet thick and is supported by three rows of columns approximately twenty two feet on center i with the two outer rows located directly under the two heat exchangers. The center row of these columns is continued through the first floor to provide additional support for the roof. ; V The roof supports two fan rooms on one end of the building and two air inlet rooms on the. opposite end of the building. Both of these rooms extend the width of the building and are approximately 23 feet wide with a partially open face covered with a bird screen. A concrete overhang is provided and serves as a .
- missile barrier for the open face.
The outside doors are protected against tornado missiles by concrete missile barriers. CCW heat exchanger maintenance sumps are located in the basemat at one end of the structure. The ' sump has a capacity equal to the fluid contents of the shell inside of one heat exchanger. There are floor drain sumps located at the opposite end of the structure. : 4 The CCW Heat Exchanger Structures shall be located a minimum of 50 feet away from any hydrogen storage area to preclude loading to the structure from a potential hydrogen burn. .; i: An underground tunnel is connected to each CCW Heat Exchanger Structure from the Nuclear Annex ! for the CCW piping. The top of the tunnels basemat is at the same elevation as the top of the CCW Heat Exchanger Structure basemat. . 11.2.3- Elevations
- El.121'-9" Top of roof of fan / air filter room 4
- El. I1l'-9" Top of Roof
_ , . _ . _ _ = _ _ _
System 80+ Design ControlDocument
- El. 91*-9" Top of the first floor (1 foot above grade)
- El. 73'-9" Pnttom of basemat 11.2.4 Codes and Standards The codes and standards applicable to Seismic Category I buildings shall be met.
11.2.5 Imads In addition to the minimum design loads requirements of Section 5.1 of this appendix, the following additional specific load requirements shall be met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply. 11.2.5.1 Dead lead (P) The weight of each heat exchanger when full of water is approximately 250 Kips excluding the heat exchanger saddle and leg supports. The heat exchanger support is not covered by this criteria and shall be designed in accordance with the rules of ASME Bc iter and Pressure Vessel Code, Section III, Division l 1, Subsection NF. 11.2.5.2 Live Load (L) The CCW Heat Exchanger Structure shall be designed for the following live loads. Itegl Live Load e Fan and Air Inlet Room 150 psf
- Roof 100 psf
- First floor 150 psf
- Basemat 250 psf 11.2.5.3 Temperature leads (T )
The normal concrete surface operating temperature within the building ranges from 60*F to 90*F. The ambient temperature range outside of the building shall be assumed to range from -10*F to 100*F (See Section 5.1.1.5 of this appendix). Site specific provisions may be taken to minimize the effects of th; structural temperature gradient produced by these conditions. 11.2.5.4 Seismic Loads (E') The seismic accelerations shall be as specified in the Table 3.8A-3. 11.2.5.5 Internal Flooding The structure sump is designed to collect water due to flooding resulting from a potential rupture of the CCW or Station Service Water (SSW) piping. ANvoved Design Materin!- Design of SSC (2/9 51 Page 3.8AM
System 80+ oesign contrat Document i O) N,_/ 11.2.5.6 Other Loads All abnormal loads (i.e., P., T., R,, Y), Y, and Y,) are zero. I 11.2.6 Loading Combinations and Acceptance Criteria 1 11.2.6.1 Concrete The requirements of Section 5.2.2 and 8.0 of this appendix shall be met. 11.2.6.2 Structural Steel The requirements of Section 5.2.3 and 8.0 of this appendix shall be met. l 11.2.6.3 Stability j The requirements of Section 5.2.4 of this appendix shall be met. ) 1 11.2.7 Other Requirements The building is to be founded on competent structural backfill as defined in Section 10.1 of this appendix. l l The bearing pressure shall not exceed the allowable value given in Table 2.0-1. 1 . ,A 11.3 Radwaste Facility 1 11.3.1 Bi.ilding Classification j e Quality Class 2 e Safety Class NNS J l e Seismic Category II 11.3.2 Description The Radwaste Facility is a non-safety related reinforced concrete building located adjacent to and on the west side of the Nuclear Annex. The building houses the liquid and solid radioactive waste management systems. { The building is a four story L shaped reinforced concrete stmeture with a thick stepped mat foundation 3 I with the major dimensions of the L being approximately 167 ft long and 153 ft wide. The major floors are at elevations 115'-6",91'-9",70'-0" and 50'-0". The elevations at the top of the stepped mat are 50'-0" and 34'-0". The basement and the first two floors are a labyrinth of walls that create numerous compartments utilized for radwaste management system components. There is a truck bay located at elevation 91'-9". A bridge
, crane, supported just below the roof, on the west end of the building traverses the entire width of the building in the north-south direction. The area serviced by the crane is open to elevation 91'-9".
(v) Arweeed Onekn Meential. Design of SSC Pepe 3.8A-45
System 80+ Design controlDocument 11.3.3 Elevations
- El. 135'-6" Top of roof
- El. 115'-6" Top of the third floor
- El. 91'-9" Top of the second floor (Grade is at El. 90'-9")
- El. 70'-0" Top of the first floor
- El. 50'-0" Top of the basemat on the North side
- El. 34'-0" Top of the basemat on the South side 11.3.4 Codes and Standards The codes and standards applicable to Seismic Category II buildings shall be met.
11.3.5 leads l l In addition to the minimum design loads requirements of Section 5.1 of this appendix, the following l additional specific load requirements shall be met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply. 11.3.5.1 Dead Load (D) The weights for major equipment are listed in Table 3.8A-4. 11.3.5.2 Live Load (L) The live loads are listed in Table 3.8A-5. 11.3.5.3 Temperature Loads (T) The normal concrete surface operating temperature within the building ranges from 60*F to 90*F. The ambient temperature range outside of the building shall be -10*F to 100*F (Section 5.1.1.5 of this appendix). Site specific provisions may be taken to minimize the effects of the structural temperature gradient produced by these conditions. 11.3.5.4 Seismic Loads (E') The seismic accelerations shall be as specified in the Table 3.8A-6. 11.3.5.5 Internal Flooding The foundation and walls shall be designed to include the containment of the maximum inventory of the solid and liquid wasic management systems. Ammt oesw unww- cesw or ssc roon 2.sA-os
System 80+ Design controlDocument l
, 11.3.5.6 Crane Loads v
The main bridge crane at grade shall be designed for 15 tons. A bridge crane above the HIC storage resin dewatering area shall be designed for 5 tons. All monorails shall be designed for 5 tons. The crane haunches shall be designed in accordance with Section 5.1.1.2.5 of this appendix. 11.3.5.7 Other Loads All abnormal loads (i.e., P., T., R,, Yj Y, and Y,) are zero. 11.3.6 Le=* 2 Combinations and Acceptance Criteria 11.3.6.1 Concrete , The building shall be designed for the SSE using Seismic Category I criteria. The requirements of Section 5.2.2 and 8.0 of this appendix shall be met. l 11.3.6.2 Structural Steel , The building shall be designed for the SSE using Seismic Category I criteria. The requirements of Section 5.2.3 and 8.0 of this pendix shall be met. , 11.3.6.3 Stability The requirements of Section 5.2.4 of this appendix shall be met. ) 11.3.7 Other Requirements There shall be a minimum space of six inches between the Radwaste facility and the Nuclear Island to : prevent seismic interaction. The building is to be founded on competent structural backfidi as defined in Section 10.1 of this appendix. The bearing pressure shall not exceed the allowable value given in Table 2.0-1. 11.4 Service Water Pumphouse and Intake Structure 11.4.1 Building Classification
- Quality Class 1
- Safety Class 3 l
- Seismic Category 1 l 11.4.2 Description 4 The Service Water Pump Structure is classified Category I and is not included in the scope of design g certification due to its specific site design requirements. The building includes a mat type foundation and a reinforced concrete superstructure with rigid walls. The service water pump room and its supporting elements will be protected against flooding.
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Sy^ tem 80+ Design ControlDocument 11.4.3 Elevations Service water pump structure elevations are site specific. 11.4.4 Codes and Standards The codes and standards applicable to Ceismic Category I buildmgs shall be met. 11.4.5 Loads in addition to the minimum design loads requirements of Section 5.1 of this appendix, the following specific additional load requirements shall be met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply. 11.4.5.1 Dead Load (D) The weight of each Station Service Water (SSW) pump is dependent on site specific considerations. 11.4.5.2 Live Load (L) The SSW pump supports are designed for thrust loads per vendor drawings. Run Live Load Concrete Floors (by COL) psf l Roof (by COL) psf Operating Conditions: Normal water level El. (Note 1)
- Extreme low water level El. (Note 1)
Maximum water level (flood) El. (Note 1) Note 1: These elevations will be established based on site specific data. 11.4.5.3 Temperature Loads (T,) The normal concrete surface operating temperature within the building is site specific. The ambient temperature range outside of the building shall be -10'F to + 100'F (Section 5.1.1.5 of this appendix). Site specific provisions may be taken to minimize the effects of the structural temperature gradient produced by these conditions. 11.4.5.4 Seismic Loads (E') The seismic response of the structure is site specific. 11.4.5.5 Flooding Flood loads on the Service Water Pump Structure shall include internal flooding and hurricane induced wave forces. Approved Desips hinterW . Desips of SSC 12/9 5) Page 3.8A.48
i System 80+ Design ControlDocument '(o) V 11.4.5.6 ' Other Loads All abnormal loads (i.e., P., T,, R , Y). Y, and Y,) are zero. 11.4.6 Loading Combinations and Acceptance Criteria 11.4.6.1 Concrete The requirements of Section 5.2.2 and 8.0 of this appendix shall be met. 11.4.6.2 Stability The requirements of Section 5.2.4 of this appendix shall be met. 11.4.7 Other Requirements Site Specific 11.5 Turbir.e Building 11.5.1 Building Clas-ification
- Quality Class 2 b o Safety Class NNS
- Seismic Category II 11,5.2 Description The Turbine Building is located adjacent to and on the cast side of the Nuclear Annex. The Turbine Building is approximately 200 ft by 370 ft, has a ground floor, a mezzanine floor, an operating floor and a roof that has several different elevations.
The ground floor is a reinforced concrete slab. In the area of the condensers the foundation is comprised of a stepped mat. The three turbines are founded on a reinforced concrete slab that is supported by pedestals that extend down to the basemat. The outside wall above grade is a steel framed superstructure with metal siding. The major portion of the roof spans approximately 135' and is comprised of prefabricated trusses with built-up roofing consisting of metal decking. There is a 125 ton main crane and a 25 ton auxiliary crane that traverses the length of the building. The cranes are supported by the outside steel columns. Railroad service is provided at the east end of the building with the track running threugh the inside of
, the building in the north-south direction.
A] w w noneeniet. Demon et ssc rege2.sA-es
System 80+ Design ControlDocument 11.5.3 Elevations Turbine Building elevations are reference elevations only. 11.5.4 Codes and Standards The codes and standards applicable to Seismic Category 11 buildings shall be met. 11.5.5 Loads in addition to the minimum design loads requirements of Section 5.1 of this appendix, the following additional specific load requirements shall be met. Should conflicting values occur between this sectio i and Section 5.1 of this appendix, the values specified in this section apply. 11.5.5.1 Dead Load (D) The estimated weights for major equipment are listed in Table 3.8A 7. 11.5.5.2 Live Load (L) The live loads are specified in Table 3.8A-8. 11.5.5.3 Temperature Loads (T,) The normal operating temperature within the building ranges from 40*F to 100*F. The ambient temperature range outside of the building shall be -10*F to 100*F (Section 5.1.1.5 of this appendix). 11.5.5.4 Seismic Loads (E') The seismic accelerations shall be as specified in the Table 3.8A-9. 11.5.5.5 Pipe Loads Where the piping loads are not known at the time of design, beams and girders are designed for a concentrated load applied at midspan as indicated below.
- 1. In areas where the main steam and steam generator feedwater lines are located, use the weight of the lines full of water.
- 2. In areas where large bore piping is heavily concentrated:
Girders (column to column) 55 kips Primary beams (column to column) 45 kips Secondary beams 30 kips For the design of the columns, a load of 110 kips applied at the operating floor level. O Approved Design Meterial Design of SSC Page 3.8A.50
Sv tem 80+ Design ControlDocument n i 1 3. For all other areas not included above:
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Girders (column to column) 30 kips l Primary beams (column to column) 20 kips , Secondary beams 10 kips [ 11.5.5.6 Trip-out end' Thermal Ledag Trip-out and thermal loadings are provided by the turbine vendor. 11.5.5.7 Equipment Laydown Loads All floor laydown areas are designed for equipment laydown loads obtained from the equipment manufacturer. Dismantling and equipment laydown areas designed to carry the above loads are indicated on the drawings. 11.5.5.8 Pedestal Design Since the pedestals will vibrate at the forcing function, namely the RPM of the turbine, due consideration shall be given to the pedestal frequency to avoid a resonance condition. Criteria for the relative displacement for the turbine and generator shaft supports are provide by the turbine manufacturer. 11.5.5.9 Transmission Line Provision is made on the east wall of the building for pulling transmission lines. ; 11.5.5.10 Rail Loads Design of the rail bay and foundation shall include loads from trains with the heaviest equipment transported by rail. 11.5.5.11 Other Loads Other loads to be considered include reactions due to the circulating water lines, machine unbalance load, , thermal expansion of the equipment, normal unbalance of the rotating equipment, condenser vacuum and emergency loads such as short circuit torque, broken rotor blade, and bowing of the rotor. This data is provided by the turbine manufacturer. 11.5.6 Loading Combinations and Acceptwee Criteria 11.5.6.1 Concrete The concrete portion of the building shall be designed using Seismic Category I c;iteria. The requirements of Sections 5.2.2 and 8.0 of this appendix shall be met. (Q () 4eroved Doeten neenerw Dee w of SSC Page 38A 51 i
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System 80+ Design ControlDocument 11.8.6.2 Structural Steel The Turbine Building lateral resisting steel frame shall be designed for the SSE using Seismic Category I criteria. The requirements of Sections 5.2.3 and 8.0 of this appendix shall be met. 11.5.6.3 Stability The requirements of Fection 5.2.4 of this appendix shall be met. 11.5.7 Other Requirements There shall be a minimum space of six inches between the Turbine Building and the Nuclear Island to prevent seismic interaction. Only one emergency load shall be assumed to act at one given time. Emergency load factors and additional load combinations are to be considered. The building is founded on competent structural backfill as defined in Section 10.1 of this appendix. The bearing pressure shall not exceed the allowable value in Table 2.0-1. 11.6 Dike for Outdoor Tanks 11.6.1 Building Classification
- Quality Class 3
- Safety Class NNS
- Seismic Category 11 11.6.2 Description Two foot reinforced concrete dikes will surround the CVCS outdoor tanks and the Condensate Storage Tank. The wall height of the dike will be approximately six feet and the plan dimensions will be determined from the amount of liquid the dike inust contain.
j 11.6.3 Elevations l Site Specific 11.6.4 Codes and Standards l The applicable codes and standards applicable to Seismic Category I buildings shall be met. l l 11.6.5 Loads , in addition to the minimum design loads requirements of Section 5.1 of this appendix, the following I specific additional load requirements shall be met. Should conflicting values occur between this Section and Section 5.1 of this appendix, the values specified in this Section apply. l &M Des $pn Material. Design of SSC Pope 3.8A 52
i System 80+ Design ControlDocument i 11.6.5.1 Temperature Loads (T ) The ambient outside temperature range shall be -10*F to 100'F (Section 5.1.1.5 of this appendix). 11.6.5.2 Seismic Loads (E') The seismic accelerations shall be as specified in the Table 3.8A-10. 11.6.5.3 Water Slosh The affects of water slosh loads, including the wall flexibility shall be considered in accordance with Section 5.1.3.1 of this appendix. 11.6.5.4 Other Loads All abnormal loads (i.e., P,,1 ?,, Y;, Y,, and Y,) are zero. 11.6.6 Loading Combinations and Acceptance Criteria 11.6.6.1 Concrete The dikes shall be designed using Seismic Category I criteria.. The requirements of Sections 5.2.2 and 8.0 of this appendix shall be met. 11.6.6.2 Stability The requirements of Section 5.2.4 of this appendix shall be met. 11.6.7 Other Require:nents , The dike is to be founded on competent structural backfill as defined in Section 10.1 of this appendix. The bearing pressure shall not exceed the allowable value in Table 2.0-1. 11.7 Component Cooling Water Tunnel 11.7.1 Building Classification
- Quality Class 1
- Safety Class 3
- Seismic Category I 11.7.2 Description There are two component cooling water tunnels that connect the CCW Heat Exchanger Structure with the Nuclear Annn Both ends of the tunnel shall have a separation of 4 inches from the adjoining :
'(7) structure. A ' atertight rubber seal shall be used at each end of the tunnel. The inside clear span dimensions arc eleven feet by eight feet. The roof, walls and mat are constructed of reinforced concrete. The roof and walls are two feet thick and the mat is three feet thick. : Nem a ce v meww.ou w orssc rope 2.sa.s2
System 80+ Design ControlDocument 11.7.3 Elevations The component cooling water tunnel configuration is site specific. 11.7.4 Codes and Standards The codes and standards applicable to Seismic Category I buildings shall be met. 11.7.5 Loads in addition to the minimum design loads requirements of Section 5.1 of this appendix, the following specific additional load requirements shall be met. Should conflicting values occur between this Section and Section 5.1 of this appendix, the values specified in this Section apply. 11.7.5.1 Live Load (L) The tunnel floor shall be designed for a live load of 100 psf. The roof shall be designed for soil overburden pressure, AASHO H20-44 truck loading and construction equipment loading as applicable to the specific site. 11.7.5.2 Temperature Loads (T,) The normal concrete surface operating temperature and ambient ground temperature is site specific. 11.7.5.3 Seisnue Loads (E') The seismic accelerations shall be as specified in the Table 3.8A-11. The tunnel shall be seismically designed to sustain soil movement during earthquake ground motions. The structural integrity of the tunnel is evaluated by accounting for the two primary effects of earthquake motion, namely:
- 1. strains and associated stresses induced in the tunnel by the free-field vibration resulting from motions of the surrounding soil mass (seismic wave passage), and
- 2. seismically induced differential movements of the ends of the tunnel (i.e., the Nuclear Island and the CCW Heat Exchanger Structure).
Equivalent static analysis shall be performed considering the tunnel as a beam on an elastic foundation. Axial stress caused by seismic waves, soil friction, thermal expansion and differential movement shall be considered. Friction between the tunnel and the surrounding soil shall be considered using conservative estimates of the associated frictional forces. 11.7.5.4 Other Loads All abnormal loads (i.e., P., T , R., Y), Y, and Y,) are zero. O Approved Design Materia! Desists of SSC Page 18A-54
Sv~ tem 80+ Design ControlDocument . t [Q 11.7.6 . Loading Combinations and Acceptance Criteria 11.7.6.1 Concnte The tunnel shall be designed using Seismic Category I criteria. The requirements of Sections 5.2.2 and 8.0 of this appendix shall be met. 11.7.6.2 Stability r The requirements of Section 5.2.4 of this appendix shall be me;. , 11.7.7 Other Requirements The building is to be founded on competent structural backfill as defined in Section 10.1 of this appendix. The bearing pressure shall not exceed the value given in Table 2.0-1. P The Component Cooling Water Heat Exchanger Piping tunnel shall be sealed against the introduction of exterior water sources into the tunnel and shall be sealed at the interface with safety related structures to prevent flooding effects. The water seals are designed for the static pressure of water at the flood elevation. Water seals to preclude flooding of the Nuclear Annex caused by Service Water piping failure in the Component Cooling Water Heat Exchanger Structure and those at the interface to the Nuclear : Island Structure are designed to maintain integrity in the event of a Safe Shutdown Earthquake. In the event of seal failure, any credible leakage is limited to the capacity of the sump pumps in the safety O related structure.
.h 11.8 Buried Cable Tunnels, and Conduit Banks 11.8.1 Conduit Classification i
e Quality Class 1 ,
- l e Safety Class 3 j e Seismic Category I 11.8.2 Description l I Buried cable tunnels and conduit banks are reinforced concrete box type structures, generally rectangular i
in cross-section that house conduit for electrical distribution. 11.8.3 Codes and Standards l The codes and standards applicable to Seismic Category I buildings shall be met. 1 l 11.8.4 Loads i n in addii .. to the minimum design loads requirements of Section 5.1 of this appendix, the following l
-Ig* specific additional load requirements shall be met. . Should conflicting values occur between this Section and Section 5.1 of this appendix, the values specified in this section apply.
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System 80+ ,. Design ControlDocument 11.8.4.1 Dead Load (D) The weight of the contents of the cable tunnel and/or conduit bank. 11.8.4.2 Live Load (L) The structure shall be designed for soil overburden pressure, AASHO H20-44 tmck loading and construction equipment loading as applicable to the specific site. 11.8,4.3 Seismic Loads (E') The reinforced concrete buried cable tunnels and/or conduit banks shall be seismically designed to sustain soil movement during earthquake ground motions. The structural integrity of the cable tunnel and/or conduit bank is evaluated by accounting for the two primary effects of earthquake motion, namely;
- 1. strains and associated stresses irgluced in the tunnel by the free-field vibration resulting from motions of the surrounding soil mass (seismic wave passage), and
- 2. seismically induced differential movements of the ends of the tunnel (i.e., the Nuclear Island and the CCW licat Exchanger Stmeture).
Equivalent static analysis shall be performed considering the conduit tunnel as a beam on an elastic foundation. Axial stress caused by seismic waves, soil friction. thermal expansion and differential movement shall be considered. Friction between the tunnel and the surrounding soil shall be considered using conservative estimates of the associated frictional forces. 11.8.4.4 Other Loads All abnormal loads (i.e., P., T., R., Y ,3 Y,n and Y,) are zero. 11.8.5 Loading Combinations and Acceptance Criteria 11.8.5.1 Concrete The cable tunnels and conduit banks shall be designed using Seismic Category I criteria. 11.8.5.2 Stability The requirements of Section 5.2.4 of this appendix shall be met. 11.8.6 Other Requirements The cable tunnels and conduit banks are to be founded on competent structural backfill as defined in Section 10.1 of this appendix. The bearing pressure shall not exceed the value given in Table 2.0-1. O Anrowd Design Naserial. Design of SSC Page .T.8A 56
.- - . - . - _ . - . . - ~ - . -, .- . ... .- . .. -- .- -
Sv'twn 80+ oesten contrar Documart t
-' Cable tunnels and conduit banks shall be sealed against the introduction of exterior water sources into the l tunnel or bank and shall be sealed at the interface with safety related structures to prevent flooding . ;
- effects. The water seals are designed for the static pressure of water at the flood elevation. Water seals i at the interface with safety related structures are designed to maintain integrity in the event of a Safe ,
- Shutdown Earthquake. In the event of seal failure, any credible leakage is limited to the capacity of the sump pumps in the safety related structure, or the associated flooding effects are shown to be acceptable. ;
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- Aorm coneen menenw onein erssc reo, s.sa-sr L-
System 80+ Design ControlDocument Table 3.8A-1 Design Loads for Nuclear Island Category I Structures leadings Dead Live Rain & Fluid Temp.* F Load Lead Wind Soll Pressure Tornado Min / Max Structures Equipment /Imad (D) (L) (L & W) (II) (F) (W,) (T.) Remarks Nuclear Island Foundation Basemat
*
- N/A *
- N/A (TBD) Notes: 1, 2,13,14 Containment Internal Structure *
- N/A N/A
- N/A (TBD) Notes: 1, 2,12 Primary Shield Wall Operating Floor. El.146* +0* 50 psf 200 psf El. I 15' + 6- 50 psf 200 psf El 91 * + 9" 50 psf 200 psf IAwer Concrete Dish (TBD) See IRWST Crane Wall Note: 14 IRWST Water Level El. 82* + 6" Refueling Cavity Water 12 vel El.144' + 0* Note: 15 Reactor Vessel Compenents
- Note: 10 Equipment Reactor Vessel 2050K (Operating)
Steam Generators 2 @ 1770K Dry 2 @ 2600K Wet Pressurizer 445K Wet Reactor Coolant Pumps 4 @ 279K Dry 4 @ 285K Wet NSS Piping 280K Dry Reactor Drain Tank 33.9K Wet Safety injection Tank 261K Wet Polar Crane 630K (TBD) Note: 11 Approved Design Material- Design of SSC Page 3.8A-58 1 O O O
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Design ControlDocument
=N"?** #0 *- _ -- - - _ . - _ -
Table 3.8A-1 Design Loads for Nuclear Island Category I Structures (Cont'd.) Loadings Dead Live Rain & 1%id Temp.* F Load Load Wind Soil Pressure Tornado Min / Max Structures Equipment /imad (D) (L) (L & W) (H) (F) (W,) (T ) Remarks N/A N/A
- N/A Reactor Building Subsphere GBD) Notes: 1. 2 Con;ainment Support Pedestal See Lower Concrete Dish Walls 10 psf El. 115' + 6* 50 psf 200 psf El. 91'+9" 50 psf 200 psf El 70* + 0* 50 psf 200 psf I
Equipment Miniflow IIx Monorails 4 @ frBD) $ ton Cont. Spray IIx Monorail 2 @ (TBD) 5 ton Shutdown CS IIx Monorail 2 @ (TBD) 5 ton El. 50* +0* 50 psf 200 psf Equipment Containment Spray hx. 2 @ (TBD) Shutdown Cooling lix. 2 @ (TBD) Safety injection Pumps 4 @ (TBD) Cont. Spray Miniflow IIx 2 @ (TBD) Shutdown CS Miniflow IIx 2 @ (TBD)
*
- Notes: 1,2, Containment Shield Building
- N/A N/A *
(TBD) 12
* * * * *
- Notes: 1. 2, Nuclear Annex (TBD) 12 Nuclear Annex Interior Walls 10 psf Nuclear Annex Exterior Walls Notes: 3,4. ;
5, 6 Nuclear Annex Roof Stabs 50 psf 50 psf N/A Notes: 4, 7 Approved Design Material- Design of SSC Page 3.8A-59
System 80+ Design Contnd Document Table 3.8A-1 Design Loads for Nuclear Island Category I Structures (Cont'd.) Imadings Dead Lise Rain & Fluid Temp.* F Load Imad Wind Sou Pressure Tornado Min / Max Structures Equipment /Imad (D) (L) (L & W) (II) (O (W,) (T ) Remarks
* * * * *
- See Nuclear Control Complez (TBD)
Annex El. I15'+ 6* 55 psf 200 psf El. 91 * + 9- 55 psf 200 psf El. 70* + 0* 55 psf 200 psf El. 50* + 0* 55 psf 200 psf Main Steam Valve llouse 40/115 Note: 8 See Nuclear Annex El. 106' + 0* 50 psf 200 psf Emergency Feedwater Tank Areas 4 EFW Tanks II 20 @ El.100* +0* Note: 15, See Nuclear Annex See Nuclear Diesel Generator Areas (TBD) Annex Equipment Diesel Generators 2 @ 200K Bridge Crane 2 @ 20 tons El.50*+0* 75 psf 300 psf
* * * * *
- See Nuclear CVCS Area (TBD)
Annex Crane 225 ton Bridge Crane 270K (TBD) Notes: 9, II El.170' + 0* 75 psf 300 psf El.146' + 0* 75 psf 300 psf Note: 9 El.130* + 0* 75 psf 300 psf El. 115' + 0* 75 psf 300 psf Equipment Monorails for El. 91*+0* 4@2 tons El. 91 * +0* 75 psf 300 psf Rail Accessfiruck Bay Shipping Loads (TBD) Approved Design Materlo!- Design of SSC Page 3.8A-60 0 0 0
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,) N._.) N_Y System 80+ _
Design Control Document Table 3.8A-1 Design Loads for Nuclear Island Category I Structures (Cont'd.) Imadings Dead IJve Rain & Fluid Temp.'F Load Load Wind Soil Pressure Tornado Min / Max Structures Equipment / Load (D) (L) (L & W) (II) (D (W,) (T,) Remarks El 81 * + 0* (TBD) El. 70' + 0* 75 psf 300 psf Equipment Fuel Pool Purification X 4 @ (TBD) Deborat. Ion Exchanger i @ (TBD) Pre lloidup lon Exch. I @ (TBD) Boric Acid Cond. lon Exch I @ (TBD) Spare ton Exchanger lloidup Pumps 2 @ (TBD) El. 50' + 0" 75 psf 300 psf Equipment Spent Resin Slice Tanks 2 @ (TBD) Resin Sluice Pumps 2 @ (TBD) See Nuclear Fuelllandling Area (TBD) Annex Fuel Pool Crane 150 ton bridge crane 433K (TBD) Note: 11 El.170* + 0* 75 psf 300 psf El.146 +0* 75 psf 300 psf El.130' + 6" 75 psf 300 psf El. I 15' + 6" 75 psf 300 psf Fuel Pool Borated Water El. 104' + 0" bottom Note: 15 El. 144' + 0" max. Fuel Racks (TBD) Fuel Assemblics (TBD) El. 91 * + 9* 75 psf 300 psf Rail Access / Truck Bay Shipping Loads (TBD) El. 50* +0* 75 psf ' 300 psf Approved Design Material- Design of SSC Page 3.8A-61
System 80+ Design Control Document Table 3.8A-1 Design Loads for Nucicar Island Category I Structures (Cont'd.) Notes: [1] The mass of all structurca shall be included in all load combinations as dead loads. [2] All structures shall be designed for seismic loads. [3] See Section 5.1.1.3 of this appendix for design soil loads, including groundwater, Section 5.1.1.4 of this appendix. [4] See Section 5.1.1.6.1 of this appendix for thermal loads. [5] See Section 5.1.2.1 of this appendix for wind leads. [6] See Section 5.1.3.2 of this appendix for tornado loads. [7] See Section 5.1.1.2.1 of this %,pendix for added live load due to precipitation. [8] Abnormal loads due to Main Steam and Feedwater line breaks shall be considered. [9] Loads for one piece steam generator removal; i130K at El.146, Crane load would be 1920K max on one rail with 480K simultaneously on the other rail. [10] Refueling Cavity is used for temporary support of RV Components; Core Support Barrel, Upper Guide Structure, Reactor Vessel IIcad. [11] Design crane loads shall include rated lifting capacity with required safety margins, impact load factors, and applicable crane hold down requirements at crane rail. [12] Extreme external temperatures must be evaluated to determine temperatures to be combined with extreme internal temperatures. [13] Soil surcharge load on exterior walls due to construction loads. [14] Differential settlement shall be accounted for. See Section 10.1.2 of this appendix for basemat. See Section 10.4.2.2 of this appendix for Crane Wall. [15] Dynamic effects due to sloshing of water dall be included. Approved Design .'Asterial- Design of SSC Page 3.8A-62 O O O
Sy~ tem 80+ Desian controlDocument
)- Table 3.8A-2 Diesel Fuel Storage Structure, SSE Accelerations in Gs Elevation Long Direction Short Direction Vertical Roof 0.855 0.661 0.686 Basemat 0.648 0.542 0.680 a
Table 3.8A-3 Component Cooling Water Heat Exchanger Structure, SSE Accelerations in Gs Elevation Long Direction Short Direction Vertical Roof 0.727 1.213 0.690 First Floor 0.574 0.892 0.685 Basemat 0.513 0.819 0.676 l l Table 3.8A-4 Dead Weight Loads for Major Radwaste Facility Equipment ! Weight Quantity (Kips) l Location Item l Basemat North Demineralizer 15 15 Basemat North Chemical Waste Tanks 2 90 Basemat North Chemical Sample Tanks 2 90 Basemat North Detergent Sample Tanks 2 M Basemat North Laundry & Ilot Shower Tanks 2 64 Basemat North Waste Monitor Tanks 4 305 Basemat Nonh Floor Drain Tanks 2 305 Basemat North Equipment Waste Tanks 2 306 Basemat South Low Activity Spent Resin Tank 2 35 Basemat South High Activity Spent Resin Tank 1 35 Second 1 oor Low Activity Spent Resin Surge Tank 1 18 Second Floor High Activity Spent Resin Surge Tank 1 18 (O, The above weights are the operating weights of each item and includes the weight of contained fluids.
- Q/ '
I
. AppwmiDeelour Motoriel.Desiger of SSC Page 3.8A-63
1 System 80+ Design ControlDocument l Table 3.8A-5 Live Loads for Radwaste Facility h Live Load Location Area psf , Basemat All areas 250 First Floor Corridors 200 Platform Grating 150 All other areas 250 Second Floor Corridor 200 Platform Grating 150 Full Cask Storage Area 500 HIC Storage Area Notell3 Truck Bay Noter2j All other areas 250 Third Floor Corridors 150 All other areas 200 Roof Roof 100 Notes: I'3 1000 psf or HIC impet ioad I23 The truck loading shall be IIS20-44 higim7 foading. Anaroved Design Materiet - Design of SSC Page 3.BA G1
System 80+ Design crntrolDocument ( Table 3.8A-6 Radwaste Facility, SSE Accelerations in Gs Elevation NS EW Vertical Roof (by COL) (by COL) (by COL) Second Floor (by COL) (by COL) (by COL) First Floor - - (by COL) (by COL) (by COL) BS : mat (by COL) (by COL) (by COL) Table 3.8A-7 Dead Weight Loads for Major Turbine Building Equipment Weight , Location Item Quantity (Kips) Oround Floor Condenserstil 3 2750 Operating Floor Reheater 2 450 Reheater Drain Tank 2 27 1st Staged 1 b y Reheater Drain Tank 2 27 2nd Stagetz i Operating Floor / Turbine Low Pressure Turbine 3 8500 ; Pedestal Operating Floor Deacrator FW Storage Tank 1 1700 MS Drain Tankt21 2 31 Mezzanine Floor FT Heater 4 290 ) The above weights are the operating weights of each item and includes the weight of cr n ained fluids. ((The site specific SAR shall verify the above information based upon the turbine purchased.))tsi IU includes LP Feedwater Heaters 121 Supports benow the operating floor Ill COL information item: see DCD Introduction Section 3.2. Anwed Onisrr A0neria! Dugn of SSC Page 3.8A 65 I
System 80+ Design Control Document f Table 3.8A-8 Live Loads for Turbine Building h location Beams and Stabs GirderslU, Columns and I (ps0 Footings (ps0 Basemat floor, concrete t 250z) goot2) l 100)1 100 Basement floor, grating Mezzanine floor, concrete t 250z) 100t2) Mezzanine floor, grating 125 75 l 175 Operating floor, concrete 250'1 Operating floor, grating 125 100 All other grating and checkered plate 100W 75DI floors O til Carrying over 400 sq ft of floor area. (2) Check for equipment laydown and stator erection loadings. 131 Grating, checkered plate and framing in trucking aisle is designed for a 5 ton capacity fork lift truck. 141 The slabs are designed for 750 psf in the laydown area as shown on the mechanical arrangement drawings. 151 Accumulation oflive load from all grating floors, walkways and miscellaneous platforms does not exceed 25 kips to any column or footing. Approwmf Deslen heatorial * $esiste of SSC page 3.gA.66
System 80+ Design controlDocument i i (h t . v) Table 3.8A-9 Turbine Building, SSE Accelerations in Gs j Elevation NS EW Ve:11 cal , Ground Floor (by COL) (by COL) (by COL) Mezzanine Floor (by COL) (by COL) (by COL) -i Operating Floor (by COL) (by COL) (by COL) . I Table 3.8A-10 Dikes for Outdoor Tanks, SSE Accelerations in Gs ; Elevation NS EW Vertical i Foundation (by COL) (by COL) (by COL) i i Table 3.8A-11 Component Cooling Water Tunnel, SSE Accelerations in Gs Elevation NS EW Vertical O-j Roof 0.54 0.56 0.62 I Mat 0.54 0.56 0.62 4 O , + L ..:n Design hiesenle!- Design of SSC Pope 3.8A-67 i i t
System 80+ Design controlDocument O' This Figure Intentionally Blank, see Figure 3.8-5, Sheet 4 O1! I 1 l l I i l l Nuclear Island Structures - Plan at Level 2 Hgure 3.8A 1 Sheet 1 of 2 i 1 l l Amund Des &n merew on&n et ssc Page 2.s se >
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l Effective Page Listing L Appendix 3.8B l l 1-Pages Date i, ii 1/97
. iii 11/%
i iv - vi Original l 3.881 through 3.8B-61 Original l Lo ) l t-i O r ..::o.oonaneurw w orssc tusn raue 1. n
" System 80+ Design ControlDocument . i (] Appendix 3.8B
.v t
Structural Design Details l i a t Conists Page i
.1.0 Objective and Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
3.8B-1 2.0 Description of the Nuclear Island Critical Areas . . . . . . . . . . . . . . . . . . . . . . . 3.8B-1 ; 3.0 Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-1 [ 4.0 Loads and Load Combinations . . . . . . ........... . . . . . . . . . . . . . . . 3. 8B- 1 4.1 Loading Combinations for Seismic Category I Concrete Structures . . . . . . . . . . . 3.8B-2 f 4.1.1 Service Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 B-2 4.1.2 Factored Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8B-2 4.2 Loading Combinations for Seismic Category I Steel Structures . . . . . . . . . . . . . . 3.8B-2 3 4.2.1 Service Load Conditions .....................................3.8B-2 d 4.2.2 Factored Load Conditions ........................ ........... 3.8B-2 . I 5.0 Analyses and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-3 5.1 Areas I A,1B, and 1C - East Face of Shield Building Wall . . . . . . . . . . . . . . . . 3.8B-3 5.1.1 Description of Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-3 5.1.2 General Loads . . . . . . . . . . . . . . . . . . .... .. . . . . . . . . . . . . . . . . 3. 8B-3 5.1.3 Governing Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-3 5.1.4 Analysis Methods and Results . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 B-4 5.1.5 Typical Reinforcing Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-5 5.1.6 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... ....... 3.8B-5 5.2 Area 2 - East End Wall Adjacent to Turbine Building . .... . . . . . . . . . . . . . 3.8B-6 l 5.2.1 Description of Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-6 ; 5.2.2 General Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8B-6
'5.2.3 Governing Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-6 l 5.2.4 Analysis Methods and Results . . . . . . . . . . . . . . . ........ .......... 3586 l 5.2.5 Typical Reinforcing Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-7. !
5.2.6 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8B-7 5.3 Areas 3A and 3B - Emergency Diesel Room Interior and Exterior Walls . . . . . . . . 3.8B-7 5.3.1 Description of Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-7 5.3.2 General Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-7 5.3.3 Governing Load Combinations ...............................3.8B-7
'5.3.4 Analysis Methods and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 B-8 s 5.3.5 Typical Reinforcing Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . 8 B-9 5.3.6 Conclus ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3. 8B-9 -(. )'
5.4. Areas 4, 6 and 7 - Containment Pedestal, Dish and Support . . . . . . . . . . . . . . . . . 3.8B-9 Amwomt onw nonaww- onw or ssc nrasi rose a ; i
Srtem ?O + Design ControlDocument Contents (Cont'd.) Page 5.4.1 Description of Area . . . . . . . . . . . . . . ......... ........... ..... 3.8B-9 5.4.2 General Loads . . . . . . . . . .. . .. .................. ... 3.8B-10 5.4.3 Governing Load Combinations and Results . . . ........... ..... .... 3.8B-10 5.4.4 Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............ 3.8B-16 5.4.5 Typical Reinforcing Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 3.8B-16 5.4.6 Conclusions . . . . . . . . . . . . . . . . .... ........................ 3.8B-19 5.5 Areas 5A, 5B and SC - South Section of Shield Buil . . . . . . . . . . . . . . . . . . . 3.8B-20 5.5.1 Description of Areas . . . . . . . . . . . . . . . . . . ........... ......... 3.8B-20 5.5.2 General loads . . . . . . . . . . . . . . ............... ............. 3.8B-20 5.5.3 Governing Load Combinations ...... .... .... ... ........... 3.8B-20 5.5.4 Analysis Methods and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-21 5.5.5 Typical Reinforcing Details . . . . . . . . ........ .. ........ .. ... 3.8B.22 5.5.6 Conclusion . . . . . . . . . . . .................................. 3.8B-23 5.6 Area 8 - Steam Generator Enclosure Radial Wall . .................... 3.8B-23 5.6.1 Description of Area . .. . ................................ 3.8B-23 5.6.2 General Loads . . . . ............. ........................ 3.8B-23 5.6.3 Governing Load Combinations . . . . . . . . . . . . . ... ...... ......... 3.8B-23 5.6.4 Analysis Methods and Results . . . . ............. ...... ........ 3.8B-23 5.6.5 Typical Reinforcing Details . .... .......... .......... .. .... 3.8B-24 5.6.6 Conclusion . . . . . . . . ........... .... . . .. .. ....... 3.8B-24 5.7 Arca 9 - Spent Fuel Pool Wall . . . . . . . . . . . . ... . ... ...... 3.8B-24 5.7.1 Description of Area . . . ...... ... . .. .. ... .. ... 3.8B-24 5.7.2 General Loads . . . . . . ... . ..... .......... .... ..... .. 3.8B-24 5.7.3 Governing Load Combbations . . . .. ... . .. ......... .... 3.8B-24 5.7.4 Analysis Methods and Results .. .. ..... ..... ... . . ... 3.8B-25 5.7.5 Typical Reinforcing Details . . . .... .......... .... .. ....... 3.8B-25 5.7.6 Conclusion . . . ...... . .. .. . . ........ .... . .. ... 3.8B-25 5.8 Area 10 . . ......... .. ..... . . . . ...... ... .. 3.8B-26 5.8.1 Description of Area . . . ..... .. . ... . . . .... .... 3.8B-26 5.8.2 General Loads . . . . . .... . ..... .. . .. . ....... ... 3.8B-26 5.8.3 Governing Load Combinations . . . .......... . .. ... . .. .... 3.8B-26 5.8.4 Analysis Methocs and Results . ....... . . . ... .... ....... 3.8B-26 5.8.5 Typical Reinforcing Details . ... .. ... . ... ........... . 3.8B-27 5.8.6 Conclusion . . . . . . . . . . . . . . . .. . .. . ... .. . ....... 3.8B-27 5.9 Area 11 - North-West End Wall . ... .. . .... . ........ 3.8B-27 5.9.1 Description of Area . ...................................... 3.8B-27 5.9.2 General Loads . . . . . .. ............. .. .. ..... . ... 3.8B-27 5.9.3 Governing trud Combinations . . . . . . . . ........ .......... ..... 3.8B-28 5.9.4 Analysis Methods and Results .... .. ... . ...... .... .... 3.8B.28 5.9.5 Typical Reinforcing *.)etails . ... . ......... ................. 3.8B-28 i 5.9.6 Conclusion . . . . ...... ..... ....... .. .............. .. 3.8B-29 1 5.10 Area 12 - Interior Structure Steel Column . . . . . . . . . . .............. 3.8B-29 l 5.10.1 Description of Area . . . . . . . ....... .................. . .. . 3.8B-29 5.10.2 General Loads . . . . . . . . . . . . . . . . . . . ............ ... ..... 3.8B-29 5.10.3 Governing Load Combinations . . . . ............. .......... 3.8B-29 5.10.4 Analysis Methods and Resuks ...... .. .. .. ...... ...... 3.8B-29 5.10.5 Design Details . . . . . . . ..... . ...... ......... .. ... 3.8B-30 5.10.6 Conclusion . . ........ .. . ............ .... .. . .. . 3.8B-30 Alvvoved Design Material Design of SSC Page iv
Sv: tem 80+ Design ConVolDocument
-i
'W i t Contem's (Cont'd.) Page 5.11- Area 13 - Nuclear Island Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-30 3 5.11.1 Description of Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-30 5.11.2 General Loads . . . . . . . . . . . . . . . . . . . . . . . .................... 3 8B-30 - i 5.11.3 Governing I. cad Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.83-30 5.11.4 Analysis Methods and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-30 , 5.11.5 Typical Reinforcing Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-31 5.1 1. 6 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-31 l l 6.0 Typical ACI-318 Chapter 21 Ductility Connection Details . . . . . . . . . . . . . . . . 3.8B-31 7.0 Non-Nuclear Island Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.BB-32 7.1 Diesel Fuel Storage Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . 3.8B-32 7.1.1 Description of Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-32
'7.1.2 Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-32 7.1.3 Loads and Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . 3.8B .22 7.1.4 Analyses and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B 32 7.1.5 Concl usion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-33 7.2 Component Cooling Water Heat Exchanger Structure . . . . . . . . . . . . . . . . . . . . 3.8B-34 ,
7.2.1 Description of Structure . . . . . . . . . . . . . . . . . . . . . . . . . ............ 3.8B-34 7.2.2 Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-34 7.2.3 Loads and Load Combinations . . . . . . . . ......................... 3.8B-34 , 7.2.4 Analyses and Results ............................... ....... 3.8B-34 7.2.5 Conclusion . . . . . . . . . . . . ..... ......................... . 3.8B-35 ( 7.3 Component Cooling Water Tunnel ..................... ......... 3.8B-35 7.3.1 Description of Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3.8B-35 7.3.2 Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-36 ' 7.3.3 Loads and Load Combinations . . . . . . . . . . . . . . . . ... ............. 3.8B-36 <
- 7.3.4 Analyses and Results . . . . . . . . . . . . ... ... .................. 3.8B-36 7.3.5 Concl u:; ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-37 :
Tables Page
]
1 3.8B.1 Areas identified for Detailed Design ... .. . .. .. .. ....... 3.81i-38 Figuns Page 3.83-1 Structural Critical Areas Section B-B . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-39 l 3.63-2 Typical Ductility Reinforcing Details Details & Notes .............. 3.8B-49 3.8B-3 Nuclear Island Basemat to Exterior Wall Connection with Four Foot Basemat Extens ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8B-52 i 3.8B-4 Nuclear Island Basemat to Exterior Wall Connection Details . . . . . . . . 3.8B-53 3.8B-5 Diesel Fuel Storage Structure . . . . . . . .. .................. 3.8B 3.8B-6 Diesel Fuel Storage Structure keinforcing Details . . . . . . . . . . . . . . . . 3.8B-56
- g 3.8B 7 General Arrangement CCW Heat Exchanger Structure . . . . . . . ...... 3.8B-57 3.8B-8 CCW Heat Exchanger Structure Reinforcing Details ............... 3.8B-58 3.8B-9 . CCW Heat Exchanger Structure Reinforcing Details ............... 3.8B-59 I m uewdomeoneaww ou w orssc noe v
Syatem 80+ Design ControlDocument Figures (Cont'd.) Page 3.8B-10 Component Cooling Water Tunnel . . ........................ 3.8B-60 3.8B-11 Component Cooling Water Tunnel Reiaforcing Details . . . . . . . ... . 3.8B-61 O O ApprendDesign Materie! Design of SSC page vi
system 80+ .___ oesian coneer occamaer s 1.0 Objective and Scope This Appendix presents analysis results and typical main reinforcing design for thirteen selected areas of
- the System 80+ Seismic Category I Nuclear Island structure, the Diesel Fuel Storage Structure, Component Cooling Water Heat Exchanger Structure, and Component Cooling Water Tunnel using the criteria in Appendix 3.8A. Based on the general arrangement of major structural elements and !
components, the thirteen Nuclear Island areas are selected to provide representative design details for ; structural elements having both typical and unique design requirements. Design details for the steel 2 containment are included in Section 3.8.2. ; i In addition to the evaluation of the thirteen a*eas, shear requirements have been calculated and capacities demonstrated for all major shear walls of the Nuclear Island. ! The resulting design forces and moments presented in this Appendix are from use of a conservative j - envelope of design loads. Reinforcing details presented are typical details to develop the capacity , i required to envelope these forces and moments. Ductility reinforcing requirements for concrete sections' ! i. are provided in Section 6.0 of this appendix. The design review demonstrates that it is feasible to design , and construct the structures as configured in the general arrangements presented in Chapter 1. The . 8
- structural analysis report, Section 3.8.4.5.4, will document that the final design details for the Nuclear i Island structure meet the analysis and design criteria of Section 3.8. ;
Design and analysis details of the Diesel Fuel Storage Structure, Component Cooling Water Heat 4 Exchanger Structure and Component Cooling Water Tunnel are provided in Section 7.0 of this appendix. J 2.0 Description of the Nuclear Island Critical Areas ~ The location and description of the thirteen areas are identified in Table 3.8B-1. The areas are shown in Figure 3.8B-1, Sheets 1-5. 3.0 Analysis Methods s
- The Nuclear Island is analyzed to account for both global and local effects of design basis loads described in Appendix 3.8A.
4 The complete Nuclear Island is founded on a common basemat and is analyzed as a monolithic structure.
' A three dimensional finite element model of the Nuclear Island is developed and equivalent static global loading conditions are applied to the scructure. These results are combined using the loading combinations identified in Section 5.2 of Appendix 3.8A. The global results from the three dimensional finite element model are combined with local analysis results to determine forces and moments for the design of the walls, columns and slabs. ;
i The analysis methods are described in further detail in Appendix 3.8A, Section 6.1. j 4.0 Loads and Load Combinations . , The loads evaluated fr de Nuclear Island are addressed in Appendix 3.8A, Section 5.1. i
- --._' W asE W *o00lpFofSSC Page 3.061
.m.m- -r,. -:,.- - , ,;, - . , , .,,-,- , , ,,,, -,,- w .w , , x, y. ,, . . ., - - -c ,
.-,v p
- System 80+ Design Control Document The following loading ccmbinations from Table 3.8-5 and Appendix 3.8A are used for analysis and design of Category I structures and their components. See Appendix 3.8A, Section 5.0 for definitions. 4.1 Loading Combinations for Seismic Category I Concrete Structures 4.1.1 Service Load Combinations (1) U = 1.4D + 1.7L (2) U = 1.2D + 1.7W (3) U = 1.4D + 1.7F + 1.7L + 1.7H + 1.7W (4) U = (.75)(1.4D + 1.7F + 1.7L + 1.7H + 1.7To + 1.7Ro) (5) U = (.75)(1.4D + 1.7F + 1.7L + 1.7H + 1.7W + 1.7To + 1.7Ro ) 4.1.2 Factored Load Combinations (1) U = D + F + L + H + To + Ro + E' (2) U = D + F + L + H + To + R o+ W, (3) U = D + F + L + H + T, + R, + 1.5 P, (4) U = D + F + L + H + T, + R, + 1.0P, + 1.0(Y,+Yj +Y,) + E' 4.2 Loading Combinations for Seismic Category I Steel Structures 4.2.1 Service Load Conditions 4.2.1.1 Elastic Allowable Strength Design (1) S=D+F+L+H (2) S=D+F+L+H+W (3) 1.3 S = D + F + L + H + Ro + To (4) 1.3 S = D + F + L + H + W + Ro + To 4.2.1.2 Plastic Design (1) Y = 1.7 (D + F + L + H) (2) Y = 1.7 (D + F + L + H + W) (3) Y = 1.3 (D + F + L + H + To + Ro) (4) Y = 1.3 (D + F + L + H + W + To + Ro) 4.2.2 Factored Load Conditions 4.2.2.1 Elastic Allowable Strength Design (1) 1.4 S = D + F + L + H + R o + To + E' (2) 1.4 S = D + F + L + H + R o + To + W, (3) 1.4 S = D + F + L + H + R, + T, + P. (4) 1.6 S = D + F + L + H + R, + T, + (Y,+ Y;+ Y,) + E' + P. O Agnprowd Des &n Meterial . Design of SSC Pope .1.88-2
Sy-tem 80+ Design ControlDocument i 4.2.2.2 Plastic Design Methods (1) Y = 1.0 (D + F + L + H + R, + To + E') (2) Y = 1.0 (D + F + L + H + R o + T o + W) (3) Y = 1.0 (D + F + L + H t R, + T + 1.5 P.) (4) Y = 1.0 (D + F + L + H + R, + T + Y, + Y) + Y, + E' + P ) 5.0 Analyses and Results 5.1 Areas IA, IB, and IC - East Face of Shield Building Wall 5.1.1 Description of Area Area 1 defines the sections of walls located on the east face of the Shield Building. The wall sections include areas I A, IB, and IC. Area l A is the east-west wall section located Ong column line 17 and is immediately east of the Shield Building wall. This four feet thick wall is located between the top of the basemat at elevation 50'+0" and the floor slab at elevation 115'+6". This area of wall is part of the shear wall that resists the east-west seismic forces in the Nuclear Annex. This shear wall also functions as a divisional wall separating mechanical and electrical safety-related equipment. Area IB is the north-south wall section adjacent to the east side of the Shield Building. It is located along column line E at its intersection with column line 17. This four feet thick wall extends from the top of the basemat at elevation 50'+0" to the floor s'ab at elevation 91'+9". This wall is a shear wall that resists north-south seismic forces in the Nuclea Annex. Area IC is the east side of the Shield Building wall. The Shield Building wall is a four feet thick right cylinder that extends from the basemat to the top of the Nuclear Annex roof where the thickness changes to three feet. The Shield Building has an inside radius of 105 feet. The Shield Building serves as a missile and wind barrier to protect the steel containment vessel. The Shield Building also provides an annulus region where containment leakage is contained and exhausted through the Annulus Ventilation system. 5.1.2 General Loads . The loads applicable to areas IA, IB, and 1C are summarized in Appendix 3.8A. Table 3.8A-1. 5.1.3 Governing Load CombinaHens Area IA Shear (in-plane) 4.1.2(4) (out-of-plane) 4.1.2(4) Bending 4.1.2(4)
/'~N V Axial '(tension) 4.1.2(4)
(compression) 4.1.2(1)
? 22:W neonerial Desips of SSC Page 3.88-3
SyQtem 80+ Design ControlDocument Area IB Shear (in-plane) 4.1.2(1) (out-of-plane) 4.1.2(1) Bending 4.1.2(1) Axial (tension) 4.1.2(4) (compression) 4.1.2(1) Area IC Shear (in-plane) 4.1.2(4) (out-of-planc) 4.1.2(1) Bending 4.1.2(4) Axial (tension) 4.1.2(4) (compression) 4.1.2(1) 5.1.4 Analysis Methods and Results The in-plane loads on areas 1 A and IB are predominantly shear loads from the SSE. The in-plane forces l are obtained from the application of these loads to the static three dimensional finite element model The out-of-plane loads on the wall are predominantly SSE seismic loads. The out-of plane resultant forces and moments are determined by hand calculation. The design forces and moments for Area 1A and IB are: Area 1A Shear (in-plane) 280 kips /ft (out-of-plane) 12 kips /ft Moment 190 ft-kips /ft Axial (tension) 40 kips /ft (compression) 240 kips /ft Area IB Shear (in-plane) 235 kips /ft (out-of-plane) 24 kips /ft Moment 130 ft-kips /ft Axial (tension) 50 kips /ft l (compression) 205 kips /ft kyroved Design Materia! Design of SSC Page 3.884
Sy: tem 80+ Design ControlDocummt The in-plane loads on area IC are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three dimensional finite element model. The out-of-plane loads on the wall are predommantly from the accident temperature differential from a postulated Annulus Ventilation System failure. The out-of plane resultant forces and moments are determined by hand calculation. The design forces and moments for Area IC are: Shear (inglane) 200 kips /ft (out-cf-plane) 86 kips /ft Moment (2 way bending) 402 ft-kips /ft 118 ft-kips /ft Axial (tension) 140 kips /ft (compression) 250 kips /ft 5.1.5 Typical Reinforcing Details Area IA Wall Thickness 4 feet
#18 at 12" vertical steel each face
#18 at 12" horizontal steel each face Shear ties not required Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
Area 1B Wall Thickness 4 feet
#14 at 12" vertical steel each face
#14 at 12" horizontal steel each face Shear ties not required Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
Area IC Wall Thickness 4 feet
#18 at 12" vertical steel each face f #14 at 12" horizontal steel each face Shear ties - #5 horizontal ties at 12" x 12"
. Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
5.1.6 Conclusion j The Area 1 concrete section strengths determined from the criteria in Appendix 3.8A -Je sufficient to resist the design basis loads.
)
N/ - Appre$1d Desirs Afsterie!* Destpr of SSC Page ll 88-5 i
Sy^ tem 80+ Design ControlDocument 5.2 Area 2 - East End Wall Adjacent to Turbine Building h 5.2.1 Description of Area Area 2 is a segment of the exterior wall at the East end of the Nuclear Island adjacent to the Turbine Building. The wall extends from the top of the basemat at elevation 50'+0" to the top of the roof at elevation 146'+0". The walls in this area are four feet thick. Out-of-plane lateral support is provided : to the walls by the floor slabs on the interior of the structure. The wall is arranged and designed to , I function as a major structural shear wall in addition to providing protection for the safety related equipment. 5.2.2 General Loads The loads applicable to Area 2 are summarized in Appendix 3.8A, Table 3.8A-1. The out-of-plane passive soit pressure loads are the predominant loads. The Nuclear Island evaluation credits the passive soil pressure loads to resist sliding. 5.2.3 Governing Load Combinations Area 2 Shear (in-plane) 4.1.2(1) (outef-plane) 4.1.2(1) Bending 4.1.2(1) Axial (tension) 4.1.2(1) (compression) 4.1.2(1) 5.2.4 Analysis Methods and Results The Area 2 wall is analyzed as a structural shear wall. The in-plane forces are obtained from output l computed by the application of these loads to the static three-dimensional finite element model. The out- ! of-plane loads on the wall are predominantly soil pressure loads with the effect of the SSE. The out-of plane resultant forces and moments are determined by local two dimensional frame models. l The design forces and moments for Area 2 are: ; i Shear (in-plane) 219 kips /ft l (out-of-plane) 273 kips /ft Moment 910 ft-kips /ft l Axial (tension) 50 kips /ft 1 (compression) 277 kips /ft O
. Approwd Design Material . Design of SSC Pope 3.88-6 l
l I System 80+ Desian controlDocument l
-i fs 'y 5.2.5 Typical Reinforcing Details
( Area 2 Wall Thic! mess 4 feet
#18 at 12" vertical steel,2 layers each face (below elevation 90'+3") l
#14 at 12" vertical steel,2 layers each face (above elevation 90'+3") ;
#11 at 12" horizontal steel,2 layers each face ;
Shear ties - #6 horizontal ties at 12" x 12" , Additional ductility reinforcing shall be provided as described in Section 6.0 of this apper. dix. 5.2.6 Conclusion The Area 2 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to
- resist the design basis loads.
5.3 Areas 3A and 3B - Emergency Diesel Room Interior and Exterior Walls 5.3.1' Description of Ares > Diesel generator areas exist on the north and south side of the Nuclear Annex. The interior wall is Area 3A and exterior wall is Area 3B. D Area 3A, the interior wall, extends from the top of the basemat at elevation 50'+0" to the top of the roof b slab at elevation 91'+9". This four feet wall continues upward ending at the top of the roof slab at elevation 191'+0". The wall at Area 3A functions as an east-west structural shear wall. Area 3B, the exterior wall, extends from the top to the basemat at elevation 50'+0" to the top of the roof slab at elevation 91'+9". This five feet exterior wall spans between the basemat and the roof slab. This wall also functions as an East-West shear wall. 5.3.2 General leads The loads applicable to Area 3 are summarized in Appendix 3.8A, Table 3.8A-1. The predominant loads i i on the exterior wall are from the out-of-plane soil pressure loads. Passive soil pressure was considered in the design of the er.teric r walls. The Nuclear Island evaluation credits the passive soil pressure loads to resist sliding. q Construction crane icd mis also considered on the exterior wall. Lateral bracing of the exterior wall is considered during construction due to the vertical span of the wall. 5.3.3 Goveming Load Combinations y Area 3A Shear (in-plane) 4.1.2(1) (out-of-plane) 4.1.2(4). L.)
= ..= on y neesaw.oesy orssc roue 3.asa ;
System 80+ Design ControlDocument Bending 4.1.2(1) Axial (tension) 4.1.2(4) (compression) 4.1.2(1) Area 3B Shear (in-plane) 4.1.2(4) (out-of-plane) 4.1.2(4) Bending 4.1.2(4) Axial (tension) 4.1.2(4) (compression) 4.1.2(1) 5.3.4 Analysis Methods and Results The in-plane loads on Area 3A are predominantly shear loads from the SSE. The in-plane forces are obtained from output computec by the application of these loads to the static three-dimensional finite element model. The out-of-plar.e loads on the wall are predominantly SSE seismic loads. The out-of-planc resultants and moments are determined by hand calculations. The design forces and moments for Area 3A are: Shear (in-plane) 208 kips /ft (out-of-plane) 14 kips /ft Moment 70 ft-kips /ft Axial (tension) 40 kips /ft (compression) 550 kips /ft The in-plane loads on Area 3B are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three-dimensional finite element model. The out-of-plane response of the wall is predominantly bending from soil pressure loads. The out-of-plane resultants and moments are determined by hand calculations and by local two-dimensional frame analysis models. The design forces and moments for Area 3B are: Shear (in-plane) 253 kips /ft (out-of-plane) 218 kips /ft Moment 1756 ft-kips /ft Axial (tension) 40 kips /ft (compression) 230 kips /ft O Approwd Design Motorial Design of SSC Page 3.88-8
. . - . - -.- - _ - - . _ - - . ... - . - .- . ~ . .- . ..
i Svstem' 80 + Design ControlDocument > l h, ' j 5.3.5 Typical Reinforcing Details
- Area 3A . Wall Thickness 4 feet ,
< #14 at'12" vertical steel each face
#14 at 12". horizontal steel each face .
Shear ties not required ; Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix. 5 Ama38' . Wall 1Mckne== 5 feet . a
- #18 at 12" venical steel,3 layers each face
. #18 at 12" horizontal steel each face
- Shear ties - #5 horizontal ties at 4" x 12*
Additional ductility reinforcing 'shall be provided as described in Section' 6.0 of this appendix. 5.3.6 Conclusion , The Area 3 concrete section strengths determined from the criteria in Appendix 3.8A are sufficiert to resist the design basis loads. 4- ' The' exterior wall requires lateral shoring during construction to withstand the potential overburden pressure loads from construction cranes. 5.4 Areas 4, 6 and 7 - Containment Pedestal, Dish and Support > 5.4.1 Description of Area
- j. This area comprises the primary structural components supponing the Steel Containment Vessel (SCV) and its internal structures. The SCV is supported by the pedestal and outer dish. The outer dish is supported by the lower crane wall, the pedestal, the floor slab at elevation 91'+9", and the radial walls. j i
This section addresses the design of these structural components, specifically described as follows: ; i
-. Area 7 Pedestal - Solid mass of concrete below the SCV, above the basemat, centered under the SCV, l nominally 66 feet in diameter.
I Area 6 and 7. Outer Dish - Concrete shell (3 feet thick) directly outside the SCV, below the floor slab at elevation 91'+9", above the pedestal.
- Area 6 Lower Crane Wall - Circular wall (4 feet thick) at radius 67 feet, below outer dish, above the a
basemat. LArea 4 Radial Walls - Walls radiating from the pedestal at 45 degree intervals, below the outer dish,
;g ' above the basemat.
l -
- Area 6 91'+9" Slab - Floor slab at elevation 91'+9", specifically that section between the top of the i outer dish and the first supporting wall at radius 84 feet. I 4preemt cos(ps afesenef - Desen et SSC Pope 3.83-9
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System 80+ Design ControlDocument 5.4.2 General Loads Loads to be resisted by the outer dish and its supporting structure include local loads and global loads. Global loads are obtained from the static three-dimensional finite element model results and are considered, as appropriate, in conjunction with localloads such as self weight and inertial forces. Loads from inside the SCV are distributed to the outer structure through the inner dish and the internal structure. In detennining the loads to be resisted by the outer dish, it is assumed that there is no transfer of longitudinal shear between the inner and outer dishes, thus there is no composite action be. tween the two 3 feet thick members. The inner dish supports loads such as self weight, water in the IRWST, and inenial forces associated with each. The outer dish supports loads such as its self weight and inertial forces. Included in the loads for the inner and outer dishes is a reaction force between them. Determining the magnitude of this reaction requires a consideration of relative displacements of the inner and outer dishes for each applicable load case. This is accomplished by modeling a typical section of the outer dish and the corresponding section of the inner dish, applying the appropriate loads to each, and determining deflections. 'Ihe reaction (equal and opposite for the two models for a given load case) is determined as that force necessary to cause equal midspan deflections of the two models, when combined with other applicable loads. There is no seal between the SCV and the inner dish. Therefore, it is assumed that expansion of the SCV due to internal pressure is resisted directly by the outer dish. As pressure is applied, membrane stresses develop in the SCV, along with an external force from the outer dish. The magnitude of this external force is calculated as the force required to restrain the SCV in the shape of the deflected concrete under all other loads, for each applicable load case. This force is inversely proportional to the deflection at a given point on the dish. Since the top of the SCV is free to expand vertically, the outer dish will only resist the horizontal component of the pressure. The tangential component of this horizontal force is relieved by movement of the SCV along the dish due to meridional strain, thus, only the normal component of the horizontal force contributes to the load on the outer dish. 5.4.3 Governing Load Combinations and Results Outer Dish: Circumferential Section Tension / flexure: No load combinations result in tensile axial loads. The critical load combination for maximum moment with zero axial load is: Load Combination: 4.1.2(4) M = 351 ft-kips /ft O 4swend Deshyn Ataterial Design of SSC Pane 3.88-10
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System 80+ Design ControlDocument
'(av) Compression / flexure:
The critical load combinations for maximum moment with compressive axial forces are: Load Combination: 4.1.l(4) P = 122 kips /ft M = 144 ft-kips /ft Load Combination: 4.:l.2(1) P = % kips /ft M = 316 ft-kips /ft
- Load Combination: 4.1.2(3)
P = 527 kips /ft M = 117 ft-kips /ft Out-of-Plane Shear: The critical load combination for out-of-plane shear is: Load Combination: 4.1.2(4) V = 51 kips /ft
.p L, Outer Dish: Radial Section (Inside the Crane Wall) i Tension / flexure:
The critical load combinations'for maximum moment with tensile axial forces are: Load Combination: 4.1.2(3) P = 5 kips /ft M = 325 ft-kips /ft Load Combination: 4.1.2(4) P = 74 kips /ft M = 206 ft-kips /ft Load Combination: 4.1.2(4) P = 67 kips /ft M = 219 ft-kips /ft Competssion/ flexure: To determine the critical combinations of compressive axial forces and associated moments, the variance , in spacing as well as the change in moment along the member must be taken into account. Three sections along the member are considered; the first 4 feet outside the pedestal, the last 4 feet inside the crane wall,
. ['t
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'and the remaining midsection. The smallest axial compression for a given load case is combined with the three moments; M(first), M(mid), and M(last). The e-itical combinations are:
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l System 80+ Design ControlDocument Load Combination: 4.1.2(3) P = 133 kips /ft M(first) = 552 ft-kips /ft M(mid) = 377 ft-kips /ft M(last) = 461 ft-kips /ft Load Combination: 4.1.l(1) P = 14 kips /ft M(first) = 143 fr-kips /ft M(enid) = 69 ft-kips /ft M0ast) = 119 ft-kips /ft Load Combination: 4.1.2(1) P = 111 kips /ft M(first) = 158 ft-kips /ft M(mid) = 92 ft-kips /ft M(last) = 121 ft-kips /ft Load Combination: 4.1.2(4) P = 81 kips /ft M(first) = 214 ft-kips /ft M(mid) = 85 ft-kips /ft M(last) = 151 ft-kips /ft Out-of Plane Shear: The critical load combination for out-of-plane shear is: Load Combination: 4.1.2(3) The shear varies linearly from -94 kips /ft to +88 kips /ft. Outer Dish: Radial Section (Outside the Crane Wall) Tension / flexure: The critical load combinations for maximum moment with tensile axial forces are: Load Combination: 4.1.2(1) P = 40 kips /ft M = 81 ft-kips /ft Load Combination: 4.1.2(4) P = 24 kips /ft M = 127 ft-kips /ft O Apnneoved Design Meteriel. Ssipn of SSC Page 3.88-12
ti System 80+ Deslan controlDocument l
' Compression / flexure:
The critical load combinations for maximum moment with compressive axial forces are: Load Combination: 4.1.2(1) . P = 68 kips /ft M = 158 ft-kips /ft Load Combination: 4.1.2(1) 3 P = 49 kips /ft M = 123 ft-kips /ft Load Combination: 4.1.2(3) P = 24 kips /ft M = 88 ft-kips /ft Load Combination: 4.1.2(3)
' P = 107 kips /ft M = 292 ft-kips /ft Out-of-Plane Shear:
I The critical load combination for out-of-plane shear is: n- ~ 'h - Load Combination: 4.1.2(3) V = 54 kips /ft Slab at Elevation 91'-9" Tension / flexure:
- - All load combinations result in compressive axial forces.
Compression / flexure: , The critical load combination for maximum moment with compressive axial force is: Load Combination: 4.1.2(1) P = 4 kips /ft M = 305 ft-kips /ft
- Out-of-Plane Shear:
The critical load combination for out-of-plane shear is: Load Combination: 4.1.2(1) t \ V V = 66 kips /ft z--...: w mem w ce + erssc rose 3.as-ts 1
- w
System 80+ Design ControlDocument In-Plane Shear: The in-plane shear for this member is obtained from the global three-dimensional finite element model results. The applicable forces are combined and transposed into the local coordinate system. The critical load combination is: Load Combination: 4.1.2(4) V = 165 kips /ft Radial Wall The in-plane shear and axial forces for this member are obtained from the global three-dimensional finite element model results. The total forces are obtained by combining the critical loads for several elements and transposing them into local coordinates. The forces envelope the critical load combinations by taking the largest forces for each element. The in-plane moment for this member is calculated as the product of the in-plane shear times the height of the wall. The out-of-plane moment, determined by local analysis, is much smaller than the in-plane moment (0.4%) and is neglected. Tension / flexure: The critical load combinations for maximum moment with tensile axial force are: Load Combinations: 4.1.2(1) and 4.1.2(4) P = $610 kips M = 82,110 ft-kips Compression /ficxure: The critical load combinations for maximum moment with compressive axial force are: Load Combinations: 4.1.2(1), and 4.1.2(4) P = 8262 kips M = 82,110 ft-kirs Out-of-Plane Shear: The critical laad combination for outef-plane shear is: Load Combination: 4.1.2(1) V = 2 kips /ft in-Plane Shear: The critical load combinations for in-plane shear are: Load Combinations: 4.1.2(1), and 4.1.2(4) V = 84 kips /ft Approved Design MaterW Design of SSC Page 3.88-14
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1 System 80+ Design ControlDocument Lower Crane Wall The in-plane shear for this member is obtained from the global three-dimensional model results. The applicable forces are combined and transposed into the local coordinate system. The in-plane moment is calculated as the product of the in-plane shear times the height of the wall. The out-of-plane moment, determined by local analysis, is increased by assuming an eccentricity of the axial load to account for possible misalignment of the upper and lower crane walls. Tension / flexure: j l The critical load combinations for maximum moment with tensile axial force are: Load Combinations: 4.1.2.(1),4.1.2(3), and 4.1.2(4) : P = 6800 kips M = 274,250 ft-kips Compnession/ flexure: The critical load combinations for maximum moment with compressive axial force are: Load Combinations: 4.1.2(3), and 4.1.2(4) P = 12,250 kips M = 274,250 ft-kips Q Out of Plane Shear: The critical load combination for out-of-plane shear is: t Load Combination: 4.1.2(3) j V = 64 kips /ft In. Plane Shear: The critical load combination for in-plane shear is: , Load Combination: 4.1.2(4) V = 177 kips /ft Pedestal The shear and axial forces for this member are obtained from the static three-dimensional finite element model results. The total forces are obtained by combining the critical loads for several elements. The forces envelope the critical load combinations by taking the largest forces for each element. Because of the extremely low height to width ratio and the large horizontal shear, the pedestal is treated as a corbel l
'and the design is based on Section 11.9 of ACI 349-90.
1 Annred Doespn 60eserial- Desigrr of SSC Page 3.88-r5 >
System 60+ Design controlDocument The critical load combinations are: Load Combinations: 4.1.2(1), and 4.1.2(4) F(horizontal) = 116,314 kips P(compression) = 136,840 kips P(tension) = 34,210 kips 5.4.4 Analysis Methods The outer dish transfers load to the pedestal, lower crane wall, radial walls, and the 91'+9" slab by two-way slab action. Two perpendicular one foot wide strips are considered to represent this structure. These strips are modelled as beam elements in a two4imensional frame model analysis. One model is a typical section of the outer dish and its supponing structure, cut radially from the pedestal to the reactor building shell wall. Perpendicular to this model is a typical section of the outer dish and its supponing radial walls, cut around the dish. A third model represents a typical section of the inner dish, spanning continuously around the inner dish, directly inside the outer dish strip. In general, in-plane forces are derived from the static three-dimensional finite element model results, and out of-plane forces from local frame model results. These forces are combined, as appropriate, to obtain the values used for design. The radii of the curved members are such that arc length and chord length are essentially equal. Therefore, the members are considered to be straight between suppon points for modelling purposes, i.e., initial curvature is neglected. 5.4.5 Typical Reinforcing Details Outer Dish Iloop steel (around the dish): 1 layer # 14's at 8" o.c. each face. These bars are mechanically spliced into continuous rings. Ties:
- 5 Closed ties around the hoop steel are placed 6" o.c. for 72 inches on both sides of each supponing wall, and 4" o.c. within the dish above the wall.
Radial steel (up the dish): 1 layer # 14's at 12" o.c. each face at the top of the dish, reducing to 5.2" o.c. at the bottom end of the dish. A second layer directly over the first from a point 7.31 feet up from the bottom of the dish to a point 8.25 feet up from the outside face of the crane wall. wmr any wraw. onion or ssc roue sse-se
3 System 80+ oesign controlDocument , fm . Q The first layer on both faces extends into pedestal where half of the bars turn down in a standard hook at 33" into the pedestal for the bottom face and 39" into the pedestal for the top face. The remaining bars : in both faces are straight bars 105" into the pedestal. The first layer on both faces extends into the floor slab at elevation 91'-9" where the bars are mechanically spliced to the slab bars. : Ties:
# 5 Closed hoops around the radial steel and # 5 stirrups with the following spacings:
From the hook on the lower end to the face of the pedestal: 8 closed hoops @ 4" o.c. From the face of the pedestal: I closed hoop @ 2" 14 closed hoops @ $ 1/2" o.c. 25 stirrups @ 11 1/2" o.c. 14 closed hoops @ 51/2" o.c. 16 closed hoops @ '4" o.c., and 21 closed hoops @ 61/2" o.c. % 91'+9" Slab i Radi=1 steel: 1 layer # 14's each face, spaced 12" o.c. at the top of the dish increasing to 13.2" o.c. over the supporting wall at radius 84 feet. These bars are mechanically spliced to the dish bars and continue away from the center of the reactor building. Ties:
# 5 stirrups @ 16.5" o.c. on each radial bar.
Circumferential steel:
# 11's @ 12" o.c. each face. .
- These bars are tapped spliced to form continuous rings.
Radial Wall Vertical steel: 2 layers # 11's @ 12" o.c. cach face.
^
_ a: Deelyn nesteniel- Desigre of SSC Pope 3.8s-17
System CO + Design ControlDocument Bars extend into the basemat a development length. At the top, these bars extend to the far face of the dish where they end in a standard hook. Ties:
# 4 closed ties @ 4" o.c. along each vertical bar for a distance from the dish at the top and from the basemat at the bottom equal to 1/6 the span but at least 36 inches. # 4 stirrups @ 15" o.c. for the remainder of the wall.
Ilodzontal steel:
# ll's @ 12" o.c. each face.
At the crane wall, these bars extend to the far face where they end in a standard hook. At the pedestal, these bars penetrate a development length past the first ring of vertical bars. Some of these bars terminate in the dish where they end in a standard hook on the far face. Lower Crane Wall Vertical steel: 2 layers # ll's @ 10" o.c. each face. These bars extend into the basemat a development length. At the top, these bars extend to the far face of the dish where they end in a standard hook, one layer each face turning away from the building center in a 90 degree hook, the other layer turning toward the building center in a 180 degree hook. Ties:
# 4 closed ties @ 4" o.c. along each vertical bar for 48 inches down from the dish,48 inches up from the basemat, and from 48 inches above the top of the 70'-0" slab to 48 inches below the bottom of the slab. # 5 stirrups @ 121/2" o.c. along the remainder of the wall.
IIorizontal steel:
# ll's @ 10" o.c. each face.
Lap splice bars to form continuous rings. O Approvenf Des / prs Acatorial.Desigre of SSC Pope .3.88-18
. System 80+ Desian controlDocument .(%J n) Pedestal ,
Vertical steel:
# 18 bars are placed in concentric rings distributed throughout the pedestal. The radius, spacing, (in ;
inches) and the number of bars are: ! f Ring Quantity Radius Spacmg 1 120 392 20.53 2 % 368 24.09 j 3 96 344 22.51 ! 4 % 320 20.94 5 80 2% 23.25 : 6 80 272 21.36 l 7 74 248 21.06 8 66 224 21.32 9 60 200 20.94 ! 10 52 176 21.27 l 11 46 152 20.76 12 38 128 21.16 13 32 104 20.42 14 24 80 20.94 i Total 936 These bars are anchored at the top by welding to a 13/4" x 9" x 10" plate. At the bottom, these bars penetrate the full depth of the basemat, which is more than a development length for a # 18. Ties: L
# 14's @ 8" o.c. around the outer ring from the dish down to the basemat.
Hodrontal steel:
# 9's @ 12" o.c. both directions. These bars cover tl.e top of the pedestal. At the edge, the hoop bars from the top face of the dish are continued into the pedestal to cover an embedment length on the ends ,
of the # 9's. 5.4.6 Conclusions The Areas 4,6 and 7 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. It is feasible to design and construct the structural components t Page 3.8619
~
.* ; . 2 % nietenia! Deeipur of SSC l
System 80+ Design ControlDocument considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. In both the analysis and design of these components, simplifying assumptions are made which result in a conservative design. A more detailed analysis and design would likely result in a more economical design for some components, but in no case would the capacity as given be insufficient. All components are constructible as detailed herein, therefore, any design resulting from more detailed analyses will be constmetible. 5.5 Areas 5A,5B and SC - South Section of Shield Building Wall 5.5.1 Description of Areas Area SA includes the south section of the Shield Building wall at the intersection of the North-South shear wall at column line K. It extends from the basemat foundation at elevation 50'+0" to the roof at elevation 156'+0". The Shield Building wall is a four feet thick right cylinder up to the top of the Nuclear Annex roof where the thickness ir, decreased to three feet. The Shield Building has an inside radius of 105 feet. Area 5B includes a portion of the East-West shear wall Dng column line 11 where it intersects with column line K. This wall extends from the basemat foundat on at elevation 50'+0" to the floor slab at elevation 130'+6". The portion between elevations 106'+0" and 130'+0" encloses the Main Steam Valve House (MSVII). The portion between elevations 70'+0" and 104'+0" encloses the Emergency Feedwater (EFW) Tank. This wall is 4 feet thick. Area SC includes a portion of the North-South shear wall along column line K where it intersects with the Shield Building wall and the East-West shear wall at column line 11 (Area 5B). This wall extends from the basemat at elevation 50'+0" to the floor slab at elevation 91'+9" inside the Shield Building, and to the floor elab at elevation 106'+0" outside the Shield Building. This wall is 6 feet thick inside the Shield Building below elevation 70' +0" and four feet thick elsewhere. A portion of the wall encloses the EFW Tank between elevations 70'+0" and 104'+0" similar to Area 5B. 5.5.2 General Loads The loads applicable to Areas SA, SB and SC are summarized in Appendix 3.8A, Table 10-1. These loads include hydrodynamic effects from the Emergency Feedwater tanks, pressurization of the Main Steam Valve House and pipe reaction loads from a potential Main Steam or Main Feedwater line ruptures. 5.5.3 Goveming Load Combinations Area SA Shear (in-plane) 4.1.2(4) (out-of-plane) 4.1.2(2) Bending 4.1.2(2) Axial (tension) 4.1.2(4) (compression) 4.1.2(1) Approved Des}pn Meterint. Design of SSC Pope 3.88-20
4 h System 80+ Danfgrs CorrealDocumerst q
'] Area 5B ' j i
Shear (in-plane) 4.1.2(4) l 4- (out-of-plane) . 4.1.2(1) l
- Bending '4.1.2(3)
Axial (tension): 4.1.2(4)
,(compression) '4.1.2(1) :
Area SC - ~ Shear (in-plane)' 4.1.2(4) 1 (out-of-plane) 4.1.2(2)
; Bending - 4.1.2(1)
; Axial (tension) 4.1.2(4)
- (compression) '4.1.2(1) ,
5.5.4 Analysis Methods and Results ,
'I '
- i. .The in-plane loads on Area 5A are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three dimensional finite element model. The out-of-plane loads on the wall are predominantly bending loads from the SSE and i
\s Main Steam' pipe reactions. The out-of-plane resultants and moments are determined by hand l
p calculations. The design forces and moments for Area 5A are: Shear (in-plane) 187 kips /ft (out-of-plane) ~ 65 kips /ft i
. Momer t 242 ft-kips /ft
- . Axial (tension) 86 kips /ft l l . (compression) 350 kips /ft 4
The in-plane loads on Area 5B are predominantly shear loads from the SSE. The in-plane forces are obtained from' output computed by the . application of these loads to the static three dimensional finite " element model. The out-of-plane loads in the Main Steam Valve House (MSVH) area are predominantly
- SSE and pressure / temperature loads from potential pipe ruptures in the MSVH. The out-of-plane loads l in the Emergency Feedwater Tank (EFW) area are predominant SSE induced hydrodynamic loads. The i out-of-plane resultants and moments are determined by hand calculations.' i
.i
' The design forces and moments for Area 5B are: ;
Shear L (in-plane) > 271 kips /ft a O. (out of-plane) ; 48 kips /ft ' : I
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System 80+ Design ControlDocument Moment 411 ft-kips /ft Axial (tension) 147 kips /ft (compression) 283 kips /ft The in-plane loads on Area 5C are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three-dimensional finite element model. The out-of-plane loads on the wall are predominantly SSE loads and SSE induced hydrodynamic loads from the EFW tank. The out-of-plane resultants and moments are determined by hand calculations. The design forces and moments for Area SC are: Shear (in-plane) 400 kips /ft (out-of-plane) 45 kips /ft Moment 528 ft-kips /ft Axial (tension) 107 kips /ft (compression) 334 kips /ft 5.5.5 Typical Reinforcing Details Area SA Wall Thickness 4 feet (MSVH area)
#18 at 12" vertical steel,2 layers each face #18 at 12" horizontal steel, 2 layers each face Shear ties - #5 horizontal ties at 4" x 12" Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
Area 5B Wall Thickness 4 feet
#18 at 12" vertical steel each face #18 at 12" horizontal steel each face Shear ties - #4 horizontal ties at 4" x 12" Additional ductility remforcing shall be provided as described in Section 6.0 of this appendix.
AntaSC Wall Thickness 6 feet (Inside Shield Building From Top of Basemat to Bottom of Slab at Elevation 70'+0")
#18 at 12" vertical steel,2 layers each face #18 at 12" horizontal steel, 2 layers each face Shear ties - #4 horizontal ties at 4" x 12" Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
Approwd Design Material- Design of SSC Page .i.88-22
Sy tem 80+ Deskn control Document 1 5.5.6 Conclusion The Area 5 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. It is feasible to design and construct the structural components considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific l site conditions, within those parameters. 1 The main steam line piping is assumed to be 32 inches in diameter. The main steam line anchor is ) assumed to have an 80 inch diameter bearing plate. A minimum separation of 6' 6" from the centerline l of the main steam line to any other discontinuity such as a wall, slab, opening or other possible failure plane should be maintained. Any separations less than 6'6" shall be analyzed and designed on a case by case basis. The 6'6" distance l is the radius of the shear failure cone with the bearing plate assumed, 5.6 Area 8 - Steam Generator Enclosure Radial Wall 5.6.1 Description of Area Area 8 includes the Steam Generator Compartment wing wall inside the in-containment Refueling Water Storage Tank (IRWST) below elevation 91'+9". This wall is 3 feet thick and is supported on three sides by the floor slab at elevation 91'+9", the dish structure which forms the bottom of the IRWST, and the vertical side wall of the IRWST. The wing wall is skewed 45 degrees from the East-West and North-South direction. 5.6.2 General Loads l The loads applicable to Area 8 are summarized in Appendix 3.8A, Table 3.8A-1. These loads include hydrodynamic effects in the IRWST. 5.6.3 Governing Load Combinations Area 8 Shear (in plan:) 4.1.2(4) (out-of-plane) 4.1.2(4) l Bending 4.1.2(4) Axial (tension) 4.1.2(4) (compression) 4.1.2(4) 5.6.4 Analysis Methods and Results The in plane loads on Area 8 are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three-dimensional finite element model. The out-of plane loads on the wall are predominantly bending loads from the SSE and IRWST pressure transients. The out-of-plane forces and moments are determined by hand calculations. L = o n on neeeu w . neeon erssc rare 2.ss-22
System 80+ Design ControlDocument The design forces and moments for Area 8 are: Shear (in-plane) 228 kips /ft (out-of-plane) 63 kips /ft Moment 320 ft-kips /ft Axial (tension) 29 kips /ft (comprest, ion) 201 kips /ft 5.6.5 Typical Reinforcing Details Area 8 Wall Thickness 3 feet
- 18 at 12" vertical steel,2 layers each face
- 18 at 12" horizontal steel each face Shear ties - #4 horizontal ties at 4" x 12" Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
5.6.6 Conclusion The Area 8 concrete section strengths deter.ained from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. It is feasible to design and construct the structural components considered. The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. 5.7 Area 9 - Spent Fuel Pool Wall 5.7.1 Description of Area Area 9 is the wall between the spent fuel pool and the refueling canal, at Elevation 104'+0" to 146'+0", Column line 17-18 @ Column line T. The wall provides a barrier to isolate the spent fuel pool from the fuel transfer canal to allow maintenance on the fuel transfer system. A weir gate in the wall is removed to transfer fuel between the spent fuel pool and the refueling canal. 5.7.2 General Loads The loads applicable to Area 9 are summarized in Appendix 3.8A, Table 3.8A-1. These loads include hydrodynamic and thermal loads from the spent fuel pool. 5.7.3 Governing Load Combinations Area 9 Shear (in-plane) 4.1.2(4) (out-of-plane) 4.1.2(4) O Altroved Desspn Materal Design of SSC Pege 3.88 24
System 80+ oesian controlDocument }
\
Q Bending 4.1.2(4) Axial . (tension) 4.1.2(4) i (compression) 4.1.2(4) l 5.7.4 Analysis Methods and Results The in-plane loads on Area 9 are predominantly shear loads from the SSE. The in-plane forces and moments are obtained from the global static three-dimensional finite element model results. Out-of-plane forces and moments are obtained by applying the out-of-plane loads to a local static three-dimensional finite element model of the wall. These forces are then considered in conjunction with the l loads from the global finite element model results to determine design forces and moments for the wall. , Horizontal reinforcing is designed for the maximum out-of-plane bending about a vertical axis, due to local loads. Vertical reinforcing is designed for the maximum out-of-plane bending about a horizontal l axis, due to local loads, combined with the maximum tension produced by global loads. Out-of-plane ! shear is determined by local analysis, i The predominant forces are out-of-plane shear and bending forces from the hydrostatic and inertial forces associated with the water in the spent fuel pool, inclading sloshing effects. Also significant are the ! thermal effects from the heat generated by spent fuel. l 1 The design forces and moments for Area 9 are: Shear .in-plane) ( 282 kips /ft (out-of-plane) 179 kips /ft l Moment (2 way bending) 2704 ft-kips /ft 4807 ft-kips /ft Axial ~ (tension) 80 kips /ft ) 5.7.5 Typical Reinforcing Details Area around weir gate notch in wall controls. Area 9 Wall Thickness 6 feet
#18 at 8" vertical steel,2 layers each face
#18 at 6" horizontal steel,3 layers each face Shear ties - #5 horizontal ties at 18.5" x 6" Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
5.7.6 Conclusion The Area 9 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to
' resist the design basis loads. It is feasible to design and construct the structural components considered.
L .::Dee@ ateennd. Design et SSC Page .T.88-25
System 80+ Design ControlDocument The assumptions envelope the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. 5.8 Area 10 5.8.1 Description of Area Area 10 includes the Main Steam Valve House exterior wall along column line H between cohunn lines 23 and 25 and elevations 106'+0" and 130'+0" on the North, side of the building. This wall is 5 feet thick and includes piping penetrations / anchors for the Main Steam and Main Feedwater Systems. 5.8.2 General Loads The loads applicable to the Area 10 wall are given in Table 3.8A-1 of Appendix 3.8A. These loads include Main Steam and Feedwater piping anchor loads including pipe rupture loads. short duration temperature effects of 300*F from main steam line rupture. and internal compartment pressurization due to a potential Main Steam line rupture. 5.8.3 Governing Load Combinations Area 10 Shear (in-plane) 4.1.2(4) (out-of-plane) 4.1.2(4) Bending 4.1.2(4) Axial (tension) 4.1.2(4) (compression) 4.1.2(1) 5.8.4 Analysis Methods and Results The in-plane loads on Area 10 are predominantly shear loads from the SSE. The in-plane forces are obtained from output computed by the application of these loads to the static three-dimensional finite element model. The out-of-plane load on the wall is predominantly bending from SSE loads and pipe reaction loads from potential Main Steam line ruptures. The out-of plane resultant forces and moments are determined by hand calculation. The design forces and moments for Area 10 are: c Shear (in-plane) 319 kips /ft (out-of-plane) 358 kips /ft Moment 1591 ft-kips /ft Axial (tension) 187 kips /ft (compression) 246 kips /ft O 44weved Des # Materal . Des & of SSC Pope 188-26
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; .3 5.8.5 Typical Rainforcing Details Lj Area 10 - ~ Wall 11h=== 5 feet- k i
- _ #18 at.12" venical steel, 2 layers each face ]
#18 at 12" horizontal steel, 3 layers each face Shear ties - 2 #5 horizontal ties at 4" x 12"
' Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix. :
5.8.6 - Ca=ria lan ; 1 The Area 10 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to - f resist the design basis loads. It is feasible to design and construct the structural components considered. '; The assumptiocs envelope the given parameters so that the design presented is adequate for any specific l site conditions,'within those parameters. To accamm Ise punching shear requirements, the anchor ')
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embedment design must incorporate excess punching shear, or alternately rupture loads may be reduced
- by more detailed analysis. 'l i
7 The main steam line piping is assumed to be 32 inches in diameter. The main steam line anchor is ! i: assumed to have an 80 inch diameter bearing plate. A minimum separation of 7'6" from the centerline i of the main steam line to any other discontinuity such as a wall, slab, opening or other possible failure _ i i plane should be maintained. The main feedwater !ine located in this area is assumed to be 24 inches in 4
- diameter and contains a 72 inch diameter bearing plate. The minimum separation of the main feedwater centerline to the ec'ge of any other discontinuity is 7'0".
1 Any separation less than 7'6" for the main steam lines and 7'0" for the main feedwater line is analyzed - , and designed on a case by case basis. These distances are the radii of the shear failure cones of the main , steam line and main feedwater line with the bearing rings assumed. ! l r 5.9 Area 11.- Nonth-West End Wall l 5.9.1 Description of Area j Area 11 is the' nonhern part of the west side end wall adjacent to the Radwaste Building. The wall } extends from the top of the basemat at elevation 50'+0", to the top of the roof in the Fuel Handling area j [ at elevation 191'+ 0". The walls in this area are four feet thick. Out-of-plane lateral support is provided to the walls by the floor slabs on the interior of the structure. The wall is arranged and designed to l function as a major structural shear wall in addition to providing protection for the safety related i equipment. q 1
~5.9.2 General Loads s The loads applicable to Area 11 are summarized in Appendix 3.8A, Table 3.8A-1. The out-of-plane
. passive soil pressure loads are the predominant loads in the lower elevations of the wall. The Nuclear Island evaluation credits the passive soil pressure loads to resist sliding. Local loads resulting from the
; wall mounted suppons for the spent fuel pool bridge crane are also included.'
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System 80+ Design ControlDocument 5.9.3 Governing Load Combinations Area 11 Shear (in-plane) 4.1.2(1) (out-of-plane) 4.1.2(1) Bending 4.1.2(1) Axial (tension) 4.1.2(1) ) l (compression) 4.1.2(1) 5.9.4 Analysis Methods and Results The Area 11 wall is analyzed as a stmetural shear wall. The in-plane forces are obtained from output computed by the application of these loads to the static three-dimensional finite element model. The out-i of-plane loads on the wall are predominantly soil pressure loads with the effect of the SSE in the lower elevations. The effects of the spent fuel pool bridge crane loads in combination with thermal loads are predominant in the upper portion of tie wall. The out-of plane resultant forces and moments are determined by local two dimensional frame models. The wall is analyzed and designed to resist the spent fuel pool bridge crane bending and axial loads. The bending effects dissipate below elevation 91'+9" The design forces and moments, excluding the spent fuel pool bridge crane loads, for Area 11 are: Shear (in-plane) 274 kips /ft (out-of-plane) 183 kips /ft Moment 938 ft-kips /ft Axial (tension) 240 kips /ft (compression) 512 kips /ft 5.9.5 Typical Reinforcing 9etails Area 11 Main Steel Wall Thickness 4 feet
- 14 at 12" vertical steel 2 layers each face (above elevation 90'+3")
#18 at 12" vertical steel 2 layers each face (below elevation 90'+3")
- 11 at 12" horizontal steel 2 layers each face Shear ties - #5 horizontal ties at 12" x 12" Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix.
O Alveoved Design Mosenind
- Design of SSC Page 3.88-28
i Sy-tem 80 + Design ControlDocument n 5.9.6 Conclusion (] The Area 11 concrete section strengths determined from the criteria in Appendix 3.8A are sufficient to I resist the design casis loads. l 5.10 Area 12 - Interior Structure Steel Column l l 5.10.1 Description of Area l Area 12 includes a total of 48 steel columns which are arranged in a circular pattern within the steel containment vessel outside the polar crane wall. These columns are supported by a floor slab at elevation 91'+9" and in turn support the slab at 115'+6". The top of each column is located 15.5 feet farther out on a radius than the bottom of the column placing the column at an angle. This arrangement results in a compression load on the bottom slab and a tension load in the top slab from the column. The edge of the elevation 115'+6" slab where the top of the column is attached is close to the steel containment vessel making the integrity of the slab and column critical. 5.10.2 General Loads The loads applicable to the Area 12 columns are given in Table 3.8A-1 of Appendix 3.8A. These loads include operating and accident thermal loads. 5.10.3 Governing Load Combinations /\ V Area 12 Bending 4.2.2(4) Axial (compression) 4.2.2(4) 5.10.4 Analysis Methods and Results The columns have pinned end connections. The end forces are obtained from the static three-dimensional finite element model results with the exception of the thermal reactions which are calculated by hand methods and superimposed in the appropriate load combinations. The moments in the column resulting from seismic and dead load are also calculated by hand methods. Due to the flexibility of the column, the peak accelerations from instructure response spectra are used for the inertia response from the SSE loads. The design forces and moments for Area 12 are: Moment (Biaxial) 68 ft-kips 46 ft-kips Axial (compression) 1119 kips p G 4erms one en www. onion or ssc rose ras-2s
i i l System 80+ oesign controlDocument ! l 5.10.5 Design Details A W14x257 A36 steel column is determined to be adequate for the potential forces and moments in the columns. Both end connections are pinned connections which allow rotation in a vertical plane coincident with the axis of the column and a containment radial through the column. The connections at the slab are loaded with horizontal tension and compression forces of approximately 664 kips. The compression load is transferred into the lower slab, at elevation 91' +9" with an embedded kicker plate assembly. The tension load is transferred into the upper slab, at elevation 115'+6" with an embedded tie back assembly. 5.10.6 Conclusion The Area 12 W14x257 A36 columns designed from the criteria in Appendix 3.8A are sufficient to resist the design basis loads. The connections details can be designed such that the load can be transferred into the slabs effectively without exceeding the design capacity of the slabs. 5.11 Area 13 - Nuclear Island Basemat 5.11.1 Description of Area This area is the foundation basemat. It is ten feet thick, and supports the entire Nuclear Island. The top of the basemat is at elevation is 50'-0". 5.11.2 General Loads The loads applicable to the Area 13 basemat are given in Table 3.8A-1 of Appendix 3.8A. 5.11.3 Governing Load Combinations Area 13 Shear 4.1.2(4) Bending 4.1.2(4) 5.11.4 Analysis Methods and Results The Area 13 basemat responds in bending and shear loads from the SSE and dead load. The forces and moments are obtained from apphcation of these loads to the static three-dimensional finite element model. The moment evaluated is the maximum moment experienced by the basemat. The shear evaluated is from a representative area under a primary shear wall. Most of the basemat will not require any shear reinforcing. The design forces and moments for Area 13 are: Moment: 3545 ft-kips /ft Shear: 68.3 kips /ft Approved Design Material Design of SSC Page 3.8B-30
i'
' System 80+ Deslan controlDocument The basemat is symmetrically reinforced to resist the potential moments as a result of differential settlement of the foundation. The capacity of the basemat to withstand differential settlement is ;
determined by calculating the deflection at the edge of the mat that would occur if the maximum moment were developed in the center. The maximum deflection in the basemat relative to the ce iter of the Nuclear Island at the four exterior walls is: Wall Delta North 20 in South 20 in . East 25 in West 49 in 5.11.5 Typical Reinforcing Details Basemat Thickness 10 feet l
#18 at 12" horizontal steel,2 layers each face each direction Shear ties - (When required) ;
#10 vertical ties at 12" x 12"
, i Most of the basemat will not require any shear reinforcing.
Additional ductility reinforcing shall be provided as described in Section 6.0 of this appendix. 5.11.6 Conclusions The Area 13 concrete basemat strength determined from the criteria in Appendix 3.8A is sufficient to resist the design basis loads. It is feasible to design and construct the nuclear island foundation basemat. The design envelopes the given parameters so that the design presented is adequate for any specific site conditions, within those parameters. Stress concentrations exist in the areas around sumps that require additional detailed analyses and design. The maximum differential settlements that can be tolerated by the basemat are calculated based on the moment capacity. Any settlements less than those shown are acceptable from the standpoint of stress in the basemat. A settlement monitoring program ensures that proper consideration is given to actual settlements during and after construction. 6.0 Typical ACI-318 Chapter 21 Ductility Connection Details The System 80 + design incorporates ACI-318 Chapter 21 ductility requirements as identified in Appendix 3.8A, Section 6.2.1.1. O Figure 3.8B-2 is provided as a supplement to illustrate the ductility steel requirements in Appendix 3.8A. (d Typical details are shown with a description and a reference to the ACI-318 Code section associated with L a:ouw noww. onw or ssc roue 3.as.sr c
Sy~ tem 80 + Design ControlD=curnent the detail. The actual spacing, dimensions, reinforcing bar sizes,. bend angles, etc., are obtained from the .ACI-318 Code sections. The details provided are for illustration purposes only. 7.0 Non-Nuclear Island Structures 7.1 Diesel Fuel Storage Structure 7.1.1 Description of Structure The Diesel Fuel Storage Structure is a two bay, partially embedded, sing;e-story reinforced concrete building; symmetrical about its north-south axis. Each bay houses a single diesel fuel oil tank. The specified concrete compression strength is 4,000 psi and the specified minimum yield strength of the reinforcing steel is 60,000 psi. 7,1.2 Analysis Methods The Diesel Fuel Storage Structure is analyzed for the design loads described in Appendix 3.8A to determine the global and localized member forces for which the structure must be designed. The structure is analyzed using a linear elastic three-dimensional finite element flat plate type model supported on clastic soil springs. Thermal and equivalent static loads corresponding to the various individual loading conditions identified in Sections 3.8A.5.1 and 3.8A.11.1.5 are applied to the structure model and the resulting member forces and moments computed. The resulting member forces are combined in accordance with the load combinations, specified in Sec'. son 5.2.2 of Appendix 3.8A, to determine the design loads for the critical sections. 7.1.3 Loads and Load Combinations The Diesel Fuel Storage Structure is evaluated for the loads and load combinations specified in Sections 3.8A.5.1 and 3.8A.5.2, respectively, for Seismic Category I concrete structures. The major loadings affecting the design of the structure are dead loads (i.e., self weight and equipment weight from the diesel fuel storage tanks), temperature, static and dynamic lateral soil and ground water pressures, wind loads, earthquake loads, and tornado loads. The cAtic~.) loed combinations are equations 5.2.2.1 (1st eqn.),5.2.2.1 (4th eqn.), and 5.2.2.2 (1st egn.) of Section 3.8h.5.2, i.e., U = 1.4D + 1.7L U = 0.75 (1.4D + 1.7F + 1.7L + 1.7H + 1.7To + 1.7R o) U = D + F + L + H + To + R, + E' 7.1.4 Analyses and Results The reinforced concrete members of Seismic Category I structures are designed to the criteria specified in ACI 349 and NRC Regulatory Guide 1.142, except as modified by Appendix 3.8A (see 3.8A.6.2). Approved Design Materia!- Design of SSC Page 3.88-32
Sy~ tern 80+ Design Control Document ,m ( " ) In general, symmetrical reinforcing steel (i.e., the same area and configuration on opposite faces of members), is provided except in local areas. Concrete joints shall be detailed in accordance with the criteria specified in ACI 318, Chapter 21 (see Section 3.8A.6.2.1.1.1 and Section 6.0 of this appendix). Foundation Mat: The primary flexural reinforcing for the 2' - 3" thick foundation mat consists of a rectangular grid of #11 at 12 inches each way/each face, [i.e.,1.56 id/ft]. No transverse shear reinforcing is required. East and West Walls: The primary flexu al reinforcing for these 2' - 3" thick walls consists of a rectangular grid of #11 at 8 inches each way/each face, [i.e., 2.34 in2 /ft]. No transverse thear reinforcing is required. North and South Walls: The primary thural reinforcing for these 2' - 3" thick walls consists of a rectangular grid of #11 at 6 inches each way/each face, [i.e., 3.12 in2 /ft]. O Transverse shear reinforcing consisting of #6 at 12 inches is required in both directions for the entire wall area. The shear steel extends 9 feet down from the top of the roof. ( ,/ Center Wall: The primary flexural reinforcing for this two-foot thick wall consists of a rectangular grid of #11 at 6 inches each way/each face, [i.e.,3.12 in2 /ft]. Compression ties are required for the top half of the wall. No transverse shear reinforcing is required. Roof: The primary flexural reinforcing for these 2' - 3* thick walls consists of a rectangular grid of #11 at 6 inches each way/each face [i.e.,3.12 id/ft). No transverse shear reinforcing is required. 7.1.5 Conclusion The concrete and reinforcing steel section strengths of the Diesel Fuel Storage Structure are sufficient to resist the design basis load and load combination criteria specified in Sections 3.8A.11.1 and 3.8A.S.0. Typical reinforcing details are shown in Figures 3.8B-5 and 3.8B-6. /O s, Apiprowd Destyn Material Design of SSC Page 3.88-33
Sy^ tem 80+ Design ControlDocument 7.2 Component Cooling Water Heat Exchanger Structure 7.2.1 Description of Structure The Component Cooling Water Heat Exchanger Structure is a single bay, partially embedded, two-story reinforced concrete building. The top floor houses two heat exchangers supported on saddles which spread the loadings to the supporting floor and column system. The specified concrete compression strength is 4,000 psi and the specified . minimum yield strength of the reinforcing steel is 60,000 psi. 7.2.2 Analysis Methods The Component Cooling Water Heat Exchanger Stmeture is analyzed for the design loads described in Appendix 3.8A to determine the global and localized member forces for which the stmeture must be designed. The structure is analyzed using manual ec.mputations which consider the structure to be comprised of linear elastic one-way wall and slab panels. Thermal and equivalent static loads corresponding to the various individual Imding conditions identified M Sections 3.8A.5.1 and 3.8A.11.2.5 are applied to the one-way panel models and resulting member fores and moments computed. The resulting member forces are combined in accordance with the load combinations, specified in Section 5.2.2 of Appendix 3.8A, to determine the design loads for the critical sections. 7.2.3 Loads and Load Combinations The Component Cooling Water Heat Exchanger Structure is evaluated f or the loads and load combinations specified in Sections 3.8A.5.1 and 3.8A.5.2, respectively, for Seismic Category I concrete structures. The major loadings affecting the design of the structure are dead loads (i.e., self weight and equipment weight from the CCW heat exchangers), temperature, static and dynamic lateral soil and ground water pressures, wind loads, earthquake loads, and tornado loads. The critical load combinations are equations 5.2.2.1 (1st egn.), 5.2.2.1 (4th egn.), and 5.2.2.2 (1st egn.) of Section 3.8A.5.2, i.e., U = 1.4D + 1.7L U = 0.75 (1.4D + 1.7F + 1.7L + 1.7H + 1.7To + 1.7Ro) U = D + F + L + H + To + Ro + E' 7.2.4 Analyscs and Results The reinforced concrete rar.nbers of Seismic Category I structures are designed to the criteria specified in ACI 349 and NRC IF.:gulatory Guide 1.142, except as modified by Appendix 3.8A (see 3.8A 6.2). In general, symmetrical reinforcing steel (i.e., the same area and configuration on opposite faces of members), is provided except in local areas. Concrete joints shall be detailed in accordance with the ' criteria specified in ACI 318, Chapter 21 (see Section 3.8A.6.2.1.1.1 and Section 6.0 of this appendix). Appnewed Ues@n htatenaf Design of SSC Page 3.88-34
Sy: tem 80+ Design ControlDocument
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Foundation Mat: The primary reinforcing for the four-foot thick foundation mat consists of a rectangular grid of #9 at 10 inches each face, [i.e.,1.20 id/ft) in the long direction and #11 at 6 inches each face, [i.e., 3.12 id/ft.] in the short direction. No transverse shear reinforcing is required. East and West Walls (Short Direction): , The primary reinforcing for these 2' - 3" thick walls consists of a rectangular grid of #11 at 6 inches each face, [i.e., 3.12 in 2/ft] vertically and #11 at 10 inches each face, [i.e.,1.87 id/ft.] horizontally. , No transverse shear reinforcing is required. North and South Walls (Long Direction): The primary reinforcing for these 2' - 3" thick walls consists of a rectangular grid of #11 at 6 inches 2 vertically each face and #11 at 10 inches horizontally, [i.e., 3.12 id/ft and 1.87 in /ft, respectively). No transverse shear reinforcing is required. Floor Slab at Elevation 90'-9": q) The primary reinforcing for the 3'-0" floor slab consists of a rectangular grid of #10 at 10 inches each 2 face, [i.e.,1.52 in /ft) in the long direction and #11 at 10 inches each face, [i.e.,1.87 id/ft.] in the short , direction. No transverse shear reinforcing is required. Roof Slab at Elevation 110'-9": . The primary reinforcing for the 2' - 3" two-foot thick roof consists of a rectangular grid of #11 at 10 inches each way/each face [i.e.,1.87 id/ft). No transverse shear reinforcing is required. 7.2.5 Conclusion The concrete and reinforcing steel section strengths of the Component Cooling Water Heat Exchanger Structure are sufficient to resist the design basis load and load combination criteria specified in Section; 3.8A.11.2 and 3.8A.5.0. Typical reinforcing details are shown in Figures 3.8B-7 through 3.8B-9. 7.3 Component Cooling Water Tunnel 7.3.1 Description of Stnicture m The Component Cooling Water Tunnel is a single compartment, fully embedded, one-story reinforced concrete structure. The tunnel houses and protects the Component Cooling Water piping which is routed from the corresponding Nuclear Island pipe chase to the basement of the Component Cooling Water Heat Anpremt onien masaw- onion or ssc rose 3.ss-ss
Design ControlDocument ystqm 80+ Exchanger Structure. The tunnel is attached at one end to the Nuclear Island Pipe Chase and the Component Cooling Water Heat Exchanger Structure at the other end via flexible connections. The flexible connections allow differential movement between the three structures without transferring loadings. The specified concrete compression strength is 4,000 psi and the specified minimum yield strength of the reinforcing steel is 60,000 psi. 7.3.2 Analysis Methods The Component Cooling Water Tunnel is analyzed for the design loads described in Appendix 3.8A to determine the global and localized member forces for which the structure must be designed. The structure is analyzed using manual computations which consider the structure to be comprised of linear elastic one-way wall and slab panels. The lateral loads on the tunnel were evaluated using a linear clastic frame model with a unit width. Thermal and equivalent static loads corresponding to the various individualloading conditions identified in Sections 3.8A.5.1 and 3.8A.11.7.5 are applied to the equivalent frame model and resulting member forces and moments computed. The resulting member forces are combined in accordance with the load combinations, specified in Section 5.2.2 of Appendix 3.8A, to determine the design loads for the critical sections. 7.3.3 Loads and Load Combinations The Component Cooling Water Tunnel is evaluated for the loads and load combinations specified in Sections 3.8A.5.1 and 3.8A.5.2, respectively, for Seismic Category I concrete structures. The major loadings affecting the design of the structure are dead loads (i.e., self weight and equipment weight from the piping systems), AASHO H2044 truck overburden pressure, temperature, static and dynamic lateral soil and ground water pressures, tornado loads, and canhquake loads (inchiding seismic inertia and wave passage). Seismically induced forces due to differential movements are eliminated by providing flexible connections, at each end of the tunnel, which are capable of accommodating the movements without transferring loads. The critical load combinations are equations 5.2.2.1 (1st egn.),5.2.2.1 (4th egn.), and 5.2.2.2 (1st egn.) of Section 3.8A.S.2, i.e., i U = 1.4D + 1.7L l U = 0.75 (1.4D + 1.7F + 1.7L + 1.7H + 1.7T, + 1.7Ro ) l l U = D + F + L + H + To + Ro + E' 7.3.4 Analyses and Results The reinforced concrete members of Seismic Category I structures are designed to the criteria specified in ACI 349 and NRC Regulatory Guide 1.142, except as modified by Appendix 3.8A (see 3.8A.6.2). In general, symmetrical reinforcing steel (i.e., the same area and configuration on opposite faces of ; members), is provided except in local areas. Concrete joints shall be detailed in accordance with the l criteria specified in ACI 318 Chapter 21 (see Section 3.8A.6.2.1.1.1 and Section 6.0 of this appendix). Approved Destgre hinterint- Desigrs of SSC Page 3.88-36
Sy' tem 80+ Design ControlDocument r ( Foundation Mat: , The primary reinforcing for the three-foot thick foundation mat consists of a rectangular grid of #9 at 12 inches each way/each face, [i.e,1.00 in2 /ft). No transverse shear reinforcing is required. Walls: The primary reinforcing for these two-foot thick walls consists of a rectangular grid of #9 at 12 inches each way/each face, [i.e.,1.00 inz /ft). No transverse shear reinforcing is required. Roof: The primary reinforcing for these two-foot thick roof slabs consist of a rectangular grid of #11 at 10 inches each way/each face, [i.e.,1.87 inz /ft]. No transverse shear reinforcing is required. 7.3.5 Conclusion The concrete and reinforcing steel section strengths of the Component Cooling Water Tunnel are S j sufficient to resist the design basis load and load combination criteria specified in Sections 3.8A.11.7 and 3.8A.5.0. Typical reinforcing details are shown in Figures 3.8B-10 and 3.8B-ll. 4 v i 4 l i l j \ v 1 l Pn . .:: Design neeenriel- Deelour of SSC Page 3.88-37
System 80+ Design ControlDocument Table 3.8B-1 Areas Identified for Detailed Design Area Description Section Elevation Col. Line/ Azimuth 1 Shear & Shield Building Wall IA 50 to 115+6 D-F @ 17 1B 50 to 91+9 E17 IC 50 to 146 16-18. E-F 2 East Wall @ Turbine Building 2 50 to 146 B14 3 Diesel Gen. Room Ext & Int. Walls 3A 50 to 93 N23 3B 50 to 93 N25 4 Subsphere Radial Wall 4 50 to 80+9 225*, R33-R65 5 Shear Wall and Slab @ Emerg. FDW SA 50 to 156 K12 Pump Room and CCW Pump Room
$B 50 to 130+6 Kil SC 50 to 106 K10-K13 6 SCV Anchorage Region 6 70 to 91 +9 R76.5 7 SCV Support Pedestal 7 50 to 62 K17 & R33 8 S/G Wing Wall @ IRWST 8 70 to 91 +9 L15 9 Spent Fuel Refueling Canal Wall 9 104 to 146 T17-18 10 Main Steam Valve House Wall 10 106 to 130 H23-25 11 Nuclear Annex Wall @ Radwaste 11 50 to 191 U19-20 Building 12 Interior Structure Steel Columns 12 91 +9 to 112 N/A l
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System 80+ Design ControlDocument 3.9 Mechanical Systems and Components V ((The COL applicant referencing the System 80+ Standard Design will develop and/or provide site and , plant-specific information on mechanical systems and components design. Information will include:
- Verification of computer codes that are not precertified,
- Reconciliation of as-built with as-designed piping configurations,
- Documentation of pressurizer safety valve adjustments,
- Documentation of radiographic examination of welds,
- Identification of safety-related components which utilize snubbers,
- Description of energy absorbing and/or non-linear piping restraints, if used, and
- Qualification and preoperational testing of safety-related pumps and valves and a detailed pump and valve inservice testing (IST) program.))l 3.9.1 Special Topics for Mechanical Components ;
3.9.1.1 Design Transients The following information identifies the transients used in the design and fatigue analysis of ASME Code ,V Class I components, reactor internals and component supports. Cyclic data for the design of ASME , Code Class 2 and 3 components, as applicable, are discussed in Section 3.9.3. All transients are classified with respect to the component operating condition categories identified as Level A (Normal), B (Upset), C (Emergency) and D (Faulted) and testing as defined in the ASME Code, Section III. The transients specified below represent conservative estimates for design purposes only and do not purport to be accurate representations of actual transients, or necessarily reflect actual operating procedures; nevertheless, all envisaged actual transients are accounted for, and the number and severity of the design transients exceeds those which may be anticipated during the life of the plant. Pressure and temperature fluctuations resulting from the normal, test, upset, emergency and faulted transients are computed by means of computer simulations of the reactor coolant system, pressurizer, and steam generators. Design transients are detailed in the equipment specifications. The component designer then uses the specification curves as the basis for design and fatigue analysis. In support of the design of each Code Class 1 and CS component, a fatigue analysis of the combined effects of mechanical and thermal loads is performed in accordance with the requirements of Section III of the ASME Code. The purpose of the analysis is to demonstrate that fatigue failure will not occur when the components are subjected to typical dynamic events which may occur at the power plant. The rules of the ASME Boiler & Pressure Vessel Code, Section III, Division 1, Subsection NB: Class 1 Components, are used for performing fatigue evaluations of System 80+ components. The fatigue O I COL information item; see DCD Introduction Section 3.2.
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System 80+ Design ControlDocument curves (S-N Curves) contained in Appendix'l to Section til are used as the basis for performing all of the fatigue analyses of System 80+ components. Observations of significant environmental degradation of the cyclic behavior of materials in LWR environments are primarily related to high strain ranges, slow strain rates, high oxygen contents of LWR primary water environments, high sulfur contents of carbon and low alloy steels, and low flow rate conditions. The absence of any one of these conditions is sufficient to preclude any significant environmental degradation of the fatigue behavior of materials exposed to typical PWR primary coolant ' environment. Since System 80+ components are not exposed to high oxygen content environments at elevated temperatures, and no carbon or low alloy steel is directly exposed to the primary coolant, no significant environmental degradation of the cyclic behavior of System 80+ comoonents will occur. The existing S-N Curves are applicable to the 60 year design life of System 80+ components because:
- 1. the RCS, including all primary components, core support and internal structt.res, and the pressurizer surge line are either stainless steel clad materials, inconel or wrought stainless steel construction,
- 2. the primary system water chemistry controls require control of dissolved oxygen content in the primary system prior to operation above 150'F, and
- 3. no carbon or low alloy steel materials are exposed to the primary coolant environment.
The fatigue analysis is based upon a series of dynamic events depicted in the respective component specifications. Associated with each dynamic event is a mechanical, incrmal-hydraulic transient presentation along with an assumed number of occurrences for the event. The presentation is generally simple and straightforward, since it is meant to envelope the actual plant response. The intent is to present material for purposes of design. Similarly, the characterization of a given dynamic event with a specific name is unimportant. Any plant dynamic occurrence with consequences which fall within the envelopes associated with one of these dynamic events is by definition represented by that dynamic event. The fundamental concept ensures that the consequences of the normal and upset conditions which are expected to occur in the power plant are enveloped by one or more of the dynamic event portrayals in the component specifications. The number of occurrences selected for each dynamic event is conservative, so that in the aggregate, a 60-year useful life is provided by this design process. Design Ical wmeinatiora for ASME Code Class 1,2, and 3 components are given in Section 3.9.3. Design loading combinations for Code Class CS internals structures are presented in Section 3.9.5.2. The principal design bases of the reactor coolant system (RCS) and reactor internals structures are given in Sections 5.2 and 3.9.5, respectively. The System 80+ design basis initiating events and frequencies used in the stress analysis of primary system Code Class I and Class CS components are shown in Table 3.9-1. The System 80+ design basis events are nearly identical to System 80. The format has been revised to detail the individual transients considered in the design basis. A small number of events (e.g., frequency control) have been added in response to the EPRI Utilities Requirement Document (URD). The 60-year frequency of occurrence for all events was re-computed based on the latest industry databases (References 48 and 49). Particular emphasis was placed on severe transients (i.e., feeding cold feedwater to a hot steam generator) and/or Approved Design historial Design of SSC Page 3.9-2
I Svetem 80+ Design ContmlDocument transients with a high design frequency of occurrence (i.e., frequency control). Where appropriate, excessive conservatism has been removed from the System 80 values. The .Iting System 80+ events l 4 and frequencies conservatively represent the 60-year design basis. j
- The design basis events are classified as normal, upset, emergency, faulted and test. The normal and test l events are planned operations that will occur during the life of the plant. Upset events are occurrences ;
that may occur during the life of the plant (i.e., anticipated operational occurrences). Emergency and j
~ faulted events are not expected to occur (i.e., accidents) but are included in the design basis for additional '
design margin. The normal and test events are selected by reviewing the expected plant operations. The upset, emergency and faulted events are determined by reviewing industry databases (References 48 and 1
- 49) for events that have occurred, or that may be postulated to occur, based on observed plant behavior. l Normal and test event frequencies are determined by summing the number of expected plant operations l' over the 60-year design life. The frequencies for upset, emergency and faulted events are determined
> on a probabilistic basis utilizing industry databases (see References 48 and 49). The stated 60-year design frequency of occurrence (see Table 3.9-1) is always greater than the expected frequency of occurrence. t
- Conservative mathematical models and methodology are used to determine the thermal-hydraulic ,
consequences of the design basis events on individual plant components. The design margin is further enhanced by enveloping similar events and using the most conservative thermal-hydraulic consequences ,
- to represent a composite group. The group frequency is then determined by algebraically summing the individual design frequericies.
Pressure and thermal stress variations associated with the design transients are considered in the design j i . of supports, valves, and piping within the reactor coolant pressure boundary (RCPB). In addition to the design transients listed above and included in the fatigue analysis, the loadings produced by the SSE are also applied in the design of components and support structures of the RCS. For the number of cycles pertaining to fatigue effects of cyclic motion associated with the SSE, refer to Section 3.7.3.2. Design load combinations for ASME Class 1,2 and 3 components are given in Section 3.9.3. ; 3.9.1.2 Computer Programs Used in Stress Analyses : l 3.9.1.2.1 Code Class Systems, Components, and Supports The following paragraphs provide a summary of the applicable computer programs used in the stress and ; structural analyses for ASME Code Class systems, components, arxl supports in the CESSAR-DC scope. l The summaries include individual descriptions and applicability data. The computer codes employed in 1 these analyses have been verified in conformance with design control methods, consistent with the quality
- assurance program described in Chapter 17.
i If computer codes that are not listed below are used, they will be compared to NRC-benchmarked or l approved codes. \lComputer codes usedforpiping dynamic analysis will be benchmarked in accordance with NUREG/CR-6128.))2 See also Section 3.9. j i 2 ( NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction , Section 3.5. Anwowar Deewn neuerior- Deewn or ssc rope 3.s 3
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System 80+ Design C7ntrolDocument l 3.9.1.2.1.1 MDC STRUDL The MDC STRUDL computer program provides the ability to specify characteristics of framed structure and three-dimensional solid structure problems, perform static and dynamic analyses, and reduce and 9l! i i combine results. l l Analytic procedures in the pertinent portions of MDC STRUDL apply to framed structures. Framed ! structures are two- or three-dimensional structures composed of slender, linear members that can be ; represented by properties along a centroidal axis. Such a structure is modeled with joints, including support joints, and members connecting the joints. A variety of force conditions on members or joints l can be specified. The member stiffness matrix is cenguted from beam theory. The total stiffness matrix l of the modeled structures is obtained by appropriaten ambining the individual member stiffness. The stiffness analysis method of solution treats the joint displacements as unknowns. The solution procedure provides results forjoints and members. Joint results include displacements and reactions and joint loads as calculated from member end forces. Member results are member end forces and distortions. The assumptions governing the beam element representation of the structure are as follows: linear, clastic, homogeneous, and isotopic behavior, small deformation, plane sections remain plane, and no coupling of axial, torque, and bending. The program is used to define the dynamic characteristics of the structural models used in the dynamic seismic analyses of the reactor coolant system components. The natural frequencies and mode shapes of the structural models and the influence coefficients which relate member end forces and moments and support reactions to unit displacements are calculated. The influence coefficients are calculated for each dynamic degree-of-freedom of each mass point and for each degree-of-freedom of each support point. The ANSYS computer code (Section 3.9.1.2.1.13) is also used as an alternate to MDC-STRUDL for defining the dynamic characteristics of the reactor coolant system and seismically analyzing it. The program can perform either time-history analysis or spectrum anaiysis using the modal super position technique. Support reactions. member loads and joint acceleration are computed by back substituting from the modal coordinates to physical coordinates through the applicable transformation matrices and then combining modal contributions from each individual mode included in the response analysis. MDC STRUDL is a program which is commercially available and has had sufficient use to justify its applicability and validity. Extensive verification of the C-E version has been performed to supplement the public documentation. The version of the program in use at C-E was developed by the McDonnell Automation Company / Engineering Computer International and is run on the IBM computer system. MDC STRUDL is described in more detail in Reference 1. 3.9.1.2.1.2 C-E MARC The C-E MARC program is a general purpose nonlinear finite element program with structural and heat transfer capabilities. It is described in detail in Reference 2. C-E MARC is used for stress analysis of regions of vessels, piping or supports which may deform plastically under prescribed loadings. It is also used for clastic analyses of complex geometries where the graphics capability enables a well defined solution. The thermal capabilities of C-E MARC are used for complex geometries where simplification of input and Eraphical output are preferred. O Artwoved Dwyn Matedel Design of SSC Page 194 m
System 80+ Desfvt ControlDocumewst C-E MARC is the C-E modified version of the MARC program, which is commercially available and . O has had sufficient use to justify its applicability and validity. Extensive verification of the C-E version
' has been performed to supplement the public documentation.
l t
-3.9.1.2.1.3 PICEP The PICEP program calculates the flow through a crack in a pipe. PICEP uses the simplified engineering
- approach for elastic-plastic fracture analysis for finding the crack opening displacement and area. Fluid calculation options include single and two-phase flow as well as allowance for friction. PICEP was i developed.by EPRI. ;
3.9.1.2.1.4 SUPERPIPE ; L .
- SUPERPIPE is a linear finite element program for the static and dynamic analysis of piping systems.
These systems'may include such components as bends, elbows, tees, reducers, socket or butt welds, flexible couplings, and fianges, with the appropriate flexibility factors and stress indices accounted for. Support types may include rigid, spring, constant-force, snubber, anchor, or user-specified, and may have
.. any desired orientation.
5- ' Analyses performed include thermal, weight, applied load, frequency and mode shape, response } spectrum, and time-history. Following the static and dynamic analysis phase, the program performs a complete ASME B&PV Code, Section III Class 1 stress check, combining analysis results in any manner [ specified by the user to create the appropriate loading cases applicable for each of the ASME code stress equations. The user also supplies the number of occurrences of each steady-state and transient load state, ;
. with which the program performs a complete fatigue damage calculation.
SUPERPIPE, which is commercially available software, was developed by ABB-Impell and is described in detail in Reference 20. SUPERPIPE was verified in accordance with ABB-Impell's Quality Assurance l Manual. 3.9.1.2.1.5 DFORCE l The computer code program DFORCE calculates the internal forces and moments at designated locations
- in a piecewise linear structural system, at each time step, due to the time history of relative displacements
- of the system mass points and boundary points. The program also selects the maximum value of each L . component of force or moment at each designated location, and the times at which they occur, over the entire duration of the specified dynamic event.
I I The program forms appropriate linear combinations of the relative displacements at each time step and performs a complete loads analysis of the deformed shape of the structure at each time step over the . entire duration of the specified dynamic event. The program is used to calculate the time dependent reactions in structural models subjected to dynamic excitation which are analyzed by the CEDAGS - program. ; To demonstrate the validity of the DFORCE program, results for test cases were obtained and shown to be substantially identical to those obtained for an equivalent analysis using MDC STRUDL i 1 Ammar oeaon uneenior- Deeon or ssc . rene 3.s-s i i
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System 80+ Design ControlDocument 3.9.1.2.1.6 SG LINK SG LINK determines steam generator and snubber stroke and building interface boundaries for the steam generator snubber lever system. The program verifies the kinematics of the snubber lever linkage systems based on input motions of the steam generator lug and detailed snubber lever system geometry. 3.9.1.2.1.7 CEDAGS The computer program CEDAGS (C-E Dynamic Analysis of Gapped Structure) performs a piecewise linear direct integration solution of the coupled equations of motion of a three dimensional structure which may have clearances or gaps between the structure and any of its supports or restraints (boundary gap 3 or between points within the structure (internal gaps). The contacted boundary points may be oriented in any selected direction and may respond rigidly, elastically, or plastically. The structure may be subjected to applied dynamic loads or boundary motions. The CEDAGS program is used to calculate the dynamic respouse of piecewise linear structural systems subjected to time varying load forcing functions resulting from postulated pipe break conditions. To demonstrate the applicability and validity of the CEDAGS program, the solutions to an extensive series of test problems were obtained and shown to be substantially identical to results obtained by hand calculations or alternate computer solutions. 3.9.1.2.1.8 CE177, Head Penetration Reinforcement Program This program calculates reinforcement available and reinforcement required for penetrations in hemispherical heads. The technique described in paragraph NB-3332 of the ASME Code, Section III is used. This program is used to perform preliminary sizing and reinforcement calculations for hemispherical heads in the reactor vessel. The program was verified by comparisons of program results and hand calculated solutions of classical problems. 3.9.1.2.1.9 CE102, Flange Fatigue Program This program computes the redundant reactions, forces, moments, stresses. and fatigue usage factors in a reactor vessel head, head flange, closure studs, vessel flange, and upper vessel wall for pressure and thermal loadings. Classical shell equations are used in the interaction analysis. This program is used to perform the fatigue analysis of the reactor vessel closure head and vessel flange assembly. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.10 CE105, Nozzle Fatigue Program This program computes the redundant reactions forces, moments, and fatigue usage factors for nozzles in cylindrical shells. This program is used to perform the fatigue analysis of reactor vessel nozzles and steam generator feedwater nozzle. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. O Approved Design hintene! Design of SSC Pope 3.94
-'Sv' tem 80+ Desion ControlDocument .
, 3.9.1.2.1.11- CEC 26, Edge Coef'icients Program j
- ' This code calculates the coefficients for edge deformations of conical cylinders and tapered cylinders ;
. when subjected to axisymmetric unit shears and moments applied at the edges. {
This program is used to perform the fatigue analysis of reactor vessel wall transition. The program was i F verified by comparisons of program results and hand-calculated solutions of classical problems. :
-i
- 3.9.1.2.1.12 CE124, Generalized 4 x 4 Program ;
i Tnis program computes the redundant reactions, fcrces, moments, stresses, and fatigue usage factors for l the reactor vessel wall at the transition from a thick to thinner section and at the bottom head juncture. This program is used to perform fatigue analysis of reactor vessel bottom head juncture. The program ! ] - was verified by comparisons of program results and hand-calculated solutions of classical problems. t i i 4 3.9.1.2.1.13- ANSYS : ANSYS is a large-scale, general-purpose, finite element program for linear and nonlinear structural and :
- thermal analysis. This program is commercially available. Additional descriptive information on this
! code is provided in Section 3.9.1.2.2.2. This program is used for numerous applications for all components in the areas of structural, fatigue, thermal and eigenvalue analysis. The program was verified
~
by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.14 - CE301, The Structural Analysis for Partial Penetration Nozzles, Hester Tube Plug Welds, and the Water Level Boundary of the Pressmiser Shell Program i g This program computes various analytical parameters, primary plus secondary stresses and stress - intensities, peak stresses and stress intensities, and the cyclic fatigue analysis with usage factors at cuts i of interest. This program is utilized to satisfy the requirements of Section III, of the ASME B&PV Code. This program is used in the fatigue analysis of partial penetration nozzles in the pressurizer and piping.
- . The program was verified by comparisons of program results and hand-calculated solutions of classical )
problems. 3.9.1.2.1.15 CE223, hf tructure S Interaction Program This code calculates redundant loads, stresses, and fatigue usage factors in the primary head, tubesheet, secondary shell, and stay cylinder for pressure and thermal loadings. 1 This program is used in the fatigue analysis of the steam generator primary structure. The program was l verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.16 - CE362; Tube-To-Tubesheet Weld Program This code performs a three body interaction analysis of the tube-to-tubesheet weld juncture. The code calculates primary, secondary, and peak stresses and computes range of stress and fatigue usage factors. I
'(3 ' This program is used in the fatigue analysis of steam generator tube-to-tubesheet weld. The program was 1 V verified by comparisons of program results and hand-calculated solutions of classical problems.
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System 80+ Design ControlDocument 1 3.9.1.2.1.17 CE286, Support Skirt Loading Program i This code calculates the stresses in the conical support skirt of the steam generator for external loads. I This program is used in the structural analysis of steam generator support skirt. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.18 CE210, Principal Stress Program This code sums stresses for three load conditions and computes principal stress intensity, stress intensity range, and fatigue usage factor. This program is used in the fatigue analysis of steam generator components. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3,.9.1.2.1.19 CE211, Nozzle Load Resolution Program i This is a special purpose code, used to calculate stresses in nozzles produced by p ping loads in combination with internal pressure. This program is used in the fatigue analysis of steam generator nozzles. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.20 KIN 12100 Program This is a general purpose finite difference heat transfer program. This program is used for steady-state and transient thermal analysis. This program is used in numerous thermal relaxation analyses for all components. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.21 CEFLASII-4A This is a code used to calculate transient conditions resulting from a flow line rupture in a water / steam flow system. The prograrn is used to calculate steam generator internal loadings following a postulated main steam line break. This program is used in a steam line break accident structural analysis. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. 3.9.1.2.1.22 CRIBE This is a one-dimensional, two-phase thermal hydraulic code, utilizin; s momentum integral model of the secondary flow. This code was used to establish the recirculation rae and fluid mass inventories as a function of power level. The code is commercially available and has had sufficient use to justify its applicability and validity. This program is used for determining steam generator performance. The program was verified by comparisons of program results and hand-calculated solutions of classical problems. O Approved Desten Meterial- Design of SSC Page 3.9-8
System 80+ Design ControlDocument 3.9.1.2.1.23 FAST 2 I i V FAST 2 is a computer code originated by Shelltech Associates for the analysis of vessel-nozzle intersections. It uses closed form asymptotic results for the solutions of the thin shell equations. FAST 2 calculates stress and deflections of a cylindrical vessel or spherical head with a cylindrical pipe intersecting the vessel wall. Vessel geometries are idealized as a horizontal cylinder on two saddle supports, a horizontal cantilevered cylinder fixed on the left end, or a horizontal cantilevered spherical head. Spherical heads are simply supported at their base such that all points at the base remain in a vertical plane. Radial expansion and local rotation is a function of the head stiffness and the stiffness of an attached cylindrical vessel which may be included in the model at the user's option. The loading conditions available in FAST 2 are nozzle loads, vessel end loads, internal pressure, and thermal loads. Nozzle loads are applied at the nozzle / vessel intersection. Shear loads are not considered. Vessel end loads are external landings applied at the right end of the vessel. Internal pressure can be applied to any combination of vessel and pipe. Thermal loads are uniform thermal expansion parameters for each portion of the defined model. The code has the capability of modeling stiffening rings at either or both ends of a cylindrical vessel, and at the vessel / head junction for a spherical head. End caps or vessel heads on a cylindrical vessel may be modeled by stiffening rings representing the equivalent stiffness of the head or cap. FAST 2 has been used by Shelltech Associates in the development of WRC Bulletin No. 297. (n) v 3.9.1.2.1.24 PC-PREPS The evaluation and design of pipe support frames and baseplate is performed using PC-PREPS. PC-PREPS is a personal computer based, integrated pipe support analysis software package. It is interactive, menu-driven, with built-in structural analysis and graphics capability. The package is totally self-contained, except for a word processor used for the final calculation document production. All operations, including the finite element analysis, are performed on the personal computer. PC-PREPS allows a pipe support analyst to prepare data, view associated graphics, and execute frame and baseplate analyses. It can automatically perform load combinations and convert loads computed with pipe stress software to the pipe support frame, and from the frame to any of the defined baseplate. The post-processing capabilities of PREPS include AISC and NF Code checks, maximum displacement checks, weld stress check, and local stress check. PC-PREPS has been qualified by comparison to other software performing similar calculations and to manual calculations. 3.9.1.2.1.25 LIDOP The LIDOP program computes the local crush characteristics of a pipe section for use in the analysis of pipe motion and subsequent impact on structural targets or pipe rupture restraint structures. The program will generate crush rigidities and deformation energies for r,ssurized or unpressurized piping in the following geometries: r) V Approwd Design Material Desn]n of SSC Page 3.9 9
System 80+ Design ControlDocument
- Ring crush against flat rigid surface.
- Indent or straight pipe against rigid cylinder.
- 1.5D pipe elbow (extrados) against a flat rigid surface.
- Pipe bend (extrados) against a flat rigid surface.
- Indent of straight pipe against a rectangular block.
Both dynamic effects and material properties are considered in generation of the crush characteristics. Unpressurized force-displacement and energy-displacement characteristics of pipe and elbows are generated from empirical equations which are based on experimental data. Pressurization effects, based on fluid displacement during deformation, are superimposed on the unpressurized characteristics. The overall dimensions of the contact area, where applicable, are generated by empirically corrected geometric relationships. Dynamic effects of elbows are empirically determined from an experimental comparison of static and dynamic impact of spheres. Dynamic effects of all other geometries and elbows in certain cases are based on the results of finite element computer simulations of rings impacting flat, rigid surfaces. The effects of material properties are determined from empirical relationships based on computer predictions. 3.9.1.2.1.26 TIMIIIS6 The TIMillS6 program performs modzl superposition time history analysis for lumped mass / stick models and response spectra calculations. 3.9.1.2.1.27 RELAPS O RELAPS is used to perform transient analysis of thermal-hydraulic systems with water as the fluid. RELAPS uses a five equation two-phase flow continuity equations, two phasic momentum equations and an overall energy equation augmented by the requirement that one of the phases is assumed saturated. In this model, only two interphase constitutive relations are required, those for interphase drag and interphase mass exchange. Models are included for abrupt area changes, choking, mass transfer interphase drag, wall friction and branching. The program requires numerical input data that completely describes the initial fluid conditions and geometry of the system being analyzed. The output consists of variables necessary to describe the transient state of the system being analyzed. 3.9.1.2.1.28 REPIPE REPIPE computes the loading time histories on a piping network based upon the results from computer program RELAPS hydrodynamic analysis of the contained fluid. The RELAP5 time-varying pressure, momentum flux and energy states throughout a fluid system containing water, steam, and/or a two phase mixture are used as in input to the REPIPE program to produce time histories for input to the piping stress analysis program. REPIPE distributes the RELAPS control vohune forces to the structural network nodes by a process based upon fluid momentum balance principle and newtons third law of motion. The output from REPIPE consists of dynamic loads on the pipe, organized into force vs time tables. Approved Design Materiet Desogn of SSC Page 39-10
t Sy tem 80+ Deslan conuot Document ! i 3.9.1.2.1.29 CCN-318 ,
\ CCN-318 is a computer program used to evaluate the design of rectangular cross section attachments on ,
- ASME Class 2 and 3 Piping following the requirements of ASME Code Case N-318. The program ,
checks for Code Case limiutions, calculates the required coefficients and then checks local stress in the :
; pipe wall. In addition, it also evaluates the adequacy of fillet and partial penetration welds. The results of the analysis are compared to ASME Code Allowables.
i t 3.9.1.2.1.30 CCN-392 CCN-392 is.a computer program used to evaluate the design of circular cross section attachments on f ASME Class 2 and 3 piping following the requirements of ASME Code Case N-392. - The program j checks for code case limitations, calculates the required coefficients and then checks the local stress in ;
' the pipe wall. In addition, it also evaluates the adequacy of fillet and partial penetration welds. The- ,
. results of the analysis are compared to ASME Code Allowables.
3.9.1.2.1.31: TRANS2A TRANS2A is a computer program which determines radial temperature distributions and gradients in a l
. pipe wall experiencing fluid temperature excursions. TRANS2A determines these temperature i distributions by solution of the unsteady one-dimensional axisymmetric heat transfer equation. For aid l in Class _1 piping analysis values of the thermal gradients AT: and AT2 and the average temperatures T, ;
and/cr T3are calculated (and printed)in accordance with ASME BPVC Section III Article NB-3650. To be of more aid to the analyst in choosing values of the average and temperature gradient data to be input to the combined stress analysis, TRANS2A evaluates the actual histories of the thermal stress terms O according to the equations of Section III, Article NB-3650 with as many as ten sets of stress indices and summarizes them in a table by extreme and time of occurrence. l i 3.9.1.2.2 Code Class CS Internals, Fuel and CEDMs The following computer programs are used in the static and dynamic analyses of reactor internals, fuel, and CEDMs. 3.9.1.2.2.1 MRI/STARDYNE The MRI/STARDYNE program uses the finite element method for the static and dynamic analysis of two-and three-dimensional solid structures subjected to any arbitrary static or dynamic loading or base
- acceleration. In addition, in.itiai displacements and velocities may be considered. The physical structure to be analyzed is modcied with finite elements that are interconnected by nodes. Each element is t constrained to deform in accordance with an assumed displacement field that is required to satisfy i I
continuity across element interfaces. The displacement shapes are evaluated at nodal points. The equations relating the nodal point displacements and their associated forces are the element stiffness relations and are a function of the element geometry and its mechanical properties. The stiffness relations for an element are developed on the basis of the theorem of minimum potential energy. Masses and - external forces are assigned to the nodes. The general solution procedure of the program is to formulate the total following equations: O homed Demipn neenerw Desten or ssc rege 3.s.11 i 4 4
- ' r w r-' -
System 80+ Design ControlDocument (1) [K] * {6} = (P} w2 [m]{q} -[K] {q} = 0 (2) where: {5} = the nodal displacement vector {P} = the applied nodal forces [m] = the mass matrix w = the natural frequencies {q} = the normal modes Equation (1) applies during a static analysis which yields the nedal displacements and finite elements internal forces. Equation (2) applies during an eigenvalue/eigenvedor analysis, which yields the natural frequencies and normal modes of the structural system. Using the natural frequencies and normal modes together with related mass and stiffness characteristics of the stmesure, appropriate equations of motion may be evaluated to determine structural response to a predescribed dynamic load. The finite elements used to date in C-E analyses are the elastic beam, plate and ground support spring members. The assumptions governing their use are as follows: small deformation, linear-clastic behavior, planc sections remain plane, no coupling of axial, torque and bending, geometric and clastic properties constant along length of element. Further description is provided in Reference 4. The MRI/STARDYNE code is used in the analysis of reactor internals. The ANSYS code (3.9.1.2.2.2) and the COSMOS code (3.9.1.2.2.9) are also used as alternatives to STARDYNE. The program is used to obtain the mode shapes, frequencies and response of the internals to predescribed static and dynamic loading. The structural components are modeled with beam and plate elements. Ground support spring elements are used, at times, to represent the effects of surrounding stmetures. The geometric and elastic properties of these elements are calculated such that they are dynamically equivalent to the original structures. The response analysis is then conducted using both modal response spectra and modal time history techniques. Both methods are compatible with the program. The program is also used to perform a static finite element analysis of the lower support structure to determine its structural stiffness. MRl/STARDYNE is commercially available software and has had sufficient use to justify its applicability and validity. Extensive verification of the C-E version has been performed to supplement the public documentation. 3.9.1.2.2.2 ANSYS ANSYS is a general purpose nonlinear finite element program with structural and heat transfer capabilities. It is described in Reference 5. ANSYS is used to perform detailed stress analyses of the 1
- fuel assembly due to combined lateral and vertical dynamic loads resulting from postulated seismic and Ap\ proved Desegn Meteriel Design of SSC Pope 3.9-12 w
Sy? tem 80+ Design C*ntrolDocument loss-of-coolant-accident conditions. Static finite element analyses of reactor internal structures, such as O flanges, expansion compensating ring and core shroud, are performed with ANSYS to determine vertical C and lateral stiffnesses and thermal stresses. ANSYS is a proprietary code and commercially available. The developers, Swanson Analysis Systems, l Incorporated have published an ANSYS verification manual with numerous examples of its usage. l l 3.9.1.2.2.3 ASIISD i The ASilSD program uses a finite element technique for the dynamic analysis of complex axisymmetric structures subjected to any arbitrary static or dynamic loading or base acceleration. The three-dimensional axisymmetric continuum is represented as an axisymmetric thin shell. The axisymmetric shell is discretized as a series of frustums of cones. Nmilton's variational principle is used to derive the equations of motion for these discrete stmetures. i This leads to a mass matrix, stiffness matrix, and load vectors which are all consistent with the assumed displacement field. To minimize computer storage and execution time, the nondiagonal " consistent" mass matrix is diagonalized by adding off-diagonal terms to the appropriate diagonal terms. These equations of motion are solved numerically in the time by a direct step-by-step integration procedure. The assumptions governing the axisymmetric thin shell finite element representation of the structure are those consistent with linear orthotropic thin elastic shell theory. Further description is provided in Reference 6. ASilSD is used to obtain the dynamic response of the core support barrel under normal operating l fs conditions and due to a LOCA. An axisymmetric thin shell model of the structure is developed. The (} ' spatial Fourier series components of the time varying normal operating hydraulic pressure or LOCA loads 1 are applied to the modeled structure. The program yields the dynamic shell and beam mode response of the structural system. l ASHSD has been verified by demonstration that its solutions are substantially identical to those obtained I by hand calculations or from accepted experimental tests or analytical results. The details of these 1 comparisons may be found in References 6 and 7. l 3.9.1.2.2.4 CESIIOCK The computer program CESHOCK solves for the response of structures which can be represented by l lumped-mass and spring syctems and are subjected to a variety of arbitrary type loadings. This is done ; by numerically solving the differential equations of motion of an n* degree of freedom system using the l Runge-Kutta-Gill technique. The equations of motion can represent an axially responding system or a laterally responding system (i.e., an axial motion, or a coupled lateral and rotational motion). The program is designed to handle a large number of options for describing load environments and includes l I such transient conditions as time <!ependent forces and moments, initial displacements and rotations, and initial velocities. Options are also available for describing steady-state loads, preloads, accelerations, gaps, nonlinear elements, hydrodynamic mass, friction, and hysteresis. 1 The output from the code consists of minimum and maximum values of translational and angular 1 accelerations, forces, shears, and moments for the problem time range. In addition, the above quantities l A are presented for all printout times requested. Plots can also be obtained for displacements, velocities V and accelerations as desired. Fur her description is provided in Reference 8. Approved Desogn Atatenal Design of SSC Pope 3.9-13
System 80+ Design ControlDocument The CESHOCK program is used to obtain the transient response of the reactor vessel intemals and fuel assemblies due to LOCA and seismic loads. Lateral and vertical lumped-mass and spring models of the internals are formulated. Various types of springs (linear, compression only, tension only, or nonlinear springs) are used to represent the structural components. Thus, judicious use of load-deflection characteristics enables effects of components impacting to be predicted. Transient loading appropriate to the horizontal and vertical directions is applied at mass points and a dynamic response (displacements and internals forces) is obtained. CESHOCK has been verified by demonstration that its solutions are substantially identical to those obtained by hand calculations or from accepted analytical results via an independent computer code. The details of these comparisons may be found in References 7 and 8. 3.9.1.2.2.5 MODSK MODSK is a C-E computer program which solves for the natural frequencies and mode shapes of a structural system. The natural frequencies and mode shapes are extracted from the system of equations: (K-Wn M) $n = 0 where: K = model stiffness ma,trix M = model mass matrix W, = natural circular frequency for the n* mode 6 = nonnal mode shape matrix for the n* mode The solution to the general eigenvalue problem is obtained using the dual Jacobi rotation method. The MODSK code is used in the analyses of reactor internals to obtain frequencies and mode shapes, and damping parameters. The results of these analyses are incorporated into overall reactor vessel internals models, which calculates dynamic response due to seismic and LOCA conditions. The MODSK program was Ceveloped by C-E. To demonstrate the validity of the MODSK program, results from lateral and veincal test problems were obtained and shown to be substantially identical to those obtained from en equivalent analysis using the cornmercially available ANSYS program (Refer to Section 3.9.1.2.2.2). 3.9.1.2.2.6 SAPIV The SAPIV computer code is a structural analysis program capable of analyzing two and three-dimensional linear complex structures subjected to any arbitrary static and dynamic loading or base acceleration. The analysis technique is based on the finite element displacement method. The structure to be analyzed can be represented using bars, beams, plates, membranes and three-dimensional finite elements. O kywoved Design histerial Design of SSC Page 3.914
Systern 80+ Design controlDocument l Structural stiffness and load vectors are assembled from the element matrices which are derived assuming (\ various displacement functions within each element whereas lumped mass matrices are used to represent inertia characteristics of the structure. In the static analysis, the assembled equations of equilibrium are solved by using a linear equation solver. Dynamic analysis capabilities include modal analysis, modal ; superposition and direct integration methods of computing dynamic response and response spectrum techniques. SAPIV has been applied to the eigenvalue and response spectra analyses of spent fuel storage racks and lifting rig structures. The SAPIV code is used in the computation of dyna.nic response of control element drive mechanisms under mechanical and seismic loads. Both modal analysis and response spectrum capabilities of the code are used to find the natural frequencies and mode shapes and the dynamic loads in CEDM components. ANSYS (3.9.1.2.2.2) is also used as an alternative to SAPlV. SAPIV is commercially available software and has had sufficient use to justify its applicability and validity. Extensive verification of the C-E version has been performed to supplement the public documentation. 3.9.1.2.2.7 CEFLASH-4B The CEFLASH-4B computer code (Reference 14) predicts the reactor pressure vessel pressure and flow distribution during the subcooled and saturated portion of the blowdown period of a Loss-of-Coolant-Accident (LOCA). The equations for conservation of mass, energy and momentum along with a representation of the equation of state are solved simultaneously in a node and flow path network representation of the primary reactor coolant system. (O, CEFLASH-4B provides transient pressures, flow rates and densities throughout the primary system following a postulated pipe break in the reactor coolant system. The CEFLASH-4B computer code is a modified version of the CEFLASH-4A code (References 15 through 17). The CEFLASH-4A computer code has been approved by the NRC (References 18 and 19). The capability of CEFLASH-4B to predict experimental blowdown data is presented in Reference 14. 3.9.1.2.2.8 LOAD LOAD calculates the applied forces of the axial internals model which is contained within water control volumes using results from the CEFLASH-4B blowdown loads analysis as input. The fluid momentum equation is applied to each volume and a resultant force is calculated. Each force is then apportioned to the various stmetural nodes contained within the volume. Use of the fluid momentum equation takes into account pressure forces, fluid friction, water weight, and momentum changes within each volume. The . I resultant forces are combined with the reactor vessel motions obtained from the reactor coolant system analysis before the structural responses are determined. The LOAD code has been verified by demonstrating that its solutions are substantially identical to those obtained from hand calculations. 3.9.1.2.2.9 COSMOS COSMOS is a general purpose program, commercially available, which can be used to perform finite e!cment, dynamic, eigenvalue, response spectra, random vibration, and other structural evaluations. The program has advanced features which include complex geometry mesh generation, color graphics and p N i A;4wond Desy Atatenal. Des # of SSC Page 3.9-15
I l System 80+ Design ControlDocument l 1 1 l postprocessing. COSMOS can solve problems using models with up to 5000 nodes and 8000 degrees of freedom. It has a library of 60 element types which can be represented with linear, isotropic, , orthotropic, non linear and composite material. Some of the non-linear capabilities include non-linear buckling, large strains, stress stiffening, kinematic hardening, gaps and friction. Some of the postprocessing features include node and element plots, color stress contour plots, various view capabilities, zoom features, shade and animation. The COSMOS program can also communicate with ANSYS (Section 3.9.1.2.2.2). 3.9.1.3 Experimental Stress Analyses \\When experimental stress analysis is used, it is performed in accordance with Appendix 11 of ASME Boiler and Pressure Vessel Code, Section ill, Division 1.}}1 3.9.1.4 Considerations for the Evaluation of the Faulted Condition 3.9.1.4.1 Seismic Category I RCS Items The major components of the reactor coolant system (RCS) are designed to withstand the forces associated with the design basis pipe breaks discussed in Section 3.6, in combination with the forces associated with the Safe Shutdown Earthquake and normal operating conditions. For structural evaluation, the design basis pipe breaks are those breaks for which leak-before-break cannot be demonstrated. Since the dynamic effects of breaks in piping systems listed in Section 3.6.2.2.1 are eliminated by leak-before-break, the pipe break loads analysis procedure considers only those branch line . pipe breaks not eliminated by leak-before-break. See Section 3.9.3 for discussion of loading combinations. Analyses are performed to generate component loads and motions due to the forces associated with branch line pipe breaks. The analyses account for the reactor vessel and supports, major connected piping and components and the reactor internals. The results of the analyses include loads on major component supports and RCS piping loads. The analyses performed for branch line breaks use the MDC STRUDL (Section 3.9.1.2.1.1) or ANSYS (Section 3.9.1.2.1.13) code. The resultant component and support reactions are specified, in combination with the appropriate normal operating and seismic reactions, for design verification by the methods discussed below and in Section 3.9.3. The system or subsystem analysis used to establish, or confirm, loads which are specified for the design of components and supports is performed on an elastic basis. 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. $(sproved Destyn Maternal
- Des @n of SSC Page 3.9-16
System 80+ Deslan Coneof Deewnent , p). ("' Whc:: an clastic system analysis is employed to establish the loads which act on components and supports, elastic stress analysis rnethods are also used in the design calculations to evaluate the effects of the loads 7 on the components and supports. ~ In particular, inelastic methods such as plastic instability and limit-analysis methods, as defined in Section III of the ASME Code, are not used in conjunction with an elastic , system analysis. The RCS and its supports, which are analyzed using clastic methods, are shown in diagram form in Figure 3.9.I. Inelastic methods of analysis are used in cases where it is deemed desirable and appropriate to permit - significant local inelastic response. In these cases, if any, the system or subsystem analysis performed , to establish the loads' which act on components and component supports are modified to include the ! , inelastic strain compatibility in the local regions of the components and component supports at which significant locsl inelastic response is permitted. Inelastic methods defined in Section III of the ASME Code as plastic instability or limit analysis mathadi ! are not used. l Reactor Internals and CEDMs [ - 3.9.1.4.1.1 t {
. See Sections 3.7.3.14 and 3.9.2.5.
' 3.9.1.4.1.2 Non-Code Itsens '
The components not covered by the ASME Code but which are related to plant safety include: l cp
- Reactor Internal Stmetures (Class IS).
4 e- Fuel. i- :
- * . Control element drive mechanisms (CEDMs).
l 'e Control element assemblies (CEAs). i [ Each of these' components is designed in accordance with specific criteria to ensure their operability as , !- it relates to safety. The fuel assembly and control element assembly design is discussed in Section 4.2. The non-code components of the control element drive mechanisms (CEDMs) are proven by testing as ! described in Section 3.9.4.4. : 3.9.1.4.2 blemic Category 1 Non-NSSS Items The analytical method for evaluating the' faulted condition uses a linear elastic model as described in l Section 3.7.3. The ASME Section III allowable stress limits will be met for faulted loads, including the ; safe shutdown earthquake and system transient loads described in Section 3.9.1. Pipe rupture restraint :
. energy absorbing members are an exception to the use of linear elastic inodels. The methods for the !
dynamic analysis of pipe whip are given in Section 3.6.2.2.2.2. ' For allowable stresses and design ! criteria. see Sections 3.6.2.3.2.4 and 3.6.2.3.2.5, respectively.
.1 i
O ! w a neeon neneerm. Dee6n er ssc rege 2.s.ry e s.m n . . . . < - a, --- , -.m - -v.,- .~- -
System 80+ Design ControlDocument 3.9.2 Dynamic System Analysis and Testing 3.9.2.1 Piping Vibrations, Thermal Expansion, and Dynamic "ffects 94 Safety-related piping systems are designed in accordance with the ASME B&PV Code, Section III. The j i preoperational test program will be conducted in accordance with ASME OM-S/G Part 3 Standard and Part 7 Guide and is applicable to the following systems:
- All ASME Code Class 1,2, and 3 systems. l
- Other high-energy piping systems inside Seismic Category 1 Structures.
- High-energy portions of systems whose failure could reduce the functioning of any Seismic Category 1 plant feature to an unacceptable safety level.
- Seismic Category 1 portions of moderate-energy piping systems located outside containment.
The supports and restraints necessary for operation during the life of the plant are considered to be parts of the piping system. The preoperational tests confirm that these piping systems, restraints, components, and supports have been adequately designed to withstand flow-induced dynamic loadings under the steady-state and operational transient conditions anticipated during service, to confirm that normal thermal motion is not restrained, and to demonstrate that piping vibrations are within acceptable levels. 3.9.2.1.1 Steady-State Vibration Essential systems and systems with the potential to experience significant vibration are monitored for steady-state vibration. The piping is monitored during normal operating and test modes along with operating modes expected to result in the most severe vibration. The piping is visually inspected and vibration movements will be measured using portable instrumentation at locations where the vibration is judged to be the most severe. When necessary, the piping will be instrumented and monitored remotely. The measured piping displacements are compared with allowable displacement limits that are based on the allowable stress amplitudes, Sci. So,is calculated in accordance with ANSI /ASME OM-S/G Part 3 Sandard. S .. is lic.iH to Sa la where. 6 Sg = 0.8 times the alternating stress (Sa) at 10 cycles from Figure 1-9.1 or at 10" cycles from Fig.1-9.2.2 of the ASME section 111 Code. a = allowable stress reduction factor: 1.3 for materials covered by Fig.1-9.1; or 1.0 for material covered by Fig.1-9.2.1 or 1-9.2.2 of the ASME Section 111 Code. O Apvweved Desogn htstanal . Desogn of SSC (11/96) Page 3.9-18
i i 1 I System 80+ Design CanarolDoewnent l The stress reduction factor ensures that the allowable stress amplitude s, is based on an allowable Section ; lil Code alternating stress latensity at a number of cycles consistent with a 60 year design life. Adequate design margin is ensured b/ employing 80% of the minimum allowable Section III Code alternating stress. .; If the measured piping displacements exceed allowable limits, one or more of the following actions are ,
- taken so that the vibration can be qualified. ,
i
* ' Analyses are performed to show that the measured displacements are acceptable. l *- Additional testing is performed to show that the peak stresses due to the vibration are acceptable. j
- The source of the excessive vibrations is eliminated. i t
t
- The pipe supporting arrangement is modified to reduce the vibration to acceptable levels. j
- 3.9.2.1.2 Tranelant Vibration Vibration monitoring is performed for systems expected to experience significant transients. The piping is instrumented to measure the system response during the transient events. The measured response is ;
cornpared with analytically predicted values from the piping stress report. If the predicted values are l exceeded, the measured response is shown to be acceptable by additional analyses or testing; or the j source of the transient is eliminated or modified to reduce the transient loadings or modifications to the pipe supporting arrangement will be made to reduce the system response to acceptable levels. 3.9.2.1.3 'Ibennal Expansion Safety-related systems that are expected to experience significant thermal movements are monitored for j thermal expansion. A preheatup walkdown is performed so that locations of potential thermal , interferences can be identified and appropriate corrective action taken prior to heatup. One complete ; thermal cycle, i.e., cold position to hot position to cold position, is monitored. The piping and ; components are visually inspected and piping displacements are monitored at predetermined locations. l The measurement locations are based on the locations of snubbers, hangers, and expected large , displacements. When necessary, the piping is instrumented and monitored remotely. i Acceptable limits of pipe disph ement, based on analytically predicted movements from the piping stress reports, are determined prior to testing. The measured displacements are compared to the acceptance limits to determine whether the piping systems are free to expand as expected. If the measured displacements are not within the acceptance limits, then analyses are performed or corrective action is ! taken, as appropriate, to ensute that neither pipe stress nor support and equipment allowables are exceeded. 3.9.2.2 blemic Qualification of Mach =aic=t Equipenent This section describes the seismic qualification methods and procedures for NSSS and non-NSSS mechanical equipment which is safety related (Seismic Category I) or non-safety related mechanical ; equipment whose failure can prevent the satisfactory accomplishment of safety function (Seismic Category II). The procedures described herein are consistent with U.S. Nuclear Regulatory Guide 1.100, Rev. 02. 1 Mechanical equipment which is designed to ASME Code requirements satisfy the procedures of RG ; 1.100. ; Mweued Deem asenner. ceew a ssc 11 ansi rene 2.s ts
System 80+ Design ControlDocument l 1 3.9.2.2.1 Nuclear Steam Supply System ! The operability of all active and passive safety-related mechanical equipment (Seismic Category I) related O to the NSSS is demonstrated by analysis and/or testing. The structural integrity of non-safety related mechanical equipment (Seismic Category II) is demonstrated by analysis and/or testing. The methods and procedures used and the results of tests and analyses that confirm implementation of the design ; criteria for safety-related mechanical equipment, including supports, are provided in Section 3.9.3, 3.9.4 and 3.9.5. 3.9.2.2.2 Non-NSSS Items The following dynamic analysis and testing procedures are used for Seismic Category I and II mechanical equipment and equipment supports. 3.9.2.2.2.1 Seismic Testing and Analysis The ability of Seismic Category I equipment to perform its functions during and after an earthquake is demonstrated by tests and/or analysis. Loadings due to plant normal operation, seismic and non-seismic vibrations shall be considered. Prior to SSE qualification, it is demonstrated that the equipment can withstand the 1/2 SSE excitation without loss of function. Damping values for equipment being qualified a, ken from Regulatory Guide 1.61 and IEEE Std. 344-1987 or from other documented test data. The ' selection of testing and/or analysis for a particular piece of equipment is based on practical considerations. When practical, the Seismic Category I operations are activated and tested during the vibratory testing. When this is not practiwl, these operations are simulated by a combination of tests and analysis. ; The structural integrity of Seismic Category II mechanical equipment is demonstrated during and after an earthquake by analysis and/or tests. Loadings due to plant normal operation, seismic and non-seismic vibrations shall be considered. Prior to SSE qualification, it is demonstrated that the equipment can I withstand the 1/2 SSE excitation without loss of structural integrity. l l Dynamic coupling between the equipment and related systems is considered. Damping values for equipment being qualified are taken from Regulatory Guide 1.61 and IEEE Std. 344-1987 or from other documented test data. 3.9.2.2.2.2 Seismic Analysis Seismic Category I equipment that is large, simple (e.g., panels, pumps and valves), and/or consumes large amounts of power and whose functional operability is assured by its structural integrity is qualified by an analysis to show that the loads, stresses, and deflections are less than the values which give assurance of proper operation. Analysis is also used to show that there are no natural frequencies of tested equipment below the frequency range capability of the test facility. i 3.9.2.2.2.3 Basis for Test Input Motion When equipment is qualified by test, the response spectrum or the time history at the point of attaciunent to the supporting structure is the basis for determining the test input motion. O1 AppwondDesign Motenet . Design of SSC Page 3.9-20 l l l l
Srtem 80+ Design Control Document 3.9.2.2_?.4 Random Vibration Input [ l
'%,/ When random vibration input is used, the actual input motion envelopes the appropriate flow input motion at the individual modes. However, single frequency input, such as sine beats, is used provided one of the following conditions are met:
e The characteristics of the required input motion are dominated by one frequency.
- The anticipated response of the equipment is adequately represented by one mode.
e The input has sufficient intensity and duration to excite all modes to the required magnitude, such that the testing response spectra will envelope J e corresponding response spectra of the individual modes. 3.9.2.2.2.5 Input Motion The input motion is applied to the vertical and one horizontal axis simultaneously. However, if the equipment response along the vertical direction is not sensitive to the vibratory motion along the horizontal direction, and vice versa, then the input motion is applied to one direction at a time. In case of single frequency input, the time phasing of the inputs in the vertical and horizontal directions is such that a purely rectilinear resultant input is avoided. 3.9.2.2.2.6 Fixture Design g The fixture design simulates the actual service mounting and causes no extraneous dynamic coupling to j V) the equipment. 3.9.2.2.2.7 Equipment Testing Equipment testing is on the prototype basis. Similarity between the equipment being tested and the installed equipment is assured. This is usually done by the vendor who supplies the equipment. 3.9.2.3 Dynamic System Analysis Methods for Reactor Vessel Core Support and Internal Structures 3.9.2.3.1 Introduction The flow-induced vibration of the reactor internals components during normal operation can be characterized as a forced response to both deterministic (periodic and transient) and random pressure fluctuations in the coolant. Methods have been developed to predict the various components of the hydraulic forcing function and the response of the reactor internals to such excitation. This analytical methodology is summarized in Figure 3.9-2. The method separates the response calculations into two groups in accordance with the physical nature of the loading i.e., deterministic or random. Methods for developing the deterministic component of the hydraulic forcing function are discussed in Section 3.9.2.3.2, while those relating to the random component are discussed in Section 3.9.2.3.3, Where complex flow path configurations or wide variations in pressure distribution are involved, the hydraulic forcing functions are formulated using a test-analysis combination method utilizing a data obtained from plant tests and/or scaled model tests. Alyveved Destgrr Material . Design of SSC Page 3.9-21
System 80 + Design ControlDocument The response of the reactor vessel core support and internal structures (to include Core Support Barrel ' Assembly, Upper Guide Structure Assembly and Lower Support Structure Assembly) to the normal operating hydraulic loads are calculated by finite element techniques. The mathematical models used in these response analyses are described in Section 3.9.2.3.4. The methods used in calculating the structural responses are discussed in Section 3.9.2.3.5. 3.9.2.3 2 Periodic Forcing Function 3.9.,2.3.2.1 Core Support Barrel Assembly An analysis based on an idealized hydrodynamic model is employed to obtain the relationship between reactor coolant pump pulsations in the inlet ducts and the periodic pressure fluctuations on the core support barrel. A detailed description of this model and subsequent solution are given in References 21 through 27. The model represents the annulus of coolant between the core support barrel and the reactor vessel. In deriving the governing hydrodynamic differential equation for the above model, the fluid is taken to be compressible and inviscid. Linearized versions of the equations of motion and continuity are used. The excitation on the hydraulic model is harmonic with the frequencies of excitation corresponding to pump rotational speeds and blade passing frequencies. The result of the hydraulic analysis is a system of equations which define the forced response, natural frequencies and natural modes of the hydrodynamic model. The forced response equations define the spatial distributions of pressure on the core support barrel system as a function of time. 3.9.2.3.2.2 Upper Guide Structure The dynamic force on the upper guide structure assembly is due to flow induced forces on the tube bank. The periodic components of these forces are caused by pressure pulsations at harmonics of the pump rotor and blade passing frequencies, and vonex shedding due to crossflow over the tubes. A series of tests on full size tubes at reactor pressure and temperature indicated no evidence of periodic vortex shedding at the Reynolds Number and turbulence levels expected in the tube bank (Reference 28). Thus, the only significant periodic force is that due to pump pulsations. Data from this same test series was utilized to determine the magnitude of these pulsations at the pump rotor, twice the rotor, blade passing, and twice blade passing frequencies. 3.9.2.3.2.3 Lower Support Structure Assembly The ICI nozzles and the skewed beam supports for the ICI support plate are excited by periodic and/or random, flow induced forces. l The periodic component of this loading is due to pump related pressure fluctuations and vortex shedding due to crossflow. High turbulence intensity caused by jetting through the flow skirt makes it unlikely that regular vortex shedding will occur (References 29 and 30), if it were assumed to occur, the maximum shedding frequency would be well below the lowest structural frequency for both the ICI suppon nozzles and skewed beams. The magnitude and frequency of this periodic force are accounted i for based on data in the literature for crossflow over both vertical (References 31 and 32) and skewed (Reference 33) isolated tubes. I Derivation of pump frequency related loads is accomplished by assuming that these periodic pressure l variations are propagated undiminished through the flow skirt from the lower portion of the core barrel - l Asywowed Desmn Material Design of SSC Page 3.9-22 l
System 80+ Design Control Document m reactor vessel annulus. The magnitude of these pulsations is based on a combination of analytical ( ) predictions, based on Reference 21, and data from previous precritical programs (References 23 and 24). 3.9.2.3.3 Random Forcing Function 3.9.2.3.3.1 Core Support Barrel Assembly The random hydraulic forcing function is developed by analytical and experimental methods. An analytical expression is developed to define the turbulent pressure fluctuation for fully developed flow (Reference 34). This expression is modified, based upon the result of scale model testing (References 35 and 36), to account for the fact that flow in the downcomer is not fully developed. Based upon tests results, an expression is developed to define the spatial dependency of the turbulent pressure fluctuations. In addition, experimentally adjusted analytical expressions are developed to define the peak value of the pressure spectral density associated with the turbulence and the maximum area of coherence, in terms of the boundary layer displacement, across which the random pressure fluctuations are in phase (References 25,26 and 27). The transient behavior of the random fluctuations during loop startup and shutdown is assumed to be identical to that of the periodic excitations. 3.9.2.3.3.2 Upper Guide Structure Results of the full size tube tests (Reference 28) showed that at normal operating conditions the shroud tubes are excited by upstream and wake produced turbulent buffeting (References 28,37 and 38). The forcing function for this type of loading can be represented as a band limited white noise power spectrum (Reference 28). The magnitude of this spectrum is computed based on data from these tests. The p resultant velocity dependent force is combined with static drag loads to compute the amplitude response V and stress levels. 3.9.2.3.3.3 Lower Support Structure Assembly The ICI nozzles and ICI support plate support beams are both subject to turbulent buffeting by the flow skirt jets. The outermost ICI nozzles and beams receive full impact of the jets before the jets decay due to fluid entrainment and the presence of inner tube rows. The force spectrum of these jets is assumed to be represented as wide band white noise. The magnitude of this spectrum is based on data in the literature for impingement of turbulent jets (Reference 39 and 40). This velocity dependent magnitude is applied to each tube, assuming no change in jet characteristics, between the outermost and inner tubes. The approach velocity for each tube is calculated from an analytical expression based on experimental data on the velocity distribution in the lower portion of the reactor vessel-core barrel annulus and the flow skirt. 3.9.2.3.4 hiathematical hiodels A finite element analysis is performed on each of the reactor internals components using mathematical models. These models are designed to provide the most efficient analysis under the most significant loading condition to which each structure is exposed. The core support barre! assembly is modeled as a shell using the ASHSD computer code (Reference 6) (Figure 3.9-3). The structure is fixed at the upper flange to determine the beam modes and frequencies. The shell modes and frequencies are found by considering the upper flange fixed and the lower flange pinned. These analyses include hydrodynamic mass effects. All significant mode shapes and frequencies are used in combination to perform the normal I ) v Approved Design Atatenet. Design of SSC Page 3.9-23
System 80+ Design ControlDocumen2 I operating deterministic response analysis. A simplified finite element model of the barrel assembly is generated on the STARDYNE computer code (Reference 4), ANSYS (Reference 5), or COSMOS (Section 3.9.1.2.2.9) for use in the random response analysis. l The control element shroud tubes in the upper guide structure assembly are modeled as beams supported at the ends by plate elements. The end plates are in turn supported by spring elements which represent the stiffness of additional surrounding structure. A typical model of this configuration is shown in Figure 3.9-4. The STARDYNE computer code (Reference 4), ANSYS (Reference 5), or COSMOS is employed to allow the same models to be utilized for modal analysis as well as deterministic and random response analysis. The lower support structure assembly is modeled in several ways. Beam and plate elements are assembled !n a comparatively coarse mesh to model the entire Instrtrnent Nozzle Assembly (Figure 3.9-5). This representation of the stmeture is used in the STARDYNE computer code (Reference 4), ANSYS (Reference 5), or COSMOS to determine the modes, frequencies and response actions of the assembly as a system. The reaction points in this model are taken at the bottom plate level of the LSS Assembly. Typical ICI nozzles (Figure 3.9-6) and Skewed Beams (Figure 3.9-7) are modeled as fine mesh beam elements reacted at the support points by spring elements representing the surrounding structure flexibility. These component models are used on the STARDYNE computer code (Reference 4), ANSYS (Reference 5), or COSMOS to provide the individual structural modes, frequencies and responses within the system. The results of both individual and system analysis are combined to provide the total response. 3.9.2.3.5 Response Analysis 3.9.2.3.5.1 Determhilstic Response The normal mode method (Reference 41) is used to obtain the structural response of the reactor internals to the deterministic forcing functions developed in Section 3.9.2.3.2. The method is applied to the appropriate finite element models described in Section 3.9.2.3.4. Generalized masses based on mode shapes and the mass matrices from the finite element computer programs are calculated for each component's modes of vibration. Modal force participation factors are based on the mode shapes and the predicted periodic forcing functions are calculated for each mode and forcing function. The generalized coordinate response for each mode is then obtr.ined through solution of the corresponding set of independent second order single-degree of freedom equations. Utilizing displacement and stress mode shapes from the finite element computer programs, the modal responses of the reactor internals are obtained by means of the appropriate coordinate transformations. Response to any specific forcing function is obtained through summation of the component modes for that forcing function. 3.9.2.3.5.2 Random Response The normal mode method (Reference 41) is used to obtain the structural response of the reactor internals subjected to random forcing functions. The random forcing functions are assumed to be of both the band limited and wide band white noise varieties as described in Section 3.9.2.3.3. Experimental and analytical expressions are used to define the force power spectral density associated with flow related turbulence and jet impact. The appropriate mathematical models described in Section 3.9.2.3.4 are used in the STARDYNE computer code (Reference 4), ANSYS (Reference 5), or COSMOS. These codes O l AMweved Design Matenal . Desogn of SSC Page 3.9-24
System 80+ Design ControlDocument p compute the response RMS displacements, loads and stresses in a multi-degree-of-freedom linear elastic structural model subjected to stationary random dynamic ioadings, such as those described in Section (~') 3.9.2.3.3. A value of 3 x RMS is used for considering peak responses to random loading. These peak values are then combined with results from other analyses (e.g., deterministic, thermal, etc.) and utilized in design verification analyses. The use of the value 3 x RMS is common design practice based upon the assumptions of Random Gaussian loading of structures made of ductile materials, as discussed in References 46 and 47. The largest response of the Core Suppon Barrel is expected to be in the " beam" mode. The simplified finite element model of this structure, described in Section 3.9.2.3.4, is used to compute these displacements. The Upper Guide Structure and Lower Support Structure do not respond to random ercitation as complete assemblies but rather experience local disturbances ofindividual components within the assemblies. The modal analyses from the finite element models of these components, (Figures 3.9-4, 3.9-5 and 3.9-7) already used for deterministic analysis, are once again utilized to determine the random responses via the normal mode procedure. 3.9.2.4 Comprehensive Vibration Assessment Program (CVAP) In accordance with Regulatory Guide 1.20 (Reference 43), a CVAP is developed for System 80+. System 80 + is designated as non-prototype Category I, per Regulatory Guide 1.20, with Palo Verde Unit 1, a Combustion Engineering System 80 Reactor as the valid prototype (Reference 44). Palo Verde Unit (~')N ( I and System 80+ design are substantially the same with regard to arrangement design, size and operating conditions. A comparison of the design arrangement is provided in Table 3.9-17. The reactor vessel internals nominal dimensional comparison is shown in Table 3.9-18. The CVAP for System 80+ design consists of an Analysis and Inspection Program. The Analysis Program consists of dynamic analyses which will be documented in an ASME Design Stress Report. In addition, flow loads and structural responses for System 80+ are compared with System 80 to confirm System 80+ design is a non-prototype Category I reactor. The Inspection Program consists of a pre-hot functional and a post-hot functional inspection of the reactor l internals. The duration of the hot functional testing are established to ensure that 10E+6 cycles of vibration will have occurred before the post-hot functional inspection. A detailed inspection of major load bearing surfaces, contact surfaces, welds, and maximum stress locations identified in the Analysis Program are performed. Photographic documentation is taken of all observations made during the pre-and post-hot functional inspections. A comparison is mad of the structures to verify that no loss in structural integrity due to flow induced vibration has occurrrsi. j 1 The Analysis Program and Inspection Program together confirm the adequacy of the analysis prediction j techniques and the structural integrity of System 80+ design according to the guidance of Regulatory ) Guide 1.20. ! l 3.9.2.5 Dynamic System Analysis of the Reactor and CEDMs Under Faulted Conditions j im
- V; Dynamic analyses are performed to determine blowdown loads and structural responses of the reactor core support, internals structures and fuel to postulated pipe break and SSE loadings and to verify the i
Approwd Desen Material. Design of SSC Pope 3.9-25
System 80+ oesign contrat Document adequacy of their design. Because of Leak-Before-Break arguments, all main RCS loop pipe breaks and all major primary branch line pipe breaks have been eliminated from consideration of dynamic effects. Internal blowdown loads due to breaks in small primary side pipes (6 inch diameter and less are considered in the design of the reactor internals. The loads due to these small pipe breaks are combined with the SSE loads by the SRSS method, and are found to represent less than a 10% increase in the SSE loads. Stress intensities for faulted conditions are governed by reactor vessel response motions from SSE and major secondary side branch line pipe breaks. Dynamic analyses are performed to determine the structural response of the Class CS and internal structures to assure that the criteria of Table 3.9-14 is achieved for the appropriate combination of pipe break and SSE loads. 3.9.3 ASME Cr ' Lass 1, 2 and 3 Components, Component Supports and Class CS Core Support > - . es ASME B&PV Code Section III Class 1,2 and 3 Piping and Components are designed and constructed in accordance with Section Ill of the ASME Boiler and Pressure Vessel Code and Code Case (s). In accordance with ASME Code, a specification is provided for piping suppons which defines the jurisdictional boundary for the NF portion of the piping support. For equipment component supports, such as those for pumps and vessels, the supports are generally furnished by the manufacturer along with the equipment. The supports are designed and classified and meet ASME Code Section 111, Subsection NF. Welding activities shall be performed in accordance with the requirements of Section III of the ASME Code. Component supports shall be fabricated in accordance with the requirements of Subsection NF of Section III of the ASME Code. Welding activities for A500 Grade B tube steel shall be performed in accordance with the requirements of AWS DI.1, "Stmetural Welding Code," (Reference 52). Visual weld acceptance criteria shall be per the Nuclear Construction Issue Group (NCIG) standard NCIG-01 (Reference 51). Reactor coolant loop piping and associated components and component supports rre designed and analyzed by Combustion Engineering. Loading conditions, stress limits, design transients, and methods of analysis for ASME Code Class I reactor coolant loop piping and associated components and component supports are discussed in Section 3.9.3.1. The site-specific SAR will provide information on the specific edition of the ASME Code used in the site-specific design. See also Section 1.8. 3.9.3.1 Loading Combinations, Design Transients and Stress Limits The loading combinations specified for the design ASME B&PV Section III Code Class I components, supports, and piping are categorized as normal, upset, emergency and faulted. The following specific loading combinations are r,pecified for design:
- The concurrent loadings associated with the Level-A (normal) plant conditions of dead weight, pressure and the thermal and expansion effects during startup, hot standby, power operation and normal shutdown to cold shutdown conditions.
O Attwoved Design Morenal- Design of SSC Page 19-26
..- . . - . . - - ~ . - - - . - - . - . - . - . . - . -- .-- . -
Srtem 80+ oeshwr conerat Document j i
- The concurrent loadings associated with either the normal plant condition or the Level-B (upset) {
'*g plant condition. The vibratory motion of the Safe Shutdown Earthquake (SSE) is included in the j
fatigue evaluation in accordance with Section 3.7.3.2. +
* ' The concurrent loadings associated with the Level-C (emergency) condition. ;
i
- The' concurrent loadings associated with the Level-A (normal) plant condition, the vibratory l motion of the SSE, and the dynamic system loadings associated with the Level-D (faulted) system l condition (postulated pipe rupture for branch line breaks not eliminated by leak before break l l
. analysis). flThe SSE andpipe rupture loadings are combined by the SRSS method in accordance l with the guidelines of NUREG-0484, Rev.1,1980 or by a more consermtive method.}}2 ;
The loading ~ The specific design transients specified for design are discussed in Section 3.9.1.1. combinations of the section and the stress limits associated with them, as defined in the Code, also apply i j
- to the internals parts which are essential to the component in performing its safety fmx. tion. j i
I . ASME B&PV Code Class 1, 2 ~ and 3 piping and components of fluid systems ne designed and { constructed in accordance with Section III of the ASME Boiler and Pressure Vessel Code. Hydrostatic .i testing is performed per Section III. j i Design pressure, temperature, and other loading conditions that provide the bases for design of fluid j systems are presented in the sections which describe the systems. _ i ' I
. Stress analysis is performed to determine structural adequacy of pressure components under the operating conditions of normal, upset, emergency or faulted, as applicable. ;
Significant structural discontinuities such as nozzles and flanges are considered. In addition to the design 1 calculation required by Section III of the ASME B&PV Code, stress analysis is also performed by l methods outlined in the Code appendices or by other methods by reference to analogous codes or other published literature. l i 3.9.3.1.1 ASME Code Class 1 Components and Supports . . Design transients for core support structures and ASME Code Class I components, supports and piping are discussed in Section 3.9.1.1. Loading combinations for ASME Code Class I components are described in Table 3.9-2. Stress limits for ASME Code Class I components, supports and piping are l described " Table 3.9-3. The operating pressures of Code Class 1 active valves are limited to the j pressures taxen from the applicable primary pressure class pressure-temperature rating of the ASME l Code, Section III, for the maximum temperatee for the applicable condition. j i , l 3.9.3,1,2 Core Support Structures (Class CS) and Internal Structurus (Class IS) :
' Design transients for core support structures and reactor internals structures are discussed in Section j 3.9.1.1; Loading combinations and stress limits are presented in Section 3.9.5. .
l
.2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction
- Section 3.5.
l Anpresent DenQn Meenrief Dee@n et SSC Page 3.9 27
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Syatem 80+ Design ControlDocument 3.9.3.1.3 ASME Code Class 2 and 3 Components and Supports Loading combinations applicable to Code Class 2 and 3 components and supports are described in Table 3.9-2. System operating conditions due to the design transients defined in Table 3.9-1, as well as any other auxiliary system specific conditions, are reviewed to determine the appropriate operating parameters to be used in the design of Code Class 2 and 3 components. The design stress limits for each of the loading conditions of the components are presented in Tables 3.9-5 through 3.9-9. Inelastic methods, as permitted by ASME Section III for Class 1 components, are not used for these components. Class 2/3 components are reviewed for thermal fatigue effects using the ASME Code Section Ill NC-3219.2, for guidance. Fatigue analysis is perfctmed in accordance with NC-3200 for these components which do not meet the NC-3219.2 criteria. 3.9.3.1.3.1 Tanks, Heat Exchangers, and Filters Pressure vessels supplied for the auxiliary systems are: Shutdown Cooling Heat Exchanger Safety injection Tanks Containment Spray Heat Exchanger Containment Spray Mini-Flow Heat Exchanger Shutdown Cooling Mini-Flow Heat Exchanger Component Cooling Water System Heat Exchangers Component Cooling Water System Surge Tanks Essential Chilled Water Compression Tanks Essential Chilled Water Refrigeration Units Diesel Generator Fuel Oil Storage Tank Diesel Generator Fuel Oil Day Tank Diesel Generator Cooling Water Surge Tank Diesel Generator Starting Air Aftercoolers Diesel Generator Starting Air Filter / Dryer Units Diesel Generator Starting Air System Air Receivers Diesel Generator Lube Oil Cooler Diesel Generator Lube Oil Sump Tank Heaters Diesel Generator Intake Turbocharger Diesel Generator Exhaust Aftercooler Diesel Generator intake and Exhaust Silencers and Air Filters Main Control Room Air Handling Units w/ Filters Main Control Room Water-cooling Coils Main Control Room Heating Coils Fuel Building Ventilation Exhaust Filter Train Reactor Building Subsphere Ventilation System Cooling Coils Reactor Building Subsphere Ventilation System Filters Annulus Ventilation System Filters Spent Fuel Pool Cooling System Heat Exchangers Station Service Water Strainers Vessel assemblies, including supports, support attachment welds, and anchor bolts, are capable of withstanding specified horizontal and vertical seismic accelerations. The seismic accelerations are applied separately at the center of gravity acting in each of two orthogonal horizontal directions and either vertical Appweved Design Matenal. Des @n of SSC Page 3.9-28
Srtem 80+ Design controlDocument
,3 direction. The stresses or reaction loads at a given point, due to the three separate analyses, are is '~' ) combined by the SR.C.s method to define a total seismic design condition. The design allowable nozzle forces and momentr act in directions that yield the highest stress which combined with the seismic loads, as determined above, and other concurrent loads.
Vessel components not subject to fluid pressure, such as supports, attachment welds, and anchor bolts, are designed to the stress criteria of ASME B&PV Code, Section III for *he loading conditions defined above. In cases where the natural frequency could not be increased to avoid amplification of the floor response of the postulated seismic input for a specific plant, the components are modeled as multi-mass systems, and their modal frequencies and maximum reactions are determined from the floor response spectra iv.- the plant. The maximum damping value used is 3% for SSE. The design point reactions due to each modal loading are combined as the sum of the absolute values or by root sum square of the modal reactions, as appropriate per recommendation of Regulatory Guide 1.92. 3.9.3.1.3.2 Valves For Class 1 valves, loading combinations are in accordance with Table 3.9-2. Stress limits are in accordance with Note (a) of Table 3.9-3 for active valves and Tabie 3.9-3 for inactive valves. ASME Class 2 and 3 valves are designed by analysis to standard rules. Stress limits are shown in Table 3.9-9 for active valves and in Table 3.9-8 for inactive valves. () 3.9.3.1.3.3 Pumps G Pumps supplied for the Auxiliary Systems are: Safety injection (active) (Safeguard) Code Class 2 Shutdown Cooling (active) (Safeguard) Code Class 2 Containment Spray (active) (Safeguard) Code Class 2 Component Cooling Water System Pumps (active) Code Class 3 Station Service Water System Pumps (active) Code Class 3 Essential Chilled Water Circulation Pumps (active) Code Class 3 Diesel Generator Fuel Oil Recirculation Pump (active) Code Class 3 Diesel Generator Fuel Oil Booster Pump (active) Code Class 3 Diesel Generator Cooling Water Circulation Pump (active) Code Class 3 Diesel Generator Cooling Water Keep Warm Pump (active) Code Class 3 Diesel Generator Starting Air System Air Compressors (active) Code Class 3 l Diesel Generator Lube Oil Transfer Pumps (active) Code Class 3 ) Diesel Generator Prelube Oil Pump (active) Code Class 3 l Reactor Building Subsphere Sump Pumps (active) Code Class 3 l Diesel Generator Building Sump Pumps (active) Code Class 3 ) Spent Fuel Pool Cooling System Pumps (active) Code Class 3 Emergency Feedwater Pumps (active) Code Class 3 The design rules and associated design stress limits applied in the design of ASME Code Class 2 and 3 l pumps are in accordance with the ASME Code, Section III, Subsections NC and ND, respectively. The (n) v results are as described herein. l Approved Design Maternel- Desogn of SSC Page 3.9-29
System 80+ Design ControlDocument Stress limits for active pumps are shown in Table 3.9-7 and stress limits for non-active pumps are shown in Table 3.9-6. Loading combinations are in accordance with Table 3.9-2. Pump assemblies, including supports, support attachment welds, and bolts, are capable of withstanding specified horizontal and vertical seismic accelerations. The seismic accelerations are applied separately at the center of gravity acting in each of two orthogonal horizontal directions and either vertical direction. The stresses or reaction loads at a given point, due to the three separate analyses, are combined by the SRSS method to define a total seismic design condition. The design allowable nozzle forces and moments act in directions that yield the highest stress when combined with the seismic loads, as determined above, and other concurrent loads. The stress criteria of the ASME Code, Section 111 are applied in the design of compenent supports to the same Code Class as the pressure boundary involved within the jurisdictional bound tries defined in the code for the loading conditions defined above. Those steel support structures whica are considered to be an extension of the building structure, but supplied with the pump assembly (i.e., bedplates), are designed to the stress criteria of the AISC Manual of Steel Construction. In addition, the Safeguard Pump assemblies are required to be capable of withstanding the design thermal transients of Section 3.9.1. # , 3.9.3.1.4 Piping and Piping Supports Piping systems classified as ASME Code Section ill Class 1,2 or 3 are designed to maintain dimensional stability and functional integrity under design loadings expected to be experienced during a 60-year design life. The as-built piping will be reconciled with the as-designed piping configurations. An as-built inspection of the pipe routing, location and orientation, the location, size, clearances and orientation of piping supports, and the location and weight of pipe mounted equipment will be performed. The inspection will be performed by reviewing the as-built drawings containing verification stamps, and by performing a visual inspection of the installed piping system. The piping configuration and component location, size, and orientation will be confirmed to be within the tolerances specified in the certified as-built piping stress report. The tolerances to be used for reconciliation of the as-built piping system with the as-designed piping system are provided in Reference 53. A reconciliation analysis using the as-built and as-designed information will be performed. The certified as-built stress report will document the results of the as-built reconciliation analysis. 3.9.3.1.4.1 ASME Code Class 1
- Piping For ASME Code Class 1 piping, the combinations of design loadings are categorized with respect to service levels, identified as Level A, Level B Level C, or Level D, as shown in Table 3.9-10.
The design stress limits for each of the loading combinations are found in ASME B&PV Code, Section 111, NB-3600. O App *oved Desh)n Materint Desogn of SSC Page 3.9-30
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Sy? tem 80+ Desio., tMirol Document
- Piping Supports O For pipe supports,- the design loading combinations are presented in Tables 3.8-5 and 3.9-12. i Pipe support members are designed to meet the requirements defined by ASME Code, Section Ill, Subsection NF. See Appendix 3.9A, Section 1.7.4, for a further discussion.
s3.9.3.1.4.2 ASME Code Class 2 and 3
- Piping For ASME Code Class 2 and 3 piping the combinations of design and service loadings are categorized with respect to system service levels identified as Design, Level A, B, C and D as shown in Tables 3.9-11. The design stress limits for each of the loading combinations are found in ASME B&PV Code, Section 111, NC/ND-3600.
L* Piping Supports k For pipe supports, the design and service loading combinations are presented in Tables 3.9-12. , Pipe support members are designed to meet the requirements defined by ASME Code, Section 111, Subsection NF. See Appendix 3.9A, Section 1.7.4, for a further discussion. 3.9.3.1.4.3 ll Functional Capability, ASME Code Cass 1, 2 and 3 To satisfy functional capability requirements, the piping is designed to meet a D/t ratio of <50 in accordance with NUREG-1367and the allowable stressfrom the ASME Code of 3.05,,, (Class I) or 3.0S, s (Class 2 and 3), but not greater than 2.05,.}}' l 3.9.3.1.5 Preparation of Design Specifications j l Design specifications for ASME Section Ill components and supports are prepared using guidelines given in Reference 42. , 3.9.3.2 Pump and Valve Operability Assurance 1 3.9.3.2.1 Active ASME Code Class 2 and 3 Pumps and Class 1,2 and 3 Valves Furnished l with the NSSS 3.9.3.2.1.1 Operability Assurance Program l Active pumps and valves are defined as pumps and valves and those components that must perform a mechanical motion in order to shut down the plant, maintain the plant in a safe shutdown condition, or , mitigate the consequences of a postulated event. The operability (i.e., performance of this mechanical motion) of active components during and after exposure to design bases events is confirmed by 1 1 I l 2
%._) NRC Staff approval is required prior to implementing a change in this information: see DCD Introduction Section 3.5.
. i Anweved Deere neeeuw- Den @n er ssc pape 2.s.3r
SYtem 80+ Design controlDocument
- Designing each component to be capable of performing all safety functions during and following design bases events. The design specification includes applicable loading combinations, and conservative design limits for active components. The specification requires that the manufacturer demonstrate operability by analysis, by test, or by a combination of analysis and test. The results are independently reviewed by the NSSS Supplier considering the effects of postulated failure modes on operability. Internale parts which are essential to the component in performing its safety function are analyzed aral/or tested as an assembled component to validate operability.
- Analysis and/or test demonstrating the operability of each design under the most severe postulated loadings. Methods /results of operability demonstration programs are detailed in Sections 3.9.3.2.1.2 and 3.9.3.2.1.3.
- Inspection of each component to assure compliance of critical parameters with specifications and drawings. This inspection confirms that specified materials and processes were used, that wall thicknesses met code requirements, and that fits and finishes met the manufacturer's requirements based on design clearance requirements.
- Shop testing of each component to verify "as-built" conditions, as defined in Sections 3.9.3.2.1.2 and 3.9.3.2.1.3.
- Startup and periodic in-service testing in accordance with ASME Boiler and Pressure Vessel Code, Section XI to demonstrate that the active pumps and valves are in operating condition throughout the life of the plant.
NSSS active pumps are listed below with a brief description of active safety function of each. NSSS active valves are listed in Table 3.9-4. Active Components Active Safety Function Safety injection pumps Operate at flow rates to runout Shutdown cooling pumps Operate at design flow Containment spray pumps Operate at design flow 3.9.3.2.1.2 Operability Assurance Program Results for Active Pumps Operability of the Safety Injection, Shutdown Cooling and Containment Spray pumps under required conditions demonstrated by analyses of the assemblies and by analyses and tests of the motors. For the safety injection, shutdown cooling and containment spray pumps, allowable stresses are not exceeded, clearances are acceptable and shaft and pedestal bolt deflections do not cause stresses to exceed the normal values. Where necessary, lumped mass models are used with the computer programs to determine the natural frequencies and displacements. The models are conservative (i.e., simplifications tend to make them more flexible). O l Approwd Design Matenal Design of SSC Page 3.9-32
Sy~ tem 80 + Design controlDocument To verify "as-built" conditions the pumps are hydrostatically tested in accordance with the ASME B&PV (n) Code, Section III to confirm acceptability of structural integrity of pressure retaining parts, tested for seal leakage, and tested for performance and NPSH characteristics in accordance with the Hydraulic Institute Standard to verify operation within specified parameters. The motors are Class IE and are tested in accordance with IEEE Standard 112A-1978 to verify operation within specified parameters. Additionally, IEEE Standard 323-1974, as endorsed by Regulatory Guide 1.89, and IEEE Standard 344-1987 as endorsed by Regulatory Guide 1.100 Rev. 2 dated June 1988, are applicable for motors to asst;re operability during and following design basis events. 3.9.3.2.1.3 Operability Assurance Program for Active Valves Safety-related active valves must perform their mechanical motion during or after design basis events. The qualification program is in accordance with IEEE Standard 344-1987 as endorsed by Regulatory Guide 1.100, Rev. 2 dated June 1988, and assures that these valves will operate during a seismic event. Qualification tests and/or analyses are conducted for all active valves. Class 1, 2 and 3 valves are designed / analyzed according to the rules of the ASME Boiler and Pressure Vessel Code, Section III, Section NB-3500, NC-3500, and ND-3500 respectively. Procurement specifications for safety-related active valves conform to the intent of Regulatory Guide 1.148 and stipulate that the valve vendor shall submit either detailed calculations and/or test data to demonstrate operability when subjected to the specification loading and stress criteria (normal through faulted conditions). The decision to accept actual or prototype test data, or analysis for operability assurance is made during the normal design and procurement process. The decision to test is based on:
,a Q e Whether the component is amenable to analysis.
- Whether proven analytical methods are available.
e Whether applicable prototype test data is available. If analysis or prototype test data is not sufficient, testing is conducted to qualify the component or to verify the analytical technique. Where appropriate, valve stem deflection calculations are performed to determine deflections due to short tenn seismic and other applicable loadings. Deflections so determined are compared to allowable clearances. It must be noted that seismic events are of short duration; thus, contact (if it occurs) does not demonstrate that operability is adversely affected. Cases where contact occurs are reviewed on a case by case basis to determine acceptability. The operability of active Code Class 1,2 and 3 components is assured through an extensive program of design verification, qualification testing and thorough surveillance of the manufacturing, assembly and shop testing of each active component. Each aspect of the design related to pressure boundary integrity and operability is either tested or verified by calculations. Procedures for testing are developed by component manufacturers and reviewed and approved before the tests are conducted. The design analyses of the component take into consideration environmental conditions including loadings developed from seismic, operational effects, and pipe loads. Where necessary and feasible, the conclusions of these analyses are confirmed by test. N ,A Approvest Desbyn Atatenet Design of SSC Page 3.9 33
Syatem 80+ Design ControlDocument j On all active valves, an analysis of the extended structure is also performed for static equivalent SSE , loads supplied at the center of gravity of the extended structure. The maximum stress limits allowed in these analyses show that structural integrity is within the limits developed and accepted by the ASME Code. The safety-related valves are subjected to a series of tests prior to service and during the plant life. Prior to installation, the following tests are performed:
- Shell hydrostatic test to ASME Section III requirements.
- Backseat and main seat leakage tests.
- Disc hydrostatic test.
- Functional tests to verify that the valve opens and closes as required when subjected to the design differential pressure and flow.
When it is not possible, due to size of valve and unavailability of test facilities, to accomplish sustained flow and pressure during opening or closing test, the valves are tested as fellows:
- 1. Closure test: The timing of flow initiation and closure initiation is such that during closure of the final 10% of valve flow area the differential pressure is at least equal to the maximum differential pressure for which the valve is being qualified.
- 2. Opening test: Rated differential pressure is established across the valve. Opening is then initiated, but full differential pressure may not be maintained after the valve is unseated.
Closing and opening times are measured.
- Operability qualification of motor operators for the environmental conditions over the installed life (i.e., aging, radiation, accident environment simulation) according to IEEE Standard 382-1972, as endorsed by Regulatory Guide 1.73.
Cold hydro qualification tests, hot functional qualification tests, periodic in-service inspections, and periodic in service operation are performed in situ to verify and assure the functional ability of the valves. These tests ensure the reliability of the valve for the design life of the plant. The valves are designed using either stress analyses or the pressure containing minimum wall thickness requirements. All the active valves are designed to have a first natural frequency which is greater than the ZPA. This is shown by suitable test or analysis. All manually-controlled, electrically-operated valves will be subject to 10 CFR 50.59 evaluations to detennine which valves Branch Technical Position ICSB 18 (PSB) may apply. All valves which are evaluated as possibly performing an " undesirable function" will be outlined in individual plant system technical specifications. In certain cases where inclusion of such valves is inappropriate, such as where manipulations are required for maintenance,10 CFR 50.59 operability determinations will be performed in accordance with plant procedures and operational directives. O Approved Desigrt Materid - Design of SSC Page 3.9-34
System '80 + Deslan controlDocument
~
c The position indication system for all safety-related valves.will be designed to meet the single failure
. criterion of Regulatory Guide 1.47.
In addition to the above, the following specific operability assurances are provided for the various type
- valves:
t 1 3.9.3.2,1.3.1 P====ticaHy Operated Valves ,
- Pneumatically operated valves are furnished by several vendors. Methods of operability demonstration i
[ ~ are summarized below. Spring actuation of the valve is the required active safety function. Loss of . ' electric power or supply air results in venting of the actuator and return of the valve to the safe position.
.Each vendor provides their own method to demonstrate valve operability. The operability for these valves is demonstrated by analysis, test or by a combination of analysis and test. The vendor considers :
i concurrent loads including seismic, design pressure and pipe loads. j 1
- The three-way solenoid valve was qualified by test and analysis to IEEE Standard 382-1972, as endorsed j by Regulatory Guide 1.73, IEEE Standard 323-1974 and IEEE Standard 344-1987 as endorsed by
. Regulatory Guide 1.100, Rev. 2 dated June 1988. Testing included thermal aging, radiation aging, wear aging, vibration endurance, seismic event simulation, and loss-of-coolant-accident. All test results provide satisfactory evidence of air solenoid valve operability.
i Limit switches, used to determine valve position, were qualified by testing and analysis to IEEE Standard ;
- j. 323-1974, IEEE Standard 344-1987 as endorsed by Regulatory Guide 1.100, Rev. 2 dated June 1988 and )
- IEEE Standard 382-1972. Switches are performance tested for aging simulation, wear aging, radiation ; exposure, seismic qualification, and design basis event environmental conditions. For valves outside of j
- containment and utilizing EA-170 limit switches, the switches are seismically qualified to IEEE Standard j 344-1987 as endorsed by Regulatory Guide 1.100, Rev. 2 dated June 1988 and were tested to sustain l radiation dosages up to 2 x 108 rads. l l
3.9.3.2.1.3.2 Motor Operated Valves -j l l Motor-operated valves are qualified by analysis as a minimum as described above. The analysis for each ; + valve assembly considers the effects of seismic loads, design pressure, and piping reaction forces to provide assurance of operability. : I 1 To provide full qualification of the motor-operated valve actuator, environmental and seismic qualification tests are conducted to simulate the following conditions: l.
- e. Inside Containment (LOCA) ;
e Outside Containment t e Seismic Qualification j e ' Steam Line Break Accident Mid-size valve actuators are subjected to complete environmental qualification consisting of inside ! contairunent and outside containment. Each qualification exposed the actuator to thermal and mechanical
~ aging, radiStion aging, seismic aging, environmental transient profile test, and steam line break. For the O' steam line break test an actuator is subjected to a very high superheated temperature to demonstrate that
-i t
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_ - _ d Ossen Aasender Doo@n of SSC page 19J5
.f
l l System 80+ Design ControlDocurnent the electrical components of the actuator never exceeds the saturated temperature corresponding to the ambient pressure for the short duration of the test. This short-term test provides evidence that the , existing qualification envelopes the steam line break for superheated temperatures as high as approximately 492*F for a few minutes (see Section 3.11). The qualification of the mid-size valve actuator is used to generically qualify all sizes of mid-size valve actuator operators for t6 envirotmental test conditions in accordance with IEEE Standard 382-1972. All sizes are constructed of the same ma:erials with components designed to equivalent stress levels, and to the same clearances and tolerances with the only difference being in physical size which varies corresponding to the differences in unit rating. All the qualifications are conducted per IEEE Standard 382-1972 and meet the requirements of IEEE Standard 323-1974 and IEEE Standard 344-1987 as they apply to valve motor actuators. Further, since the actuators perform satisfactorily without maintenance throughout the various qualifications, the valve actuators are fully qualified for use in CE Nuclear Power Generating Plants. 3.9.3.2.1.3.3 Pressurizer Safety Valves Pressurizer Safety valves are 6 x 8 valves. Operability has been successfully demonstrated by a combination of dynamic testing and analysis or by static testing. Operability was successfully demonstrated with a 6g seismic load by one vendor or with a 7.lg seismic load by another vendor. Dynamic testing of System 80+ valves is performed to demonstrate that the natural frequency of both valves is greater than the frequency at ZPA. A summary of the previous programs follows: e Vendor A Safety Valves
- 1. Natural Frequency Demonstration Vibration input wcs in it single, horizontal direction. It was established by previous experience that the horizontal direction was more significant than the vertical direction, and that there was no material difference between the various horizontal directions. The frequency of vibration was increased from 5 to 75 Hz at a rate of I octave per minute.
Accelerometers were mounted on the valve assembly. The actual natural frequency under test conditions was 38 Hz.
- 2. Operability Demonstration A series of tests demonstrated that the valve would fully open and rescat during and after a seismic acceleration. Vibration input ranged from 3 to 6g and 10 to 33 Hz. The tests were performed using saturated steam. In addition, analysis was used to establish the significance of nozzle loading. The results indicated that deformation was significantly less than the internal clearances. This loading was, therefore, neglected in the seismic operability tests.
O Approved Oesse n Matenel Design of SSC Page 3.9-36
Sy~ tem 80 + Design Control Document g3
- Vendor B Safety Valves
( "j Natural Frequency Demonstration 1. A resonance survey was performed along three orthogonal axes with one axis being the centerline of the outlet port. (Valve mounted on inlet port.) No resonant frequencies were detected in the range of 1-50 Hz on any axis.
- 2. Operability Demonstration A series of tests demonstrated that the valve would fully open and reseat during and after applying the following loading combinations: Static seismic loads up to 7.lg were applied to the valve in the direction of least bending stiffness. In addition the maximum permissible piping loads were applied concurrently. The tests were performed using saturated steam. Valve operation was satisfactory.
- EPRI Testing of Safety Valves Pressurizer safety valves were tested in the EPRI Test Program under full pressure and full flow conditions. This testing demonstrated that stable valve operation under these conditions is dependent upon the inlet pipe configuration, built up back pressure range and blowdown setting.
Prior to valve shipment, the inlet pipe configuration and built up back pressure range for the specific plant are examined by CE and the applicable valve vendor. If necessary, the valves are adjusted to provide blowdown settings which will result in stable valve operation. These blowdown settings are recommended by the vendor and approved by CE. These adjustments are () based on the results obtained in the EPRI Test Program. V Required adjustments to the valve to assure operability will be documented in plant-specific information. 3.9.3.2.1.3.4 Check Valves The check valves are characteristically simple in design and their operation is not affected by seismic accelerations or the maximum applied nozzle loads. The check valve design is compact and there are no extended structures or masses whose motion could cause distortions which could restrict operation of the valve. The nozzle loads due to maximum seismic excitation do not affect the functional ability of the valve since the valve disc is designed to be isolated from the casing wall. The clearance supplied by the design around the disc prevents the disc from becoming bound or restricted due to any casing distortions caused by nozzle load. Therefore, the design of these valves is such that once the structural integrity of the valve is assured using standard design or analysis methods, the ability of the valve to operate is assured by the design features. In addition to these design considerations, the valve also undergoes:
- Stress analysis, including the SSE loads.
- In-shop hydrostatic tests.
- In-shop seat leakage test.
)
- Periodic in-situ valve exercising and inspection, to assure the functional ability of the valve.
Approved Design Meterial- Design of SSC Page 3.9-37
System 80+ Design controlDocument 3.9.3.2.2 Non-NSSS Active ASME Code Class 2 and 3 Pumps and Class 1,2 and 3 Valves 3.9.3.2.2.1 Pumps Safety-related active pumps are subjected to in-shop tests that include hydrostatic tests of casing to 150% of the design pressure, and performance tests to determine total developed head, net positive suction head (NPSH) requirements, and other pump / motor characteristics. Vibration is monitored during the performance tests. In addition to the required testing, the pumps are designed and supplied in accordance with the following specified criteria:
- In order to ensure that the active pump will not be damaged during the seismic event, the pump manufacturer is required to demonstrate by test and/or analysis that the lowest natural frequency of the pump is greater than the ZPA. The pump, when having a natural frequency above the EPA, will be considered essentially rigid. This frequency is considered sufficiently high to avoid problems with amplification between the component and structure for all seismic areas. A static shaft deflection analysis of the rotor is performed. The natural frequency of the support is determined and used in conjunction with the plant seismic response spectra. The deflection determined from the static shaft analysis is compared to the allowable rotor clearances. The pump manufacturer is required to demonstrate the pump operability during and after the SSE.
If the natural frequency is found to be below the lowest ZPA frequency, an analysis is performed to determine the amplified input accelerations necessary to perform the static analysis. The static deflection analyses are performed using the adjusted accelerations.
- The maximum seismic nozzle loads are also considered in an analysis of the pump supports to ensure that unacceptable system misalignment cannot occu
- To complete the seismic qualification procedures, the pump motor and all appurtenances vital to the operation of the pump are independently qualified for operation during the maximum seismic event in accordance with IEEE Standard 344-1987. If the testing option is chosen, sine-beat or sweep testing for the electrical equipment is justified by satisfying one or more of the following requirements to demonstrate that multi-frequency response is negligible or that the sine-beat or sine-sweep is of sufficient magnitude to conservatively account for this effect:
- 1. The equipment response is basically due to one mode.
- 2. The sine-beat response spectrum in the region of significant response.
- 3. The floor response spectrum consists of one dominant mode and has a narrow peak at this frequency.
The degree of coupling in the equipment, in general, determines if a single or multiaxis test is required. Multiaxis testing is reauhed if there is considerable cross-coupling. If coupling is very light, then single-axis testing is justified or, if the degree of coupling can be determined, then single-axis testing can be used with the input sufficiently increased to include the effect of coupling on the response of the equipment. O Approved Design htaterial Design of SSC Pope 3.9-38
- - . ---- . - . . . . . ~_- --_
Sy: tem 80+ Design ControlDocument From this, it is concluded that the safety-related pump / motor assemblies manufacturer show that it will Q not be damaged and will continue operating under SSE loadings and will perform their intended functions.
- 3.9.3.2.2.2 Valves ,
- Safety-related active valves are subjected to the following tests:
- Shell hydrostatic tests, in accordance with ASME B&PV Code, Section III requirements.
- Backseat and main seat leakage tests.
- Disc hydrostatic tests.
- Functional tests that verify that the valve will open and close as required when subjected to the ,
design differential pressure and flow. When it is not possible, due to size of valve and unavailability of test facilities, to accomplish sustained flow and pressure during opening or closing test, the valves are tested as follows: 6
- 1. Closure test: The timing of flow initiation and closure initiation is such that during closure of the final 10% of valve flow area the differential pressure is at least equal to H aaaximum differential pressure for which the valve is being qualified.
- 2. Opening test: Rated differential pressure is established across the valve. Opening is then initiated, but full differential pressure may not be maintained after the valve is unseated.
g ( Closing and opening times are measured.
- Operability qualification of motor operators for the envirenmental conditions over the installed life (i.e., aging, radiation, accident, environment simulation) in accordance with IEEE Standards 323-1974,344-1987, and 382-1972.
After installation, cold hydrostatic tests, hot functional tests, and periodic inservice operation are i performed to verify and assure the functional ability of the valve. These tests enhance reliability of the , valve for the design life of the plant. I l The valves are designed using either stress analysis or standard design rules for minimum wall thickness j 1 requirements. On all active valves with extended topworks, an analysis is also performed for static l l equivalent SSE loads applied at the center of gravity of the extended structure. The maximum stress limits allowed in the analyses are those required by the ASME Code for the particular ASME Class of valve analyzed, in addition to these tests and analyses, valves are tested for verification of operability during a simulated seismic event by demonstrating operational capabilities within the specified limits. The valve is mounted in a manner that represents typical valve installation. The valve unit includes the operator and all appurtenances normally attached to the valve in service. The operability of the valve during SSE is demonstrated by satisfyittg the following criteria: 4 .
- i
= = onw unaw. one or ssc rose 3.s ss
System 80+ Design ControlDocument
- All the active valves with extended topworks are designed to have a first natural frequency greater than the lowest ZPA frequency. This may be shown by test and/or analysis. Valves with a first natural frequency less than the lowest ZPA frequency are discussed below.
- While in the shop and installed in a suitable test rig, the extended topworks of the valve are subjected to a statically applied equivalent seismic load. The load is applied to the valve assembly in the direction of the weakest axis of the yoke. The design pressure of the valve is simultaneously applied to the valve during the static load tests.
- The valve is then operated with the equivalent seismic static load applied (i.e., from the normal operating status to the faulted operating status). The valve must perform its safety-related function within the specified operating time limits. Three full-stroke operations are required.
- Motor opera:03 and other electrical appurtenances necessary for operation are qualified as operable during toe SSE by IEEE Standard 344-1987, Seismic Qualification Standards, prior to their installation on die v;dve.
The piping designer supports the piping in such a way that the equivalent seismic static load accelerations are not exceeded at the valve inlet and outlet support points. If the frequency of the valve with topworks, by test or analysis, is less than ZPA, a dynamic analysis of the valve is performed to determine an equivalent acceleration that is to be applied during the static test. The analysis provides the amplification of the input acceleration considering the natural frequency of the valve and frequency content of the plant floor response spectra. The adjusted accelerations are determined using margins similar to that contained in the horizontal and vettical accelerations used for " rigid" valves. The adjusted accelerations are used in the static analysis, and valve operability is assured by the methods outlined in listings B to D above, using the modified acceleration input. The above testing program applies only to valves with overhanging structures (e.g., the operator). The testing is conducted on a representative number of valves. Valves from each of the prunary safety-related design types (e.g., motor-operated gate valve) are tested. Specific valves are qualified by the tests, and the results are extended to qualify valves within a range of sizes. An analysis is conducted to prove the similarity between the tested valve and the installed ones. Due to the simple c? *acteristics of check valves and other compact valves, they are qualified by the following tests and t. mysis:
- Stress analysis of the attached piping for SSE loads.
- In-shop hydrostatic test.
- In-shop seat leakage test.
- Feriodic valve exercise and inspection to assure the functional ability of the valve.
Using the methods described, safety-related active valves in the system are qualified for operability during a seismic event. O Astroved Des @ Matenel- Desigrr of SSC Page 3.940
Sy: tem 80+ oesign controlDocument
.q 3.9.3.3 Design and Installation Details for Mounting of Pressure Relief Devices l Safety valves and relief valves are analyzed in accordance with the ASME Section III Code.
llThe method of analysisfor safety valves and relief valves suitably accountsfor the time-history ofloads i acting immediatelyfollowing a valve opening (i.e., firstfew milliseconds). The fluid-inducedforcing i functions are calculatedfor each safety valve and relief valve using one-dimensional equationsfor the ; t conservation ofmass, momentum, and energy. The calculatedforcingfunctions are applied at locations along the associatedpiping where a change influidflow direction occurs. Application of theseforcing l functions to the associatedpiping model constitutes the dynamic time-history analysis.}}' The dynamic j response of the piping system is determined from the input forcing functions. Therefore, a dynamic amplification factor is inherently accounted for in the analysis. (( Alternately, an equivalent static analysis e i may be usedfollowing the criteria of Appendix 0 of the ASME Code Section III as supplemented by the additional criteria of SRP3.9.3, Section 11.2.}}2 l Snubbers or strut-type restraints are used as required. The stresses resulting from the loads produced by the sudden opening of a relief or safety valve are combined with stresses due to other pertinent loads and are shown to be within allowable limits of the ASME Section Ill Code. Also, the analyses show that the loads applied to the nozzles of the safety and relief valves do not exceed the maximum loads specified { by the manufacturer. Jurisdictional boundaries between ASME Section III Class 1,2 and 3 component supports and the building structure are established in accordance with ASME Section III, Subsection NF. 3.9.3.4 Component Supports Jurisdictional boundaries between ASME Section Ill Class 1, 2 and 3 component supports and the
) building structure are established in accordance with ASME Section III, Subsection NF.
ASME B&PV Code Section III Class 1, 2 and 3 component supports are designed and constructed in accordance with Section III of the ASME B&PV Code and Code Case (s). Seismic Category I component supports are designed to meet the requirements of Subsection NF, Section III of the ASME Code. Welding fabrication and installation, nondestructive examination (NDE) and , acceptance standards shall be in accordance with Subsection NF, Section III of the ASME Code. In addition, visual weld acceptance criteria shall be per the Nuclear Construction Issue Group (NCIG) standard NCIG41 (Reference 51). Radiographic examinations will be accepted by a nondestructive examination (NDE) Level III examiner prior to final acceptance. See also Section 3.9. i Confirmation that facility welding activities are in compliance with the certified design commitments shall include verifications of the following by individuals other than those who performed the activity:
- 1. Facility welding specifications and procedures meet the applicable ASME Code requirements. ;
)
- 2. Facility welding activities are performed in accordance with the applicable ASME Code 1 requirements.
i i (]M 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5, i l G.-2 Des @n n0eteniel- Des &n of SJC Page 3.9 41 l
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j
System 80+ Design controlDocument \
- 3. Welding activities related records are prepared, evaluated and maintained in accordance with the ASME requirements.
- 4. Welding processes used to weld dissimilar base metal and welding filler metal combinations are compatible for the intended applications. ,
- 5. The facility has established procedures for qualifications of welders and welding operators in accordance with the applicable ASME Code requirements.
- 6. Approved procedures are available and are used for pre-heating and post-heating of welds, and those procedures meet the applicable requirements of the ASME Code.
- 7. Completed welds are examined in accordance with the applicable examination method required by the ASME Code.
Supports for ASME Section III Code Class 1,2 and 3 components are specified for design in accordance with the loads and loading combinations discussed in Section 3.9.3.1 and presented in Table 3.9-2. Component support building structures are designed to meet the criteria in Appendix 3.8A. Component supports uhich are loaded during normal operation, seismic and following a pipe break (branch line breaks not eliminated by leak-before-break) are specified for design for loading combinations of Section 3.9.3.1. Design stress limits applied in evaluating loading combinations for Level A, B or C plant conditions described in Section 3.9.3.1 are consistent with the ASME B&PV Code, Section Ill. The design stress limits applied in evaluating loading combination for Level D plant conditions described in Section 3.9.3.1 are in accordance with the ASME B&PV Code, Section 111. Loads in compression members are limited to 2/3 of the critical buckling load. Concrete expansion anchors meet the requirements of ACI-349, " Code Requirements for Nuclear Safety Related Concrete Structures" and IE Bulletin 79-02, Rev. 02, " Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts", November 8,1979, with the provisions identified in Section 3.8.4.5 and further discussed in Appendix 3.9A. See Appendix 3.9A, Section 1.7.4, for a discussion of concrete expansion anchors. Where required, snubber supports are used as shock arrestors for safety-related systems and components. Snubbers are used as structural supports during a dynamic event such as an earthquake or a pipe break, but during normal operation act as passive devices which accommodate normal expansions and contractions of the systems without resistance. For System 80+, snubbers are minimized, to the extent practical, through the use of design optimization procedures. Assurance of snubber operability is provided by incorporating analytical, design, installation, in-service, and verification criteria. The elements of snubber operability assurance for System 80+ include:
- Consideration of load cycles and travel that each snubber will experience during normal plant operating conditions.
- Verification that the thermal growth rates of the system do not exceed the required lock-up velocity of the snubber.
Appresed Denngn Meterial Design of SSC Page 3.942
l Sy~ tem 80 + Design ControlDocument
- Accurate characterization of snubber mechanical properties in the structural analysis of the 7m
( ) snubber-supported system.
- For engineered, large bore snubbers, issuance of a design specification to the snubber supplier, describing the required structural and mechanical performance of the snubber; verification that the specified design and fabrication requirements are met.
- Verification that snubbers are properly installed and operable prior to plant operation, through visual inspection and through measurement of thermal movements of snubber-supported systems during start-up tests.
- A snubber in-service inspection and testing program, which includes periodic maintenance and visual inspection, inspection following a faulted event, a functional testing program, and repair or replacement of snubbers failing inspection or test acceptance criteria.
Site-specific information will include a listing of all safety-related components which utilize snubbers, in accordance with SRP 3.9.3. ((Energy absorbing and/or non-linear piping restraints may be used on System 80+. If used, a description of the methodology used to analyze and design thepiping systems incorporating these elements will be provided with site-spectpc information.}}2 3.9.4 Control Element Drive Mechanisms 3.9.4.1 Descriptive Information of CEDM (G a i The control element drive mechanism (CEDMs) are magnetic jack type drives used to vertically position and indicate the position of the control element assemblies (CEAs). Each CEDM is capable of withdrawing, inserting, holding, or tripping the CEA from any point within its 153-inch stroke in response to operation signals. The CEDM is designed to function during and after all normal plant transients. The CEA drop time for 90% insertion is 4.0 seconds maxiraum. The drop time is defined as the interval between the time power is removed from the CEDM coils to the time the CEA has reached 90% of its fully inserted position. The CEDM pressure boundary components have a design life of 60 years. The CEDM is designed to operate without maintenance for a minimum of 1-1/2 years and without replacing components for a minimum of 3 years. The CEDM is designed to function normally during and after being subjected to seismic loads. The vibratory motion 'of the Safe Shutdown Earthquake is included in the fatigue evaluation in accordance with Section 3.7.3.2. The CEDM will allow for trippiag of the CEA during and after a Safe Shutdown Earthquake. The design and construction of the CEDM pressure housing fulfill the requirements of the ASME boiler and Pressure Vessel Code, Section III, for Class 1 vessels. The CEDM pressure housirigs are part of the reactor coolant pressure boundary, and they are designed to meet stress requirements consistent with i those of the vessel. The pressure housings are capable of withstanding, throughout the design life, all ; normal operating loads, which include the steady-state and transient operating conditions specified for the i f} qj 2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Apweved oesogn Matener . oesogn or ssc rope 3.s43
System 80+ Design ControlDocument vessel. Mechanical excitations are also defined and included as a normal operating load. The CEDM pressure housings are service rated at 2500 psi at 650*F. The loading combinations and stress limit categories are presented in Table 3.9-14 and are consistent with those defined in the ASME code. The design duty requirements for the CEDM is a total cumulative CEA travel of 100,000 feet operation without loss of function. The test programs performed in support of the CEDM design are described in Section 3.9.4.4. 3.9.4.1.1 Control Element Drive Mechanism Design Description The CEDMs are mounted on nozzles on top of the reactor vessel closure head. The CEDMs consist of the upper and lower CEDM pressure housings, motor assembly, coil stack assembly, reed switch assemblies, and extension shaft assembly. The CEDM is shown in Figure 3.9-8. The drive power is supplied by the coil stack assembly, which is positioned around the CEDM housing. Two position indicating reed switch assemblies are supported by the upper pressure housing shroud, which encloses the upper pressure housing assembly. The lifting operation consists of a series of magnetically operated step movements. Two sets of mechanical latches are utilized engaging a notched extension shaft. To prevent excessive latch wear, a means has been provided to unload the latches during the engaging operations. The magnetic force is obtained from large de magnet coils mounted on the outside of the lower pressure housing. Power for the electromagnets is obtained from two separate supplies. A control programmer actuates the stepping cycle and moves the CEA by a forward or reverse stepping sequence. Control element drive mechanism hold is obtained by energizing one coil at a reduced current, while all other coils are deenergized. The CEAs are tripped upon interruption of electrical power to all coils. Each CEDM is connected to the CEAs by an extension shaft. The weight of the CEDMs and the CEAs is carried by the pressure vessel head. Installation, removal, and maintenance of the CEDM is possible with the reactor vessel head in place; however, the missile shield placed over the reactor vessel cavity makes the CEDMs inaccessible during operation of the plant. The axial position of a CEA in the core is indicated by three independent readout systems. One counts the CEDM steps electronically, and the other two consist of magnetically actuated reed switches located at regular intervals along the CEDM. These systems are designed to indicate CEA position to within 2-1/2 inches of the true location. This accuracy requirement is based on ensuring that the axial alignment between CEAs/PLCEAs is maintained within acceptable limits. The materials'in contact with the reactor coolant used in the CEDM are listed in Section 4.5.1. 3.9.4.1.1.1 CEDM Pressure Housing The CEDM pressure housing consists of the motor housing assembly and the upper pressure housing assembly. The motor housing assembly is attached to the reactor vessel head nozzle by means of a threaded joint and seal welded. Once the motor housing assembly is seal welded to the head nozzle, it need not be removed since all servicing of the CEDM is performed from the top of the housing. The upper pressure housing is threaded into the top of the motor housing assembly and seal welded. The upper pressure housing encloses the CEDM extension shaft and contains a vent. Attwoved Destyn Material Design of SSC Page 3.944
Sy-tem 80 + Design controlDocument p 3.9.4.1.1.2 Motor Assembly ( The motor asembly is an integral unit which fits into the motor housing and provides the linear motion t to the CEA. The motor assembly consists of a latch guide tube, upper latches and lower latches. Both upper latches and lower latches are used to perform the stepping of the CEA and by proper sequencing perform a load transfer and minimize latch and extension shaft wear. The upper latch also performs the holding when CEA motion is not required. Engagement of the extension shaft occurs when the appropriate set of magnetic coils is energized. This moves sliding magnets which cam a two-bar linkage moving the latches inward. The upper latches move vertically 7/16 inches while the lower latches move vertically 3/8 inches to perform both the load transfer and stepping action. Total CEA motion per cycle is 3/4 inches. 3.9.4.1.1.3 Coil Stack Assembly The coil stack assembly for the CEDM consists of four large DC magnet coils mounted on the outside of the motor housing assembly. The coils supply magnetic fome to actuate mechanical latches for engaging and driving the CEA extension shaft. Power for the magnetic coils is supplied from two separate supplies. A CEDM control system actuates the stepping cycle and obtains the correct CEA position by a forward or reverse stepping sequence. CEDM hold is obtained by energizing the upper latch coil at a reduced current while all other coils are deenergized. The CEAs are tripped upon interruption of electrical power to all coils. Electrical pulses from the magnetic coil power programmer provide one of the means for transmitting CEA position indication. ,!" A conduit assembly containing the lead wires for the coil stack assembly is located at the side of the ( upper pressure housing shroud. 3.9.4.1.1.4 Reed Switch Assembly Two reed switch assemblies provide separate means for transmitting CEA position indication. Reed switches and voltage divider networks are used to provide two independent output voltages proportional to the CEA position. The reed switch assemblies are positioned so as to utilize the permanent magnet i h the top of the extension shaft. The permanent magnet actuates the reed switches as it passed by them. The re:d switch assemblies are provided with accessible electrical connectors at the top of the upper pressure housing. 3.9.4.1.1.5 Extension Shaft Assembly The extension shaft assemblies are used to link the CEDMs to the CEAs. The extension shaft assembly is a 304 stainless steel rod with a permanent magnet assembly at the top for actuating reed switches in the reed switch assembly, a center section called the drive shaft and a lower end with a coupling device for connection to the CEA. The drive shaft is a long tube made of Type 304 stainless steel. It is threaded and pinned to the extension shaft. The drive shaft has circumferential notches in 3/4 inch increments along the shaft to provide the ; means of engagement to the control element drive mechanism. I 10 l l l l 4prowd onien nearenet. ow or ssc Page 3.945 l l
System 80+ Design ControlDocument The magnet assembly, located in the top of the extension shaft assembly, consists of a housing, magnet ' and plug. The magnet is made of two cylindaical Alnico-5 magnets. This magnet assembly is used to actuate the reed switch position indicator and is contained in a housing which is plugged at the bottom of the housing. 3.9.4.1.2 Descri:ot on of the CEDM Motor Operation Withdrawal or usertian af the CEA is accomplished by programming current to the various coils. There are three programmised r enditions for each coil (i.e., high voltage for initial gap closure, low voltage for maintaining the gap c!osed and zero voltage to allow opening of the gap). 3.9.4.1.2.1 Operating Sequence for the Double Stepping Mechanicm The initial condition is the hold mode. In this condition, the upper iatch coil is energized at low voltage.
- Withdrawal (Ref. Figure 3.9-8)
- 1. The upper lift coil is energized causing the 7/16" upper lift gap to close lifting the CEA.
- 2. Low current is supplied to hold the CEA in the withdrawn position.
- 3. The lower latch coil is energized causing the lower latches to engage the drive shaft with 1/32-inch clearance.
- 4. The upper lift coil is de-energized allowing the upper latches to drop 7/16 inches and the drive shaft to lower 1/32 inches placing the load on the lower latches.
- 5. The upper latch coil is de-energized disengaging the upper latches.
- 6. The lower lift coil is energized lifting the drive shaft 3/8 inches.
- 7. The upper latch coil is energized engaging the upper latches in the drive shafi with 1/32-inch clearance.
- 8. The lower lift coil is de-energized allowing the lower latches to drop 3/8 inches and causing the drive shaft to drop 1/32 inches applying the load on the upper latches.
- 9. The lower latch coil is de-energized disengaging the lower latches from the drive shaft.
- Insertion
- 1. The lower latch coil is energized causing the lower latches to engage the drive shaft.
- 2. The lower lift coil is energized lifting the lower latches 3/8 inches and lifting the drive shaft 1/32 inches thus applying the load to the lower latches.
- 3. The upper latch coil is de-energized causing the upper latches to disengage the drive shaft.
O Appresed Desberr Maternet - Design of SSC Page 3.9-46
System 80+ gion controlDocument
- 4. The upper lift coil is cnergized moving the de-energized upper latch assembly up 7/16 :
inches.
- 5. The upper latch coil is energized engaging the latches with 11/32 inch clearance.
- 6. The wer lift coil is de-energized allowing the lower latch to drop 3/8 inch. The drive ,
shaft will move down 11/32 inch, stopping on the upper latch assembly, which is energized and in its up position.
- 7. The lower latch coil is de-energized disengaging the lower latches.
~
- 8. The upper lift coil is de-energized lowering the upper latch assembly with the drive shaft 7/16 inch. ,
3.9.4.2 Applicable CEDM Design Specifications The pressure boundary components are constructed in accordance with the requirements for Class i vessels per the applicable Edition and Addenda of Section III (Subsections NCA and NB) of the ASME Boiler and Pressure Vessel Code and the criteria of SRP 3.9.4, Rev. 2 Subsection II.2. The pressure boundary material complies with the requirements of Section III and IX of the ASME Boiler and Pressure Vessel Code and Code Case N-4-11. The adequacy of the design of the non-pressure boundary components has been verified by prototype accelerated life testing as discussed in Section 3.9.4.4. The reed switch position transmitter (RSPT) assembly of the CEDM is designed to comply with IEEE Standard 323-1974, standard for " Qualification of Class I Electrical Equipment for Nuclear Power Generating Stations," and IEEE Standard 344-1987, " Recommended Practice Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations", as endorsed by Regulatory Guide 1.100 Rev. 2, dated June 1988. The electrical components are external to the pressure boundary and are non-pressurized. 4 The test program to verify the CEDM design is discussed in Section 3.9.4.4. 3.9.4.3 Design Loads, Stress Limits and Allowable Deformations The CEDM stress analyses consider the following loads: l 1
- Reactor coolant pressure and temperature j
- Reactor operating transient conditions
- Dynamic stresses produced by seismic loading and design bases pipe breaks and/or LOCA loading.
- Dynamic stresses produced by mechanical excitations
. Fulllength RSPT assemblies are subjected to biaxial random multi-frequency input motions corresponding
[V]- to' design bases excitations. Testing is performed using four RSPT orientations to account for asymmetries in the design. 1 w= one nesww.oup or ssc reo. 3. w
System 80+ Design Control Document
- Loads produced by the operation and tripping of the mechanism e Dynamic stresses produced by excitations from pipe breaks other than those eliminated by LBB.
O The methods used to demonstrate that the CEDMs operate properly under seismic conditions are presented in Section 3.7.3.14. The design and fabrication of the CEDM pressure boundary components fulfills the requirements of the ASME Code, Section III, for Class I vessels. The pressure housings are capable of withstanding throughout the design life all the steady state and transient operating conditions specified in Table 3.9-14. The adequacy of the design of the CEDM pressure boundary and non-pressure boundary components has been verified by prototype accelerated life testing as discussed in Section 3.9.4.4. Clearances for thermal growth and for dimensional tolerances were investigated, and tests have proven that adequate clearances are provided for proper operation of the CEDM. The latch locations are set by a master gauge, and settings are verified by testing at reactor conditions. A weldable seal closure, per Section III of the ASME Code, is provided for the vent valve in case of leakage. The motor housing fasteners are mechanically positively captured, and all threaded connections are preloaded before capturing. The coil stack assembly can be installed or removed simply by lowering or lifting the stack, relative to the CEDM pressure housing, for case of coil replacement or maintenance. 3.9.4.4 CEDM Performance Assurance Program 3.9.4.4.1 CEDM Testing 3.9.4.4.1.1 Prototype Accelerated Life Tests The System 80+ CEDM is similar to and based on existing magnetic jack mechanisms presently in use on operating reactors such as Maine Yankee (Docket No. 50-309) and Calvert Cliffs (Docket Nos. 50-317 and 318), the 150-inch core reactors such as Arkansas Nuclear One Unit 2 (Docket No. 50-368) and San Onofre Units 2 & 3 (Docket Nos. 50-361 and 362), and is the same as the System 80 CEDM presently in use at Palo Verde (Docket Nos. 50-528,529 and 530). The significant cifferences between the System 80+ drives and pre-System 80 CEDMs are:
- The elimination of the pulldown coil.
- The use of the lift coils to perform both a load transfer function and stepping action.
The elimination of the pulldown coil required installation of a coil spring to ensure positive resetting of the latch assemblies. In addition, the drive shaft was modified by placing the teeth on 3/4-inch pitch in place of the 3/8-inch spacing of previous drive shafts to allow load transfer and stepping with the same coil. The safety release mechanism uses the same materials and clearances as on all previous magnetic Approved Design Material . Design of SSC Page 3.9-48
1 System 80+ oe-ion controlDocument ! jack mechanisms. The following describes accelerated life tests on both a pre-System 80 mechanism as I ' O well as on a prctotype System 80 CEDM. Both programs provide design verification for the System 80+ CEDM. A pre-System 80 prototype CEDM was subjected to an accelerated life test accumulatmg a muumum of ] 157,000 feet of travel on all CEDM components. In addition, the latch guide tube bearings in the motor l _ assembly saw an additional 50,000 feet of operation. The prototype mechanism was installed on a test facility which was operated at a nominal temperature l i of 600'F and 2250 psi. After 50,000 feet of operation lifting 230 pounds at 40 inches per minute, the ! l motor was removed from the test motor housing and the bearing surfaces inspected. During this i inspection it was found that excessive wear existed on the upper gripper magnet and upper gripper housing bearings. ; i , i The gripper housing magnet bearing configuration was revised and replacement parts with this revision were incorporated into the prototype mechanism. This configuration was reinstalled into the test facility and the mechanism operated as before for an additional 157,000 feet of travel. The replacement parts
. showed a wear of only .001 inches while the latch guide tube bearings had a total wear of 0.012 inches.
The mechanism at disassembly was still operational with no abnormalities. This test constituted operation equivalent to 1.5 to 2.0 times the design duty requirements of the mechanism. A prototype System 80 CEDM was assembled and installed in a test loop, where the accelerated wear test was conducted at 615'F and 2250 psi. The total weight attached to the CEDM was 450 pounds and this was moved at a nominal speed of 30 inches per minute. A total of 34,000 feet of travel was then completed without difficulty. Included in that test footage were 300 full-height gravity scrams. The mechanism motor was removed from the test facility and disassembled for inspection. The latch l
- guide tube bearings showed a maximum diametral wear of 0.003 inches with negligible wear on the gripper housing to gripper magnetic bearings. Alignment tabs, which maintain orientation of the gripper with the latch guide tube, showed extensive wear but had not caused mechanism malfunctions. These alignment tabs have been replaced in the production units with an improved design.
Upon completion of the accelerated wear test, 300 full height light weight drops were completed utilizing a 75-pound test weight. The maximum CEA drop time to 90% insertion was 2.93 seconds which met the 4.0 second criterion. All release times were less than the 0.3 seconds with normal releases completed in less than 0.200 seconds. 3.9.4.4.1.2 First Production Test A qualification test program was completed on the first production C-E magneticjack CEDM. A similar
- test program was invoked for the System 80 CEDMs. During the course of this program, over 4000 feet of travel was accumulated and 30 full height gravity drops were made without mechanism malfunction or measurable wear on operating parts. The program included the following
e Operation at 40 inJmin lifting 230 pounds (dry) at ambient temperature and 2300 psig pressure for 800 feet.
.e Six full height 230 pounds dry weight gravity drops at ambient temperature.
-e~ Operation at simulated reactor operating condition at 40 in/ min lifting 230-pound for 1700 feet,
=.= onnen unaw . oeeon a ssc rene 3.s-ts 4' 1
- - -a -
System 80 + Design ControlDocument
- Six full-height drops at simulated reactor operating conditions with 230 pounds of weight.
- An operational test at ambient temperature and 2300 psig pressure, lifting 335 pounds for 500 O
feet.
- Six full-height drops of the 335 pound weight.
- Operation at simulated reactor conditions for 1700 feet at 20 in/ min, lifting 335 pounds.
- Operation at ambient temperature and 2300 psig for 1100 feet and 20 full-height drops with an attached dry weight of 130 pounds.
The mechanism operated without malfunction throughout the test program and, upon final inspection, no measurable wear was found. 3.9.4.4.1.3 Operating Experience at the Palo Verde Nuclear Generating Station The System 80+ CEDMs are identical to those in operation at PVNGS. That experience has shown that the CEDMS operate without malfunction. 3.9.5 Reactor Vessel Core Support and Internals Structures 3.9.5.1 Design Arrangements 9e components of the reactor vessel core support structures are divided into two major parts consisting of the core support structure and the upper guide structure assembly. The flow skirt, although functioning as an integral part of the coolant flow path, is separate from the internals and is affixed to the bottom head of the pressure vessel. The arrangement of these components is shown in Figure 3.9-9. 3.9.5.1.1 Core Support Structure The major structural member of the reactor internals is the core support structure. The core support structure consists of the core suppon barrel and the lower support structure. The material for the assembly is Type 304 stainless steel. The core suppon structure is supponed at its upper end by the upper flange of the core support barrel, which rests on a ledge in the reactor vessel. Alignment is accomplished by means of four equally spaced keys in the flange, which fit into the keys in the vessel ledge and closure head. The lower flange of the core support barrel suppons, secures, and positions the lower suppon structure and is attached to the ' lower support structure by means of a welded flexural connection. The lower support structure provides support for the core by means of support beams that transmit the load to the core support barrel lower flange. The locating pins in the beams provide orientation for the lower ends of the fuel assemblies. The core shroud, which provides a flow path for the coolant and lateral support for the fuel assemblies, is also supponed and positioned by the lower support structure. The lower end of the core suppon barrel is l restricted from excessive radial and torsional movement by six snubbers which interface with the pressure l vessel wall. 0 AMvowd Design Atatorial Design of SSC Page 3.9-50
Srtem 80+ Design ControlDocument n 3.9.5.1.1.1 Core Support Barrel
- 1 The core support barrel is a right circular cylinder including a heavy external ring flange at the top end and an internal ring flange at the lower end. The core support barrel is supported from a ledge on the pressure vessel. The core support barrel, in turn, supports the lower support structure upon which the fuel assemblies rest. Press-fitted into the flange of the core support barrel are four alignment keys located 90 degrees apart. The reactor vessel, closure head, and upper guide structure assembly flange are slotted in locations corresponding to the alignment key locations to provide alignment between these components in the vessel flange region. The core support barrel assembly is shown in Figure 3.9-10.
The upper section of the barrel contains two outlet nozzles that interface with internal projections on the vessel nozzles to minimize leakage of coolant from inlet to outlet. Since the weight of the core support barrel is supported at its upper end, it is possible that coolant flow could induce vibrations in the structure. Therefore, amplitude limiting devices, or snubbers, are installed on the outside of the core support barrel near the bottom end. The snubbers consist of six equally-spaced lugs around the circumference of the barrel and act as a tongue-and-groove assembly with the mating lugs on the pressure vessel. Minimizing the clearance between the two mating pieces limits the amplitude of vibration. During assembly, as the internals are lowered into the pressure vessel, the pressure vessel lugs engage the core suppoit barrel lugs in an axial direction. Radial and axial expansion of the core support barrel are accommodated, but lateral movement of the core support barrel is restricted. The pressure vessel lugs have bolted, captured Inconel X-750, SA-637, Grade 688, Type 2 shims. The core support barrel lug mating surfaces are hardfaced with Stellite to minimize wear. The shims are machined during initial installation to provide minimum clearance. The snubber assembly is shown in Figure 3.9-11. 3.9.5.1.1.2 Lower Support Structure and Instrument Nozzle Assembly p' V The lower support structure and ICI nozzle assembly position and support the fuel assemblies, core shroud, and ICI nozzles. The structure is a welded assembly consisting of a short cylinder, support beams, a bottom plate, ICI nozzles, and ICI nozzle support plate. The lowest support structure is made up of a short cylindrical section enclosing an assemblage of grid beams arranged in egg-crate fashion. The outer ends of these beams are welded to the cylinder. Fuel assembly locating pins are attached to the beams. The bottoms of the parallel beams in one direction are welded to an array of plates which contain flow holes to provide proper flow distribution. These plates also provide support for the ICI nozzles and, through support columns, the ICI nozzle support plate. The cylinder guides the main coolant flow and limits the core shroud bypass flow by means of holes located near the base of the cylinder. The ICI nozzle support plate provides lateral support for the nozzles. This plate is provided with flow holes for the requisite flow distribution. The lower support structure and ICI nozzle assembly is shown in Figure 3.9-12. 3.9.5.1.1.3 Core Shroud The core shroud provides an envelope for the core and limits the amount of coolant bypass flow. The shroud consists of a welded vertical assembly of plates designed to channel the coolant through the core. Circumferential rings and a top and bottom end plate provide lateral support. The rings are attached to the vertical plates by means of welded ribs which extend the full length of the core shroud. A small gap is provided between the core shroud outer perimeter and the core support barrel in order to provide upward coolant flow in the annulus, thereby minimizing thermal stresses in the core shroud. The core shroud is shown in Figure 3.9-13. Four hardfaced alignment lugs, spaced 90 degrees apart, protrude (%r) Approved Design Atatorial Design of SSC Page .T.9-61
System 80+ Design ControlDocument vertically from the top of the core shroud and engage in corresponding hardfaced slots in the upper guide structure fuel alignment plate to ensure proper alignment between the upper guide structure assembly, ! core shroud, and lower support structure. 3.9.5.1.2 Upper Guide Structure Assembly The Upper Guide Structure Assembly (UGS) aligns and laterally supports the upper end of the fuel assemblies, maintains the control element spacing, holds down the fuel assemblies during operation, prevents fuel assemblies from being lifted out of position during a severe accident condition and protects the control elements from the effects of coolant cross flow in the upper plenum. The UGS assembly is handled as one tmit during installation and refueling. The UGS assembly consists of the UGS support barrel assembly and the CEA shroud assembly (Figure 3.9-14). The UGS support barrel assembly consists of UGS support barrel fuel alignment plate, UGS base plate and control element shroud tubes. The UGS support barrel consists of a right circular cylinder welded to a ring flange at the upper end and to a circular plate (UGS base plate) at the lower end. The flange, which is the supporting member for the entire UGS assembly, seats on its upper side against the pressure vessel head during operation. The lower side of the flange is supported by the holddown ring, which seats on the core support barrel upper flange. The UGS flange and the holddown ring engage the core support barrel alignment keys by means of four accurately machined and located keyways equally spaced at 90 degree intervals. This system of keys and slots provides an accurate means of aligning the core with the closure head and thereby with the CEA drive mechanisms. The fuel alignment plate is positioned below the UGS base plate by cylindrical control element shroud tubes. These tubes are attached to the UGS base plate and the fuel alignment plate by rolling the tubes into the plates and welding. The fuel alignment plate is designed to align the lower ends of the control element shroud tubes which in turn locate the upper ends of the fuel assemblies. The fuel alignment plate also has four equally spaced slots on its outer edge which engage with Stellite hardfaced lugs protruding from the core shroud to provide alignment. The control element shroud tubes bear the upward force on the fuel assembly holddown devices. This force is transmitted from the alignment plate through the control element shroud tubes to the UGS barrel base plate. The CEA shroud assembly limits cross flow and provides separation of the CEA assemblies. The assembly consists of an assemblage of large vertical tubes connected by vertical plates in a grid pattern. The tubes and connecting plates are furnished with multiple holes to permit hydraulic communication. The grid assemblage is attached to a cylinder at the periphery which in turn is supported by a flange at the UGS flange evaluation. Guides for the CEA extension shafts are provided by the guide structure support system (GSSS). The holddown ring provides axial force on the flanges of the upper guide structure assembly and the core support structure in order to prevent movement of the structures under hydraulic forces. The holddown ring is designed to accommodate the differential thermal expansion between the pressure vessel and the internals in the vessel ledge region. 3.9.5.1.3 Flow Skirt The inconel flow skirt is a right circular cylinder, perforated with flow holes, and reinforced with two stiffening rings. The flow skirt is used to reduce inequalities in core inlet flow distributions and to prevent formation of large vortices in the lower plenum. The skirt is supported by nine equally spaced machined sections that are welded to the bottom head of the pressure vessel. Atywowd Design Material Design of SSC Page 3.9-52
Srtem 80+ Design ControlDocument 3.9.5.1.4 In-Core Instrumentation Support System The in-core neutron flux monitoring system includes self-powered in-core detector assemblies, supporting i structures and guide paths and an amplifier system to process detector signals. The self-powered in-core detector assemblies and the amplifier system are described in Section 7.7. The instrumentation supporting structures and guide paths are described in this section and shown in Figure 3.9-15. The support system begins outside the pressure vessel, penetrates the bottom of the vessel boundary and terminates in the upper end of the fuel assembly. Each in-core instrument is guided over its full length by the external guidance conduit, the pressure vessel nozzles, the lower support structure ICI nozzles and the instrument guide tube of the fuel assembly. Figure 3.9-12 shows the in-core instrument support structure. The in-core instrumentation support system routes the instruments so that detectors are located in selected fuel assemblies throughout the core. An equal instrument length exists for all locations. The guide tube routing outside the reactor vessel is a simple 180' bend to the seal table. The pressure boundaries for the individual instruments are at the out-of-reactor seal table, where the external electrical connections to the in-core instruments are made (Figure 3.9-15). Each instrument has an integral seal plug which forms a seal at the instrument seal table and through which the signal cables pass. Static O-ring seals are used to seal against operating pressure.- 3.9.5.2 Design Loading Conditions ; The following loading conditions are considered in the design of the core support and internals structures. ,
- Normal operating temperature differences
'b V
- Normal operating pressure differences
- Flow loads
- Weights, reactions and superimposed loads e Vibration loads
- Shock loads
- Anticipated transient loadings in accordance with Table 3.9-1. !
- Handling loads (not combined with other loads above) i
- Appropriate DBPB, secondary side break and LOCA loads.
P 3.9.5.3 Design Loading Categories
. The design loading conditions are categorized as follows:
3.9.5.3.1 level A and Level B Ser3 ice Loadings
- This category includes the combinations of design loadings consisting of normal operating temperature and pressure differences, loads due to flow, weights, reactions, superimposed loads, vibration, shock l' \ loads, and transient loads not requiring shutdown.
oprme seemo unnerw owon or ssc rage 2.s-52 i a -~ ~ - - - , . ,
Sy stem 80 + Design ControlDocument 3.9.5.3.2 Level C Service Loadings Level C Service Loadings are derived from a loading combination of normal operating loads and the design basis pipe break (DBPB). The DBPB for reactor vessel core support and internal structures is defined as a postulated pipe break that results in the loss of reactor coolant at a rate less than or equal to the capability of the reactor coolant makeup system (i.e. less than 150 GPM). 3.9.5.3.3 Level D Service Loadings The following loading combination is considered as Level D Service Loadings. e Normal Operation Loads
- Either the Feed Water Pipe Break (FWPB) or Loss of Coolant Accident (LOCA) Loads
- Safe Shutdown Earthquake (SSE) Loads LOCA is defined as the loss of reactor coolant at a rate in excess of the reactor coolant normal makeup rate, from breaks in the reactor coolant pressure boundary inside primary containment up to, and including, a break equivalent in size to the largest primary branch line not eliminated by leak before break (LBB) criteria.
3.9.5.4 Design Bases for Reactor Internals The stress limits to which the reactor internals are designed are listed in Table 3.9-14. The operating categories and stress limits are defined in the applicable section of the Section III of the ASME Boiler and Pressure Vessel Code. To properly perform their functions, the reactor internal structures are designed to meet the deformrion limits listed below:
- Under Level A Level B and Level C service loadings, the core will be held in place and deflections will be limited so that the CEAs can be inserted under their own weight as the only driving force.
- Under service loading combinations other than Level A, B, and C service loadings that require CEA insertability, deflections are limited so that the core will be held in place, adequate core cooling is preserved, and all CEAs can be inserted. Those deflections that would influence CEA movement are limited to less than 80% of the deflections required to prevent CEA insertion.
The allowable deformation limits are established as 80% of the loss-of-function deflection limits. The significant component deflection limits are designed as follows:
- 1. Fuel lower end fitting interface with the lower support structure is deflection limited to avoid disengagement.
Apnproved Design Maternal- Design of SSC Page 3.9-54
Srtem 80+ Design controlDocument n- 2. Fuel upper end fitting interface with the upper guide structure relative displacement precludes disengagement. (]
- 3. The CEA shroud lateral deflection allows CEA insertion. l l
In the design of critical reactor vessel internals components which are subject to fatigue, the stress j analysis is performed utilizing the design fatigue curve of Figure I-9-2 of Section III of the ASME Boiler and Pressure Vessel Code. A cumulative usage factor of less than one is used as the limiting criterion. As indicated in the preceding sections, the stress and fatigue limits for reactor internals components are obtained from the ASME Code. Allowable deformation limits are established as 80% of the loss-of-function deflection limits. These liinits provide adequate safety factors assuring that so long as calculated , I stresses, usage factors, or deformations do not exceed these limits, the design is conservative. 3.9.6 Testing of Pumps and Valves Qualification, preoperational, and inservice testing (IST) of safety-related pumps and valves will be addressed in plant-specific information. See also Section 3.9. Inservice testing for safety-related pumps and valves is developed in accordance with the requirements of ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987, Parts 1,6 and 10. Table 3.9-15 lists the inservice testing parameters and frequencies for the safety-related pumps and valves. Safety-related . pumps and valves include those necessary to ensure:
- The integrity of the reactor coolant pressure boundary.
- The capability to achieve safe shutdown of the reactor and keep it in a safe shutdown condition.
- The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures in excess of 10 CFR 100 guidelines.
Table 3.9-15 also provides explanatory notes / justifications for any code defined testing exceptions. Plant-specific information will include safety-related pump and valve inservice testing details, including test schedules and frequencies, in the inspection and testing program. This program will include baseline pre-service testing to support the periodic inservice testing of the safety-related pumps and valves. Depending on the test results, the plan will provide commitment to disassemble and inspect the safety- . related pumps and valves, as described in the following paragraphs. The primary elements of this plan, including the requirements of Generic Letter 89-10 for motor operated valves, are promulgated in the subsections of Section 3.9.6. Inservice Inspection (ISI) is discussed in Sections 5.2.4 and 6.6. 3.9.6.1 Testing of Safety-Related Pumps . For each safety-related pump, the design basis and required operating conditions (including tests) under which the pump will be required to function will be established. These design (design basis and required operating) conditions include flow rate and corresponding pump head for each system mode of pump (,) v operation and the required operating time for each mode, acceptable bearing vibration levels, seismic / dynamic loads, fluid temperature, ambient temperature, and pump motor minimum voltage. Anmed oeson anesweet oessen or ssc rope 3.s-ss
Syotem 80+ Design ControlDocument Plant-specific information will include the following design and qualification requirements and acceptance criteria for these requirements. For each size, type, and model testing which encompasses design conditions will be performed to demonstrate acceptable flow rate and corresponding pump head, bearing vibration levels, and pump internals wear rates for the operating time specified for each system mode of pump operation. From these tests baseline (reference) hydraulic and vibration data for evaluating the acceptability of the pump after installation will also be developed. Test data will be used to ensure that the pump specified for each application is not susceptible to inadequate minimum flow rate and inadequate thrust bearing capacity. With respect to minimum pump flow operation, the sizing of each minimum recirculation flow path is evaluated to assure that its use under all analyzed conditions will not result in degradation of the pump. The flow rate through minimum recirculation flow paths can also be periodically measured to verify that flow is in accordance with the design specification. The safety-related pumps and piping configurations accommodate inservice testing at a flow rate at least as large as the maximum design flow for the pump application. The safety-related pumps are provided with instrumentation to verify that the net positive suction head (NPSH) is greater than or qual to the NPSH required during all modes of pump operation. These pumps can be disassembled for evaluation when Part 6 testing results in a deviation which falls within the " required action range." The Code provides criteria limits for the test parameters identified in Table 3.9-15. The detailed IST program will establish the frequency and the extent of disassembly and inspection based on suspected degradation of all safety-related pumps, including the basis for the frequency and the extent of each disassembly. Factors to be considered in the disassembly frequency and extent of disassembly include, but are not limited to: e Historical performance of the pump to identify pumps which are prone to degradation / wear.
- Analysis of trends of pump test parameters and service conditions.
- Analysis of pump components which are subject to aging and require a maintenance replacement approach (e.g., "O-Rings").
- Results of non-intrusive pump testing. The non-intrusive technologies employed may obviate the need for inspection / disassembly of safety-related pumps altogether, provided the technologies demonstrate an equivalent ability to detect pump degradation as inspection / disassembly would.
The program may be revised t.kughout the plant life to minimize disassembly based upon past disassembly experience. if OM-6 pump tests cannot be performed on the CCW or SSW pumps due to inability to repeat pump test single point flow conditions, pump curve testing will be used to assess pump degradation. The following provisions shall be complied with in the use of pump curve testing for the CCW/SSW pumps:
- Pump curves are developed, or manufacturer's pump curves are validated, when the pumps are known to be operating acceptably.
- The reference points used to develop or validate the curve are measured using instruments at least as accurate as required by the Code.
- Pump curves are based on an adequate number of points, with a minimum of five.
O Approved Design neatonal Design of SSC Page 3.9-56
System 80+ oesign controlDocument g) ( V Points are beyond the " flat" portion (low flow rates) of the curves in a range which includes or is as close as practicable to design basis flow rates.
- Acceptance criteria based on the curves does not conflict with Technical Specifications or CESSAR-DC design bases.
- If vibration levels vary significantly over the range of pump conditions, a method for assigning appropriate vibration acceptance criteria should be developed for regions of the pumps curve.
- When the reference pump curve may have been affected by repair, replacement, or routine service, a new reference curve shall be determined or the previous curve revalidated by an inservice test.
3.9.6.2 Testing of Safety-Related Valves 3.9.6.2.1 Motor-Operated Valves For each motor-operated valve assembly (MOV) with an active safety-related function, the design basis and required operating conditions (including testing) under which the MOV will be required to perform are established for the development and implementation of the design, qualification, and preoperational testing. 3.9.6.2.1.1 Design and Qualification Requirements for Motor-Operated Valves q Plant-specific information will include the following design and qualification requirements and acceptance criteria for these requirements. By testing each size, type, and model the torque and thrust (as applicable to the type of MOV) requirements to operate the MOV will be determined to ensure the adequacy of the torque and thrust that the motor-operator can deliver under design (design basis and required operating) conditions, ((Each size, type, and model will be tested under a range ofdiferential pressure and pow conditions up to the design conditions. These design conditionsincludefluidflow, diferentialpressure (including pipe break), system pressure, fluid temperature, ambient temperature, minimum voltage, and minimum and maximum stroke time requirements. nis testing of each size, type and model shallinclude test datafrom the manufacturer, peld test dataforplant-specipc dedication, empirical data supported by test, or test (such as aprototype) ofsimilar valves which support qualtpcation of the required valve where similarity must be justrped by technical data.))2 This preoperational testing will demonstrate that the results of the testing under in-situ or installed conditions can be used to ensure the capability of the MOV to operate under design conditions. Test data will be used to ensure that ((the structural capability limits ofthe individualparts of the MOV will not be exceeded under design conditions.))2 Additionalguidelines to justify prototype qualification testing are contained in Generic Letter 89-10, Supplement 1, Questions 22, 24,25, 26,27, and 28. Test data will be used to ensure that the valve specified for each application is not susceptible to pressure locking and thermal binding.
\\Re concerns and issues identiped in Generic Lener 89-10 and its supplements for MOVs will be addressedprior to plant startup, and the structural capability limits of the individualparts of the MOV will not be exceeded under design conditions.))2 The following testing requirements and acceptance 2
NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction
- Section 3.5.
Anwomt Design Moseniah Design of SSC 111/96) Page 3.9-57
i System 80+ Design ControlDocument criteria are applicable to each motor-operated valve assembly (MOV with an active safety-related function). 3.9.6.2.1.2 Pre-Operational Testing of Safety-Related Motor-Operated Valves (leach MOV will be tested in the open and closed directions under static and maximum achievable preoperational conditions using diagnostic equipment which measures torque and thrust (as applicable to the type ofMOV), and motorparameters. The MOV will be tested under various diferentialpressure andflow conditions up to maximum achievable conditions and perform a suficient number of tests to determine the torque and thrust requirements at design conditions.)) The torque and thrust requirements to close the valve to the position at which there is diagnostic indication of hard seat contact will be determined. llThe determination of design torque and thrust requirements will be made for such parameters as diferentialpressure, fluidflow, undervoltage, temperature and seismic dynamic effectsfor MOVs which must operate during these transients. The design torque and thrust requirements will be adjustedfor diagnostic equipment inaccuracies.))2 For the point of control switch trip, any loss in torque produced by the actuator and thrust delivered to the stem for increasing differential pressure and flow conditions (referred to as load sensitive behavior) will be determined. The design torque and thrust requirements will be compared to the control switch trip torque and thrust by subtracting margin for load sensitive behavior, control switch repeatability, and degradation. ((7he total thrust and torque delivered by the MOV under static and dynamic conditions (including diagnostic equipment inaccuracies and control switch repeatability) will be measured and compared to the allowable structural capability limitsfor the individualparts of the MOV.))2 Tests will be conducted for proper control room position indication of the MOV. The parameters and acceptance criteria for demonstrating that the above functional performance requirements have been fulfilled are as follows:
- As required by the safety function, the valve must fully open, or the valve must fully close with diagnostic indication of hard seat contact.
- The control switch settings must provide adequate margin to achieve design requirements including consideration of diagnostic equipment inaccuracies, control switch repeatability, load sensitive behavior, and margin for degradation.
- The motor output capability at degraded voltage must equal or exceed the control switch setting including consideration of diagnostic equipment inaccuracies, control switch repeatability, load sensitive behavior and margin for degradation.
- The maximum torque and thrust (as applicable for the type of MOV) achieved by the MOV including diagnostic equipment inaccuracies and control switch repeatability must not exceed the allowable stmetural capability limits for the individual parts of the MOV.
- The remote position indication testing must verify that proper disk position is indicated in the control room.
2 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approved Dessen Matenet - Design of $$C Page 3.9-58
System 80+ De-ign ControlDocument
,m - o Stroke time measurements taken during valve opening and closing must meet minimum and maximum stroke time requirements.
3.9.6.2.1.3 Inservice Testing of Safety-Related Motor-Operated Valves The inservice testing of MOVs will rely on diagnostic techniques which are consistent with the state of the art and which will permit an assessment of the performance of the valve under actual loading. Periodic testing per Generic Letter 89-10 Paragraphs D and J will be conducted under adequate differential pressure and flow conditions which allow a justifiable demonstration of continuing MOV , capability for design basis conditions. The detailed IST program will include the optimal frequency of this periodic verification. The frequency and test conditions will be sufficient to demonstrate continuing design basis and required operating capability. The Code provides criteria limits for the test parameters identified in Table 3.9-15 for Code inservice testing. The detailed IST program will establish the frequency and the extent of disassembly and inspection based on suspected degradation of all safety-related MOVs, including the basis for the frequency and the extent of each disassembly. Factors to be considered in the disassembly frequency and extent of disassembly include, but are not limited to:
- Historical performance of the safety-related valves to identify valves which are prone to degradation / wear.
- Analysis of trends of valve test parameters and service condition.
/'
- Analysis of valve components which are subject to aging and require a maintenance replacement
( approach (e.g., "O-Rings").
- Results of non-intrusive valve testing. Use of non-intrusive technologies may obviate the need for inspection / disassembly of safety-related valves altogether, provided the technologies demonstrate an equivalent ability to detect valve degradation as inspection / disassembly would.
The program may be revised throughout plant life to minimize disassembly based on past disassembly experience. 3.9.6.2.2 Power-Operated Valves For each power-operated (includes pneumatic , hydraulic , piston , and solenoid-operated) valve assembly (POV) with an active safety-related function, the design basis and required operating conditions (including testing) under which the POV will be required to perform will be established. 3.9.6.2.2.1 Design and Qualification Requirements for Power-Operated Valves Plant-specific information will include the following design and qualification requirements and acceptance criteria for these requirements. By testing each size, type, and model the force (as applicable to the type of POV) requirements to operate the POV will be determined to ensure the adequacy of the force that the operator can deliver under design (design basis and required operating) conditions. Each size, type, and model will be tested under a range of differential pressure and flow conditions up to the design conditions. These design conditions include fluid flow, differential pressure (including pipe break), ; I
- system pressure, fluid temperature, ambient temperature, minimum air supply system (or accumulator) k-) pressure, spring force, and minimum and maximum stroke time requirements. This testing of each si7e, L 2 Des &n Acetodel- Design of SSC Pope 3.9-59
1 System 80+ Design ControlDocument ! l type and model shall include test data from the r.unufacturer, field test data for plant-specific dedication, l empirical data supported by test, or test (such as a prototype) of similar valves which support qualification l of the required valve where similarity must bejustified by technical data. This preoperational testing will demonstrate that the results of the testing under in-situ or installed conditions can be used to ensure the capability of the POV to operate under design conditions. Test data will be used to ensure that the r,tructural capability limits of the individual parts of the POV will not be exceeded under design conditions and that packing adjustment limits are specified for the valve for each application such that it is not susceptible to stem binding. 3.9.6.2.2.2 Pre-Operational Testing of Safety-Related Power-Operated Valves Each POV will be tested in the open and closed directions under static and maximum achievable preoperational conditions using diagnostic equipment which measures or provides information to determine total friction, stroke time, seat load, spring rate, travel under normal pneumatic or hydraulic pressure (as applicable to the type of POV), and minimum pneumatic or hydraulic pressure. The POV will be tested under various differential pressure and flow conditions up to maximum achievable conditions and perform a sufficient number of tests to determine the force requirements at design conditions. The force requirements to close the valve to the position at which there is diagnostic indication of full valve closure (as required for the safety function of the applicable valves) will be determined. The determination of design force requirements will be made for such parameters as differential pressure, fluid flow, minimum pneumatic or hydraulic pressure, power supply, temperature, and seismic / dynamic effects for POVs which must operate during these transients. The design force requirements will be adjusted for diagnostic equipment inaccuracies. Total force delivered by the POV under static and dynamic conditions (including diagnostic equipment inaccuracies) will be measured to compare to the allowable structural capability limits for the assembly and individual parts of the POV. Tests will be conducted for proper control room position indication of the POV. The parameters and acceptance criteria for demonstrating that the above functional performance requirements have been fulfilled are as follows:
- As required by the safety function, the valve must fully open, or the valve must fully close with diagnostic indication of hard seat contact.
- The assembly must demonstrate adequate margin to achieve design requirements including consideration of diagnostic equipment inaccuracies and margin for degradation.
- The assembly must demonstrate adequate output capability of the power-operator at minimum pneumatic or hydraulic pressure or electrical supply (or loss of motive force for fail-safe positioning) with consideration of diagnostic equipment inaccuracies and margin for degradation.
- The maximum force (as applicable for the type of POV) achieved by the POV including diagnostic equipment inaccuracies must not exceed the allowable structural capability limits for the assembly and individual parts of the POV.
- The remote position indication testing must verify that proper disk position is indicated in the control room and other remote locations relied upon by operators in any emergency situation.
O Apprend Desmer historial Design of SSC Page 3.9-60
__ _ _ . . _ _ _ . . ~ _ _ . . - , . . _ _ ._ . . . _ __ . _ _ _ _ . _ i i
' Sv: tem 80+ oestan controlooeument ;
m o Stroke time measurements taken during valve opening and closing must meet minimum and ! maximum stroke time requirements, e For solenoid-operated valves (SOVs), the Class IE electrical requirements are to be verified. The ! SOV should be verified to be capable of performing for design requirements for energized or l
'deenergized and rate appropriately for the electrical power supply amperage and voltage. !
e .. Provide leak-tight seating which must meet a specified maximum leakage rate, or meet a leakage l rate to ensure an overall containment maximum leakage. l 3.9.6.2.2.3- Inservice Testing of Safety-Related Power-Operated Valves l All safety-related piping systems incorporate provisions for testing to demonstrate the operability of the POVs under design conditions. The inservice testing of POVs will incorporate the use of advanced non- , intrusive techniques to periodically assess degradation and the performance characteristics of the POVs. l The Part 10 tests will be performed, and valves which fail to exhibit the required performance can be i disassembled for evaluation. The Code provides criteria limits for the test parameters identified in Table
; 3.9-15. {
The detailed IST program will establish the frequency and the extent of disassembly and inspection based j on suspected degradation of all safety-related POVs, including the basis for the frequency and the extent j l of each disassembly. Factors to be considered in the disassembly frequency and extent of disassembly l
- include, but are not lirnited to
- !
i e historical performance of the safety-related valves to identify valves which are prone to l
- c degradation / wear, l
)
e analysis of trends of valve test parameters and service conditions, e analysis of valve comronents which are subject to aging and require a maintenance replacement
- approach (e.g., "O-Rings"), and
- o results of non-intrusive valve testing. Use of non-intrusive technologies may obviate the need 4 for inspection / disassembly of safety-related valves altogether, provided the technologies ;
-demonstrate an equivalent ability to detect valve degradation as inspection / disassembly would.
( The program may be revised throughout plant life to minimize disassembly based on past disassembly [ experience. { 4 3.9.6.2.3 Check Valves j i
' For each check valve with an active safety related function, the design basis and required operating -?
' L conditions (including testing) under which the check valve will be required to perform will be established. l l, ? '3.9.6.2.3.1 Design and' Qualification Rem _' M for Check Valves ! t
' Plant-specific information will include the following design and qualification requirements and acceptance criteria for these requirements. By testing each size, type, and model the design adequacy of the check valve under design (design basis and required operating) conditions will be ensured. These design
- conditions include all the required system operating cycles to be experienced by the valve (numbers of l i
- r 4pmed Deeon noneenw- Denon er ssc rene 2.s sr l
System 80+ Design ControlDocument each type of cycle and duration of each type cycle), environmental conditions under which the valve will be required to function, severe transient loadings expected during the life of the valve such as waterhemner or pipe break, lifetime expectation between major refurbishments, sealing and leakage requirements, corrosion requirements, operating medium with flow and velocity definition, operating medium temperature and gradients, maintenance requirements, vibratory loading, planned testing and methods, test frequency and periods of idle operation. The design conditions may include other requirements as identified during detailed design of the plant systems. This testing of each size, type and model shall include test data from the manufacturer, field test data for plant-specific dedication, empirical data supported by test, or test (such as prototype) of similar valves which support qualification of the required valve where similarity must be justified by technical data. Test data will be used to ensure proper check valve application, including selection of the valve size and type based on the system flow conditions, installed location of the valve with respect to sources of turbulence, and correct orientation of the valve in the piping (i.e., vertical vs horizontal) as recommended by the manufacturer. Valve design features, material, and surface finish will ensure that n)n-intrusive diagnostic testing methods available in the industry or as specified can be accommodatd. Flow through the valve will be determinable from installed instrumentation, and valve disk positions will be determinable without disassembly, such as by the use of non-intrusive diagnostic methods. Valve internal parts are designed with self-aligning features for the purpose of assured correct installation. The maximum loading on the check valve under design basis and required operating conditions will be compared to the allowable structural capability limits for the individual parts of the check valve. The qualification acceptance criteria noted above will include baseline data developed during qualification testing and will be used for verifying the acceptability of the check valves after installation. 3.9.6.2.3.2 Preoperational Testing of Check Valves Each check valve will be tested in the open and/or closed direction, as required by the safety function, under normal operating system conditions. To the extent practical, testing of the valves as described in this section will be performed under fluid temperature and flow conditions which would exist during a cold shutdown, as well as under fluid temperature conditions which would be experienced by the valve during other modes of operation. The testing will identify the flow required to open the valve to the full-open position. The testing will include the effects of rapid pump starts and stops as required by expected system operating conditions. The testing will include any other reverse flow conditions which may be required by expected system operating conditions. Disk movement during valve testing will be examined to verify the leak-tightness of the valve when fully closed. By using methods such as non-intrusive diagnostic equipment, the open valve disk stability will be examined under the flow conditions during normal and other required system operating conditions. The parameters and acceptance criteria for demonstrating that the above functional performance requirements have been fulfilled are as follows:
- During all tests modes which simulate expected system operating conditions, the valve disk fully opens or fully closes as expected based on the direction of the differential pressure across the valve.
e Leak-tightness of valve when fully closed is within established limits, as applicable.
- Valve disk positions are determinable without disassembly.
- Valve testing must verify free disk movement whenever moving to and from the seat.
A$4voved Design Matenal Design of SSC Page 3.9-62
- I i
l
- Sv= tem 80+ oestan controlDocument l
'I e The disk is stable in the open position under normal and other required system operating fluid l
flow conditions. e The valve is correctly sized for the flow conditions specified, i.e., the disk is in full open position i at normal full flow operating condition. e- Valve design features, material, and surfaces accommodate non-intrusive diagnostic testing methods available in the industry or as specified. e Piping system design features accommodate all the applicable check valve testing requirements' as described in Table 3.9-15. 3.9.6.2.3.3 Inservice Testing of Safety-Related Check Valves All safety-related piping systems incorporate provisions for testing to demonstrate the operability of check
. valves under design conditions. The inservice testing of check valves will incorporate the use of :
advanced non-intrusive techniques to periodically assess degradation and the performance characteristics i
- of the check valves. The Part 10 tests will be performed, and check valves which fail to exhibit the required perfonnance can be disassembled for evaluation. The Code provides criteria limits for the test .
. parameters identified in Table 3.9-15. i The detailed IST program will include the frequency and the extent of disassembly and inspection based !
on suspected degradation of all safety-related check valves, including the basis for the frequency and the :' extent of each disassembly. Factors to be considered in the disassembly frequency and extent of (~ disassembly include, but are not limited to: i e Historical performance of the safety-related check valves to identify valves which are prone to degradation / wear, , 4 - e Analysis of trends of valve test parameters and service conditions, e Analysis of valve components which are subject to aging and require a maintenance replacement approach, and > e Results of non-intrusive valve testing. Use of non-intrusive technologies may obviate the need , for inspection / disassembly of safety related check valves altogether, provided the technologies . demonstrate an equivalent ability to detect check valve degradation as inspection / disassembly [ would. 1-
- The program may be revised throughout plant life to minimize disassembly based on past disassembly ;
expenence. ; l 3.9.6.2.4 - Isolation Valve laak Tests j - The leak-tight integrity will be verified for each valve relied upon to provide a leak-tight function. These j
. valves include: !
o Pressure Isolation Valves (PIVs) that provide isolation of a pressure differential from one part O, of a system to another part or between systems. Pressure Isolation Valves associated with the Reactor Coolant System (RCS) are defined in Generic Letter 89-04, Attachment 1, Section 4a. L = coa + anoeww Deew or ssc mm rose 2.s-62 i
Sy^ tem 80+ Design ControlDocument The RCS Pressure Isolation Valves are listed in Table 3.9-16, and they will be tested in accordance with Table 3.9-15 and Technical Specification Surveillance Requirement 3.4.13.1.
- Temperature Isolation Valves (TIVs) whose leakage may cause unacceptable thermal stress, fatigue, or stratification in the piping and thermal loading on supports or whose leakage may cause steam binding of pumps. Safety related valves performing this duty are listed in Table 3.9-15, along with a description of specific leakage test requirements.
- Containment Isolation Valves (CIVs) that provide isolation capability for the piping systems penetrating containment. CIVs are listed along with their required testing in Table 6.2.4-1.
Those CIVs for which a Type-C leakage rate test is specified in Table 6.2.4-1 will also be tested in accordance with OM-10, Subsections 4.2.2.2 and 4.2.2.3(e) and (f). These CIVs are designated in Table 3.9-15 by the valve function CIC. Those CIVs for which a Type-C leakage rate test is not specified in Table 6.2.4-1 are designated in Table 3.915 by the valve function CIN. The CIN valve function designation indicates that these valves are listed in Table 6.2.4-1, but are not leakage rate tested in Tables 6.2.4-1 and 3.9-15. - References for Section 3.9
- 1. MDC STRUDL, McDonnell Douglas Corp., St. Louis Mo.
- 2. " MARC-CDC User Information Manual," Volume I and Volume III, MARC Analysis Corp. and Control Data Corp. Minneapolis, Minn.,1976.
- 3. Not Used.
- 4. MRl/STARDYNE-Static and Dynamic Structural Analysis System: User Information Model, Control Data Corporation, June 1,1970.
- 5. DeSalvo, G. P. and Swanson, J. A., ANSYS-Engineering Analysis System Swanson Analysis Systems, Inc., Elizabeth, Pa.,1972.
- 6. " Dynamic Stress Analysis of Axisymmetric Structures under Arbitrary Loading," Ghosh, S. and Wilson, E., Dept. No. EERC 69-10, University of Califomia, Berkeley, September 1969.
- 7. Topical Report on Dynamic Analysis of Reactor Vessel Internals Under Loss-of-Coolant Accident Conditions with Application of Analysis to C-E 800 Mwe Class Reactors," Combustion Engineering, Inc., CENPD-42, August 1972 (Proprietary).
- 8. " SHOCK, A Computer Code for Solving Lumped-Mass Dynamic Systems," Gabrielson, V. K.,
SCL-DR-65-34, January 1966.
- 9. Not Used.
- 10. Not Used.
- 11. Not Used.
- 12. Not Used.
Approved Design Atatenal- Desen of SSC Page 3.9-64
Sy' tem 80+ Design ControlDocument
- 13. . Not Used.
- 14. " Method for the' Analysis of Blowdown Induced Forces in a Reactor Vessel," Combustion Engineering, Inc., CENPD-252-P, December,1977 (proprietary).
- 15. "CEFLASH-4A: A Fortran-IV-Digital Computer Program for Reactor Blowdown Analysis,"
Combustion Engineering, Inc., CENPD-133P, August,1974 (proprietary).
- 16. "CEFLASH-4A: A Fonran-IV Digital Computer Program for Reactor Blowdown Analysis Cvlodifications)," Combustion Engineering, Inc., CENPD-133P, Supplement 2, February,1975 (proprietary).
- 17. Scherer, A. E., Licensing Manager, (C-E), Letter to D. F. Ross, Assistant Director of Reactor Safety Division of Systems Safety, LD-76-026, March,1976 (proprietary).
- 18. Parr, O. D., Chief Light Water Reactor Project Branch 1-3, Division of Reactor Licensing (NRC), Letter to F. M. Stern, Vice President of Projects (C-E), June,1975.
- 19. Kniel, K., Chief Light Water Reactors Branch No. 2. Letter to A. E. Scherer, Licensing Manager (C-E), August,1976, (Staff Evaluation of CENPD-213).
- 20. SUPERPIPE - A Computer Program for Structural Analysis and Code Compliance Verification of Piping System, Impell Corporation, Walnut Creek, California.
- 21. " Theory of Pump Induced Pulsating Coolant Pressure in PWRs," Penzes, L. E.,2nd Int. Conf.
l . on Structural Mechanics in Reactor Technology, Vol. II, Part E-F.
- 22. " Forced Vibration of a Shell Inside a Narrow Water Annulus," Horvay, G., Bowers, G., Nuclear Engr. Design V34,1975.
- 23. Maine Yankee Precritical Vibration Monitoring Program, Final Report, CENPD-93, February, 1973.
- 24. Omaha Pre-critical Vibration Monitoring Program, Final Report, CEN-8(0), May,1974.
- 25. Calvert Cliffs Analysis of Flow-Induced Structural Response, CEN-4(B).
4
- 26. Comparison of Calvert Cliffs, Maine Yankee and Ft. Calhoun Design Parameters and Flow-Induced Stnictural Response, CENPD-115 Supp.1, April,1974.
- 27. Comparison of ANO-2, Maine Yankee and Ft. Calhoun Reactor Internals Design Parameters and ,
Flow-Induced Structural Response, CEN-8(A)-P Supp.1, May 1975.
- 28. " Response of a Tube Bank to Turbulent Crossflow Induced Excitation," Lubin, B. T., ASME DET-142.
'29 . " Pressure Distribution on Circular Cylinder at Critical Reynolds Numbers," Batham, J. P., J1 Fluid. Mech., v57, pt 2,1973.
L DeeQn ateaor6e!* Deepe of SSC Page 1945
System 80+ Design ControlDocument
- 30. "Supercritical Reynolds Number Simulation for Two Dimensional Flow Over Circular Cylinders," Szecheny. I. E., Jl. Fluid Mech, v70, pt 3,1975. f
- 31. " Properties of Karman Vortex Streets," Chen, Y. N., Sulzer Technical Review,1972.
32, ' Fluctuating Lift Forces of the Karman Vortex Streets on Single Circular Cylinders and in Tube Bundles," Chen, Y. N., Part 1 (ASME-71-VIBR11), Part 2 (ASME-71-VIBR-12), Part e (ASME- ! 71-VIBR-13).
- 33. " Vortex Excited Oscillations of Yawed Circular Cylinders," King, R., ASME 76-WA/FE-16.
- 34. Mark, W. D., Chandiramani, K. L., Karnopp, R. L., Reactor Internals Vibration Study, BBN Report 151273, Oct.1967.
- 35. Final Report on Studies of Flow in a 0.248 Scale Model of the Omaha PWR, H. L. Crawford and L. J. Flanigan, August,1970, Battelle Memorial Institute.
- 36. Final Report On Studies of Flow in a 1/5-Scale Model of the Palisades PWR, CEND-358, L. A.
Schultz, D. A. Trayser, L. J. Flanigan, Battelle Memorial Institute, April,1969.
- 37. Owen, P. R., " Buffeting Excitation of Boiler Tube Vibration," Jl. of Mechanical Science, Vol.
7, N. 4,1965.
- 38. Davies, P.O.A.L., "An Experimental Investigation of the Unsteady Pressure Forces on a Circular Cylinder in a Turbulent Cross Flow," Bruun, H. H., Jl. of Sound & Vibration, V40, N.4,1975.
- 39. "A Study of Free Jet Impingement," Coleman, D. D., et al., Part 2, Jl. Fluid Mech.1971 V.45, Part 3, pg. 477-512.
- 40. " Spectra of Turbulence in a Round Jet," Gibson, M. M., J1. Fluid Mechanics V15,1963.
- 41. Harty, W. C., Rubinstein M. F., Dynamics of Structures, Prentice-Hall,1964.
- 42. " Procedure Guideline for Preparing the ASME Boiler & Pressure Vessel Code, Section III, Design Specification for the System 80+ Certified Design", S80CD-FS-PR001, Rev. 00, January 18, 1993.
- 43. " Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing," USNRC Regulatory Guide 1.20 Rev. 2, May,1976.
- 44. "A Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 1 (System 80 Prototype)," Combustion Engineering, Inc., CEN-263, Rev.1 January,1985 (Proprietary).
- 45. " Structural Analysis of Fuel Assemblies for Seismic and Loss-of-Coolant Accident Loading,"
Combustion Engineering, Inc., CENPD-178, Revision 1, August 1981.
- 46. " Random Vibrations, Elunentary Theory, Structural Dynamics and Design, Signal Analysis and Testing," University of Arizona Seminar, October 29 to November 2,1990.
Appweved Design Material Design of SSC Page 3.9-66
System 80+ Design Control Document
- 47. " Flow Induced Vibration," R. D. Blevins, Second Edition,1990.
- 48. "ATWS: A Reappraisal, Part 3: Frequency of Anticipated Transients," EPRI-NP-2230, January 1982.
- 49. " Development of Transient Initiating Event Frequencies for Use in Probabilistic Risk Assessment," NUREG/CR-3862, May 1985.
- 50. NRC Letter of September 11,1992, " Safety Evaluation on the Use of Single Earthquake Design for Systems, Structures and Components in the ABWR," Docket 52001.
- 51. NCIG-01 " Visual Weld Acceptance Criteria for Structural Welding of Nuclear Power Plants,"
Revision 2 EPRI NP-5380.
- 52. AWS D1.1, The American Welding Society, Structural Welding Code,1990.
- 53. EPRI NP-5639, " Guidelines for Piping System Reconciliation," May 1988.
(v~) A V Anume nie seenner - onew, or ssc rage 3.s-s1
System 80+ Design ControlDocument Table 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components Occurrencest2) O Normal Conditionsul
- 1. Power operation with normal i parameter variations 1,000,000t31
[ Primary and secondary changes enveloped by variations of i 100 psi and i 10'F at normal operating conditions]
- 2. Daily load cycle of 100-50-100% power 22,000
[ Power ramp from 100% to 50% over 2 horrs, stays at 50% for 2 to 10 hours, ramp to 100% over 2 hours, stays at 100% for the remainder of the 24 hour cycle]
- 3. Frequency control 800,000
[ Peak to peak turbine power changes of 10% at 2%/ min] Turbiue Puer steps of i 10 percent (15-100% power) 2,000I31 4. [ Power demands to stabilize grid disturbances] Turbine power steps of i 1 percent (5-15% power) 2,000I31 5.
- 6. Turbine power ramps of i 1 %/ min (5-15% power) 2,000l31
- 7. Turbine load rejection up to 50% (50-100% power) 40
[ Power demands to stabilize grid disturbances]
- 8. Turbine generator runback to house load (15-100% power) 40
- 9. Loss of a main feedwater pump without causing a reactor trip 40 (50-100% power)
- 10. Uncomplicated reactor trips (5-100% power) 60ldl
- 11. NSSS operations with the control systems in the manual mode 2,000 (0-5% power)
- 12. NSSS operations with the control systems in the manual mode 2,000 (5-100% power)
- 13. Opening of the FW economizer valve during power increasing 400 operations
[ Feeding the steam generators with a slug of cold feedwater]
- 14. Startup and coastdown of a Reactor Coolant Pump at hot standby 4,000 conditiou
- 15. Operation of the auxiliary spray system 300
[ Operation of the auxiliary spray system during plant cooldown] O Approwd Desen Material- Design of SSC (11/96) Page 3.9-68
System 80+ Desian contrat oocument i () V
- Table 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components (Cont'd.)
Normal Conditions (Cont'd.)n) Occurrencest2)
- 16. Tie line thermal backup 60t31
[ Turbine load demand of i 20% over 10 minutes, capability to correct overloads of lines tying grids together and prevent grid breakup]
- 17. Plant heatup 300
[Psant heatup at a maximum rate of 100'F/hr between 70 and 556*F except for the pressurizer which has a maximum rate of 200*F/hr between 70 and 653'F, cold feedwater additions to the steam generators at hot standby conditions]
- 18. Plant cooldown 300
[ Plant cooldown at a maximum rate of 100'F/hr between 556 and 70*F except for the pressurizer which has a maximum rate of 200'F/hr between 653 and 70'F] ;
- 19. Shift to high steam generator blowdown 2,000
- 20. Shift from nortnal to maximum CVCS flow rate and return 2,000
- 21. Spurious actuation of the pressurizer spray 40
- 22. Spurious actuation of the pressurizer heaters 40
- 23. Inadvertent closure of one economizer feedwater valve 40
- 24. Inadvertent isolation of one main feedwater heater 40
- 25. Spurious startup of a safety injection pump during shutdown 40 conditions
- 26. Startup and shutdown of the SCS 300
[Startup of the SCS during a plant cooldown and shutdown of the SCS during the subsequent plant heatup] I b v , L ..; Deengre a0etenet - Desigre of SSC (11/96) Page 3.9-69 b
System 80+ oesign controt Document TaNe 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components (Cont'd.) Test Conditionsu3 OccurrencesPI RCS bydrostatic test 10 1. RCS leak test 200 2. Secondary hydrostatic test 10 3.
- 4. Secondary leak test 200
- 5. SIS /SCS check valve operability test 500 SIS /SCS preoperational and maintenance test 240 6.
Upset ConditionsH1 OccurrencesPI Decrease in feedwater temperature 20 1.
- 2. Increase in feedwater flow rate 20 Increase in steam flow rate 20 3.
4 Inadvenent opening of a steam generator relief or safety valve 10 less of load (turbine speed control system operates normally) 19 5. [ Loss of electrical load and normal turbine / generator runback to house load]
- 6. Turbine trip 20
- 7. less of condenser vacuum 20 Imss of non<mergency AC power to the station auxiliaries 10 8.
- 9. less of normal feedwater flow 20
[ Subsequent actuation and cycling of cold emergency feedwater to the steam generators]
- 10. less of forced reactor coolant flow 20
- 11. Uncontrolled CEA withdrawal from suberitical or low power condition 10 O
Approved Destyn Maternal- Design of SSC (11/96) Page 3.9-70
System 80+ Deslan controlDocument Table 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components (Cont'd.) - Upset Conditions (Cont'd.)I!! Occurrencest21
- 12. Uncontrolled CEA withdrawal at power 10
- 13. Control rod misoperation, system malfunction operator error or 20 inadvertent RPCS operation
- 14. Natural circulation cooldown 30
[ Plant cooldown to SCS entry conditions without the reactor coolant pumps)
- 15. Loss of component cooling water to the letdown heat exchanger 10
- 16. CVCS malfunction that increases RCS inventory 20
- 17. Failure of small lines carrying coolant outside containment 5 (sample line break)
- 18. Closure of a single MSIV 5
- 19. Reactor Coolant Pump Seal Failure 10
- 20. loss of RCP seal injection with loss of cooling water 5
- 21. Seismic See Section 3.7.3.2 Emergency Conditions til Occurrencest21
- 1. less of load (with a failure of the turbine speed control system) 1
{ Loss of electrical load, failure of the turbine speed control system and subsequent turbine trip on overspeed] , Faulted Conditionsts j Occurrencest2)
- 1. Inadvertent opening of a SDS valve 1
- 2. Steam system piping failure (SLB) 1
- 3. Feedwater system pipe break (FWLB) 1
- 4. Reactor coolant pump rotor seizure 1
- 5. Reactor coolant pump shaft break 1
- 6. Steam generator tube rupture (SGTR) 1
- 7. Loss of coolant accident (LOCA) 1
- 8. Inadvertent opening of a pressurizer safety valve 1
- 9. Rod ejection accident 1 Notes:
10 Additional information provided in brackets for new System 80+ events and/or events with significant thermal fatigue consequences. Anwoved Desopus aseennef . Desiers of $$C 111/96) Pope 3.9-71
System 80+ oesign controloccument Table 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components (Cont'd.) Notes (Cont'd.): [2] This is the total number of occurrences considered over 60 years for each event. This frequency of occurrence is for design purposes only. The expected number of operational occurrences is lets than this value. An asterisk denotes that the frequency is in each disection (e.g., there are 2000 power steps of
+10% and 2000 power steps of -10%).
131 These are uncomplicated reactor trips. Reactor trips occurring as a consequence of other DBEs are included elsewhere in this table as pan of other events. Idl These events are not con >'dered credible in forming the basis of the NSSS. They are included to demonstrate that the NSS3 comporr.; will not structurally fail even in the highly unli'.cly event that they do occur. A design frequency of occurrence of one (1) in sixty years is assigned for all faulted events. The actual expected frequency of occurrence of these events is much less. O O Approved Design Material ~ Design of SSC Page 3.9-72 l
System 80+ Design ControlDocument i Table 3.9-2 Imading Combinations for ASME Code Class 1, 2, and 3 Componentst31 and l O Component Supports l 0-Condition Design Loading!!! Combination Design PD + DW 1 Level A (Normal)t21 po + ow Level B (Upset)t21 po + ow Level C (Emergency) PO + DW + DE Ixvel D (Faulted) PO + DW + SSE + DF [1] Legend: PD = Design pressure PO = Operating pressure DW = Dead weight /~m \ SSE = Safe Shutdown Earthquake DE = Dynamic system loadings asscciated with the emergency condition DF = Dynamic system loadings associated with pipe breaks (not eliminated by leak before break analysis) [2] As required by ASME Code Section III, other loads, such as thermal transient, and thermal gradient require consideration in addition to the primary stress producing loads listed. SSE is considered in equipment fatigue evaluations in accordance with Section 3.7.3.2. [3] For piping, see Tables 3.9-10 and 3.9-11. p-b AnweredDesso n ateterie! Dwon or ssc tr1/96) Pere 3.9-13
Systens 80+ Design ControlDocument Table 3.9-3 Stress Lhnits for ASME Code Class 1 Components, Piping, and Component Supports g Component and Piping Stress LimitsU l Component Support Stress LimitsI31 Design NB-3221, NB-3231 and NB-3652 NF-3221 or tIF-3321, and NF-3225 level A (Normal) NB-3222, NB-3232 and NB-3653 NF-3221 or NF-3321, and NF-3225 level B (Upset) NB-3223. NB-3233 and NB-3654 NF-3221 or NF-3321, and NF-3225 level C (Emergency) NB-3224, NB-3234 and NB-3655 NF-3221 or NF-3321, and NF-3225 level D (Faulted)t21 NB-3225, NB-3235 and NB-3656 NF-3221 or NF-3321, and NF-3225 Notes: [1] Stress limits listed are used as required by ASME Section Ill, and applicable addenda for all components except active components. Active components are designed to the stress limits of NB-3221 and NB-3231 for Design Conditions and the stress limits of NB-3222 and NB-3232 for all other conditions for active components. [2] For faulted condition loadings, bolts in the load path connecting two members of an NF support for Class I components are designed in accordance with Appendix XVII of the ASME Code for friction type connections with tensile stresses limited to the lesser of 0.7 Su or Sy. [3] Stress limits used are as required by ASME Wion !!! and modified by Regulatory Guide 1.124 and 1.130. Component standard supports may be designed to de limits of NF-3280. O AMweved Desgr Morerwl Desopre of SSC page 3,9 74
Sy~ tem 80+ Deslan controlDocument Table 3.9-4 Seismic I Active Valves O(3 ASME Valve System Neme Valve Section III Actuator No. (Safety Function) Type Code Class Type SI100 Irwst Recirculation Isolation (Operate) Check None 2 SI101 Irwst Recirculation Isolation (Operate) Check None 2 Si 113 Safety Injection System (Operate) Check 2 None SI 123 Safety Injection System (Operate) Check 2 None SI133 Safety Injection System (Operate) Check 2 None SI 143 Safety injection System (Operate) Check 2 None SI 157 Containment Spray System (Operate) Check 2 None SI 158 Containment Spray System (Operate) Check 2 None SI '164 Containment Spray System (Operate) Check 2 None SI 165 Containment Spray System (Operate) Check 2 None SI168 Shutdown Cooling System (Operate) Check 2 None SI178 Shutdown Cooling System (Operate) Check 2 None SI179 Shutdown Cooling System (Operate) Relief 2 None SI189 Shutdown Cooling System (Operate) Relief 2 None SI 215 Safety injection Tank (Operate) Check i None 51 217 Safety injection System (Operate) Check 1 None SI 225 Safety Injection Tank (Operate) Check 1 None
'l227 Safety injection System (Operate) Check 1 None SI235 Safety Injection Tank (Operate) Check 1 None Anprotest Desdyn Afederset. Dessyre of SSC Pepe J.9-75 L-
System 80+ Design ControlDocument Table 3.9-4 Seismic I Active Valves (Cont'd.) ASME Valve System Name Valve Section III Actuator No. (Safety Function) Type Code Class Type S1237 Safety injection System (Operate) Check 1 None Si 245 Safety injection Tank (Operate) C' s 1 None SI247 Safety injection System (Operate) Check i None SI300 CS/SCS IRWST Recirculation Gate 2 Motor lolation SI301 CS/SCS 1RWST Recirculation Gate 2 Motor Isolation SI 302 SI IRWST Recirculation isolation Gate 2 Motor S1 303 SI 1RWST Recirculation Isolation Gate 2 Motor SI 304 IRWST Isolation Gate 2 Motor SI305 IRWST Isolation Gate 2 Motor SI308 IRWST Isolation Gate 2 Motor SI309 IRWST isolation Gate 2 Motor SI310 SCS 1 Flow Control (Operate) Globe 2 Motor SI311 SCS 2 Flow Control (Operate) Globe 2 Motor S1312 SDCHX Bypass (Operate) Globe 2 Motor
$1313 SDCllX Bypass (Operate) Globe 2 M otor S1 314 SCS 1 IRWST Recirculation Line Globe 2 Motor Flow Control SI315 SCS 2 IRWST Recirculation Line Globe 2 Motor Flow Control SI321 liot leg injection (Operate) Globe 2 Motor S1322 Hot leg Injection leakage Return Globe 1 Pneumatic (Close)
SI331 llot leg injection (Operate) Globe 2 Motor Approved Desbyn Matenal Design of SSC Page 3.9-16
System 80+ Design ControlDocument l 1 (' Table 3.9-4 Seismic I Active Valves (Cont'd.) ] N, I ASME ! Valve System Name Valve Section III Actuator No. (Safety Function) Type Code Class Type SI332 Hot leg injection leakage Return Globe 1 Pneumatic (Close) j SI390 Cavity Flooding System (Open) Gate 2 Motor SI 391 Cavity Flooding System (Open) Gate 2 Motor SI392 Cavity Flooding System (Open) Gate 2 Motor , 51393 Cavlty Flooding System (Open) Gate 2 Motor SI 394 Cavity Flooding System (Open) Gate 2 Motor l
)
SI 395 Cavity Flooding System (Open) Gate 2 Motor SI396 IRWST Isolation (Close) Gate 2 Motor SI 397 IRWST Isolation (Close) Gate 2 Motor SI404 Safety injection System (Operate) Check 2 None SI405 Safety injection System (Operate) Check 2 None SI424 S1 Pump Minimum Flow 1RWST Check 2 None Return (Operate) S1426 SI Pump Minimum Flow TVST Check 2 None Return (Operate) 51 434 Safety injection System (Ohrate) Check 2 None SI446 Safety injection System (Operate) Check 2 None S1448 S1 Pump Minimum Flow IRWST Check 2 None Return (Operate) S1 451 SI Pump Minimum Flow IRWST Check 2 None Return (Operate) SI 484 Containment Spray System (Operate) Check 2 None SI485 Containment Spray System (Operate) Check 2 None t 'N- SI 522 Hot Leg Injection (Operate) Check 1 None Amrowd ouinn nearerer onw or ssc rage 3.s-n
i System 80+ Design ControlDocument j Table 3.9-4 _ Seismic I Active Valves (Cont'd.) ASME Valve System Name Valve Section III Actuator No. (Safety Function) Type Code Class Type SI 523 Hot leg injection (Operate) Check 1 None SI$32 Hot leg injection (Operate) Check 1 None 51 533 l{ot Leg Injection (Operate) Check 1 None SI540 Safety injection System (Operate) Check 1 None SI541 Safety injection System (Operate) Check 1 None SI542 Safety Injection System (Operate) Check 1 None SI543 Safety injection System (Operate) Check 1 None SI568 Shutdown Cooling System (Operate) Check 2 None SI569 Shutdown Cooling System (Operate) Check 2 None SI600 Shutdown Cooling System Isolation Globe 2 M otor (Operate) SI601 Shutdown Cooling System isolation Globe 2 Motor (Operate) SI602 Safety injection System Throttle Globe 2 Motor (Operate) 51 603 Safety injection System Throttle Globe 2 Motor (Operate) SI604 Hot Leg Injection (Operate) Gate 2 Motor 51 605 Safety injection Tank Vent (Operate) Globe 2 Solenoid SI606 Safety injection Tank Vent (Operate) Globe 2 Solenoid S1607 Safety injection Tank Vent (Operate) Globe 2 Solenoid SI608 Safety injection Tank Vent (Operate) Globe 2 Solenoid SI609 Hot Leg injection (Operate) Gate 2 Motor SI611 Safety injection Tank Fill / Drain Globe 2 Pneumatic (Operate) AMwwsed Desiger Maternal Desigte of S$C Page 3.9 78
System 80+ Design ControlDocument r Table 3.9-4 Seismic I Active Valves (Cont'd.) b] ASME Valve System Name Valve Section III Actuator No. (Safety Function) Type Code Class Type SI 613 Safety injection Tank Vent (Close) Globe 2 Solenoid SI 614 Safety injection Tank Isolation Gate 1 Motor 4 (Operate) SI616 Safety Injection System Globe 2 Motor , (Operate) SI 618 12akage Return To IRWST/RDT Globe 1 Pneumatic (Close'
' SI 621 ' m; anjection Tank Fill / Drain Globe 2 Pneumatic
,Llose)
SI 623 Safety Injection Tank Vent (Operate) Globe 2 Solenoid SI 624 Safety injection Tank Gate 1 Motor , Isolation (Operate) SI626 Safety injection Globe 2 Motor i Isolation (Operate) 4 p SI628 Leakage Return To IRWST/RDT Globe 1 Pneumatic l
\ (Close)
SI 631 Safety injection Tank Fill / Drain Globe 2 Pneumatic (Close) SI 633 Safety injection Tank Vent (Operate) Globe 2 Solenoid 4 SI 634 Safety injection Tank Isolation Gate 1 Motor (Operate) SI636 Safety Injection System (Operate) Globe 2 M otor S1638 Leakage Return To IRWST/RDT Globe 1 Pneumatic (Close) SI 641 Safety injection Tank Fill / Drain Glebe 2 Pneumatic (Close) SI 643 Safety injection Tank Vent (Operate) Globe 2 Solenoid SI 644 Safety injection Tank Isolation Gate 1 Motor (Operate) 51 646 Safety Injection Isolation (Operate) Globe 2 Motor SI 648 Leakage Return To IRWST/RDT Globe 1 Pneumatic (Close) V S1 651 Shutdown Cooling Suction (Operate) Gate 1 Motor Anwoved Desern hentenet Denmn of SSC Pope 3.9-79
Sy0 tem 80+ oesign controlDocument Table 3.9-4 Seismic I Active Valves (Cont'd.) ASME Valve System Name Valve Section III Actuator No. (Safety Function) Type Code Class Type SI 652 Shutdown Cooling Suction (Operate) Gate 1 Motor S1653 Shutdown Cooling Suction (Operate) Gate 1 Motor SI654 Shutdown Cooling Suction (Operate) Gate 1 Motor SI 655 Shutdown Cooling Suction (Operate) Gate 2 Motor SI 656 Shutdown Cooling Suction (Operate) Gate 2 Motor SI657 CSS 1 IRWST Recirculation Flow Globe 2 Motor Control (Operate) SI 658 CSS 2 IRWST Recirculation Flow Globe 2 Motor Control (Operate) SI 671 Containment Spray isolation (Operate) Gate 2 Motor SI 672 Containment Spray Isolation (Operate) Gate 2 Motor SI682 Sit Fill / Drain (Close) Globe 2 Pneumatic SI686 CSS 1 IRWST Recirculation Isolation Gate 2 Motor SI687 Containment Spray Isolation Gate 2 Motor SI688 SCS 1 IRWST Recirculation Isolation Gate 2 Motor SI693 SCS 2 IRWST Recirculation isolation Gate 2 Motor SI695 Containment Spray isolation Gate 2 Motor 51 6 % CSS 2 IRWST Recirculation Isolation Gate 2 Motor Cll189 CVCS Makeup To IRWST (Close) Check 2 None i CH 255 Seal Injection isolation (Close) Globe 2 Motor ! l C11304 SCS Purification (Close) Check 2 None CH 494 RMW Supply To RDT (CLOSE) Check 2 None i Apprownf Desen Matwiel Desmer of SSC Pope 3.9-80
Sy~ tem 80 + Design ControlDocument p Table 3.9-4 Seismic I Active Valves (Cont'd.) ASME Valve System Name Valve Section HI Actuator No. (Safety Function) Type Code Class Type CH $05 RCP Controlled Bleed-Off Isolation Globe 2 Pneumatic (Close) CH$06 RCP Controlled Bleed-Off Isolation Globe 2 Pneumatic (Close) CH 509 CVCS Makeup To IRWST Isolation Gate 2 Motor (Close) CH 515 letdown Line Isolation (Close) Globe 1 Pneumatic CH516 letdown Line Isolation (Close) Globe 1 Pneumatic CH 523 Letdown Line Isolation (Close) Globe 2 Pneumatic CH 524 Charging Line Isolation (Close) Globe 2 Motor CH 560 RDT Suction Isolation (Close) Globe 2 Pneumatic CH 561 RDT Suction Isolation (Close) Globe 2 Pneumatic CH 575 Letdown Line Isolation (Close) Globe 2 Pneumatic CH 580 RMW Supply Line To RDT Isolation Globe 2 Pneumatic (Close) CH 747 Charging Line (Close) Check 2 None CH 835 Seal Injection (Close) Check 2 None RC 200 RCS (Operate) Safety 1 None RC 201 RCS (Operate) Safety 1 None RC 202 RCS (Operate) Safety 1 None RC 203 RCS (Operate) Safety 1 None RC 244 RCS (Operate) Check i None RC 406 Safety Depressurization System Globe 1 Motor
# RC 407 Safety Depressurization System Globe Motor T 1 U
knproM Onespo Atatarael. Design of SSC Pape 3.9-81
1 System 80+ Design Control Document Table 3.9-4 Seismic I Active Valves (Cont'd.) ASME Valve System Name Valve Section III Actuator No. (Safety Function) Type Code Class Type RC 408 Safety Depressurization System Gate 1 Motor RC 409 Safety Depressurizatior, System Gate 1 Motor RC 410 Safety Depressurization System Globe 1 Solenoid RC 411 Safety Depressurization System Globe 1 Solenoid RC 412 Safety Depressurization System Globe 1 Solenoid RC 413 Safety Depressurization System Globe 1 Solenoid RC 414 Safety Depressurization System Globe 1 Solenoid RC 415 Safety Depressurization System Globe 1 Solenoid RC 416 Safety Depressurization System Globe 1 Solenoid RC 417 Safety Depressurization System Globe 1 Solenoid RC 418 Safety Depressurization System Globe 2 Solenoid RC 419 Safety Depressurization System Globe 2 Solenoid Notes: [1] (Operate) is defined as valve being capable of both opening and closing. [2] (Close) is defined as va!ve being capable of moving to or maintaining a closed position. [3] (Open) is defined as valve being capable of moving to or maintaining an open position. O Approved Design Material Design of SSC page 3.942
System 80+ orian controlDocument
" Table 3.9-5 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Vessels Service Level Stress Limits 01 Design and Level A a , s 1.0 S (a, or at) + a, s 1.5 S Level B a , s 1.1 S (a, or at) + a, s 1.65 S Level C a , s 1.5 S (a, or at) + o, s 1.80 S Level D ams 2.0 S (a, or at) + a, s 2.4 S 133 Stress limits are taken from ASME III, Subsections NC and ND (Table 3321-1).
Table 3.9-6 Stress Criteria for ASME Code Class 2 and Class 3 Inactive Pumps and Pump Supports ("N \' Stress LimitsD1 P ,,[2] Service Level Design and Level A a, s 1.0 S 1.0 (a, or at) + a, s 1.5 S Level B a , s 1.1 S 1.1 (a, or og) + a, s 1.65 S Level C ams 1.5 S 1.2 (a, or at) + a, s 1.80 S Level D ams 2.0 S 1.5 (o, or at) + a, s 2.4 S til Stress limits are taken from ASME III, Subsections NC and ND (Table 3416-1). [21 The maximum pressure shall not exceed the tabulated factors listed under P.,, times the design pressure. m
'w/
4present Destyrt nieterW Desers of SSC Page 3.9-83
Sy~ tem 80 + Design controlDocument Table 3.9-7 Design Criteria for Active Pumps and Pump Supports Servlee Level Stress Limitslil Design and Level A ASME B&PV Section III, Article NC-3400 and ND-3400 Level B a , i 1.0 S a , + a, & l.5 S level C a , i 1.2 S a , + a, i 1.65 S Level D a , i 1.2 S a,+a n i1.8S l'1 The stress limits specified for active pumps are more restrictive than the ASME B&PV Section III limits. For Service level D (membrane plus bending), stresses may exceed 1.8 S but must remain below the material yield stress, in such cases, a deflection analysis is performed to assure that the maximum displacements are within the deflection limits which will not impair the operability of the equipment. O O Apprend Desyn MaterW - Desqrn of SSC Page 3.9-84
- System 80+ oesign comrol Document 3 Table 3.9-8 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Inactive f) v Valves Service Level Stress Limits!'dl P,,,,)51 Design and Level A a , s 1.0 S 1.0 (a, or at) + a s 1.5 S Level B a , s 1.1 S 1.1 (a, or ot) + an s 1.65 S Level C ams 1.5 S 1.2 (a, or at) + on s 1.80 S Level D a, s 2.0 S 1.5 (am or at) + o s 2.4 S Notes:
133 Valve nozzle (piping load) stress analysis is not required when both of the following conditions are satisfied: (1) the section modulus and areas of every plane, normal to the flow, through the region defined as the valve body crotch are at least 110% of those for the piping connected (or joined) to the valve body inlet and outlet nozzles; and (2) code allowable stress, S, for valve body material is equal to (Vn) or greater than the code allowable stress, S, of connected piping material. If the valve body material allowable stress is less than that of the connected piping, the valve section modulus and area as calculated in (1) above shall be multiplied by the ratio of Spip,/S,g.,. If unable to comply with this i requirement, the design by analysis procedure of NB3545.2 is an acceptable altemate trcthod. [2) Casting quality factor of 1.0 shall be used. I'l These stress limits are applicable to the pressure retaining boundary, and include the effects of loads l transmitted by the extended structures, when applicable. j l 183 Design requirements listed in this table are not applicable to valve stems, seat rings, or other parts of valves which are contained within the confines of the body and bonnet. I 151 The maximum pressure resulting from Service Levels B, C, or D shall not exceed the tabulated factors listed under P., times the design pressure or the rated pressure at the applicable operating condition temperature. If the pressure rating limits are met at the operating conditions, the stress limits in this table are considered to be satisfied. I63 Stress limits are taken from ASME III, Subsections NC and ND, (Table 3521-1). (
\
Auprovost Design Asatorset- Design of SSC Pape 3.9-#5 l l
Srtem 80+ oesign controlDocument Table 3.9-9 Stress Criteria for Safety-Related ASME Class 2 and Class 3 Active g Valves W Service Level Stress LimitsUl P )'l level A ASME Section III Article 1.0 NC-3500 and ND-3500 Level B a, s ,1.0 S 1.0 a, + a s 1.5 S Level C a , s 1.2 S 1.1 a, + a, s 1.65 S level D a , s 1.2 S 1.2 am+an s1.8S Notes i through 5 of Table 3.9-8 also apply to this table. VI The stress limits specified for active valves are more restrictive than the ASME B&PV Section Ill limits. For Service Level D (membrane plus bending), stresses may exceed 1.8 S but must remain below the material yield stress. In such cases, a deflection analysis is performed to assure that the maximum displacements are within the deflection limits which will not impair the operability of the equipment. O Approved Des @s Maternal. Desern of SSC page 3,9.gg
}
- Svstem 80+ Design ConeralDocument {
Table 3.9-10 Imading Conibl==tlaas for ASME Section III Class 1 Piping - ; Service Level IAnding Combination , i' Design Design Pressure, Weight, Other Sustamed MachanientIAads ; l levelA Level A Transients, Weight, Operating Pressure, l Thermal Expansion. Anchor Movements, Other Mechanical Loads, Dynamic Fluid Loads . Level B Ievel B Transients, Weight, Coincident Pressure, , 5 Thermal Expansion, Anchor Movements, Safe (( Shutdown Earthquake,# #))I'l Other Mechanical IAnds, Dynamic Fluid IAnds Level C Maximum Pressure, Other Mechanical Loads, ] ' Weight, Dynamic Fluid 1 Ands - level D Maximum Pressure, Other Mechanical Imads, :; Weight, Safe Shutdown Earthquake, Pipe Break 'l Loads, Dynamic Fluid IAads, ((SSE SAMS (Full l 5 Range) Thermal TAMS,* 1hermal Expansion']ll") Notes: ~1he dynamic loads are combined by the square root of the sum of the squares ((1. Alternatively, a lower lent of SSE motion may be used in accordance with Section 3.7.3.2. , p& 2. Loading combinationfor Eq.12a of Reference 50. I
- 3. Loading combinationfor Eq.10; primary plus secondary stress producing load.}}l'1 1 '
L Table 3.9-11 Loading Conibinations for ASME Section III Classes 2 and 3 Piping l . Service Level Loading Combination ! Design Design Pressure, Weight level A & B Operating Pressure, Weight, Other Occasional leads (DFL, Wind)'Ibermal Expansion, Anchor Movements j Level C Maximum Pressure, Weight, Othen Occasional leads l (DFL, Tornado) l Level D Maximum Pressure, Weight, DFL, Safe Shutdown : Earthquake, Pipe Break, (( Anchor Mowments', j . Thermal Espansion#1 ]l") i a Notes: Dynamic fluid loads (DFL) are occasional loads such as safety / relief valve thrust,' steam hammer, water hammer, or loads associated with plant upset or faulted condition as applicable. l , , i ((1. Leading Combinationfor Eq.10b of Reference 50.}}\"1 l l I i- D .
) J l
I43 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5, 'f Anpowed onow nanautor. coop of ssc (2/ssj hoe 2.n1 -
.-,)
System 80+ Design ControlDocument Table 3.9-12 Loading Conditions and Load Combination Requirements for ASME Code Class 1, 2, and 3 Piping Supports Service Level Loading Combination level A Weight Thermalt!) Friction level B Weight Thermal :t1 Dynaanic Fluid leadst2) or Wind level C Weight Thermaltil Dynamic Fluid Imadst21 or Tornado Level D Weight ThermalI81 Dynamic Fluid IAadst21 SSE Inertia
' :. - eismic Movements i
hpe Break Loads O Ill Thermal conditions (including ambient temperature) to be combined to provide maximum load combinations. Ill Dynamic Fluid Loads due to safety / relief valve thrust, steam hammer, and water hammer. Approved Desip Materiel . Desip of SSC Pope 3.9-88
~. .
System 80+ oesign control Document A Table 3.9-13 Stress Limits for CEDM Pressure Housings ~b Service Level Stress Categories And Limits of Stress Intensities 11.21
- 1. Design: Design Pressure, Weight, Other Sustained NB-3221 and Figure NB-3221-1, including Mechanical leads. notes.
- 2. Level A: Normal Operating leading plus Normal NB-3222 and Figure NB-3222-1, including Operating Transients. notes.
- 3. Level B: Normal Operating loading plus Normal NB-3223 and Figures NB-3221 1 and NB-Operating & Upset Transients plus Low Cycle 3222-1, including notes.
Fatigue leading due to Safe Shutdown Earthquake (SSE)f33 Forces.
- 4. Level D: Normal Operating Loadings plus Faulted Article F-1000, Appendix F. Rules for Plant Transients plus Safe Shutdown Earthquake Evaluation of Service Conditions leading Forces plus Loads due to Design Basis Pipe Breaks with Level D Service Limits.
and/or pipe breaks not eliminated by LBB.
- 5. Testing: Testing Plant Transients Paragraph NB-3226 For the above listed operating conditions, the following limits regarding function apply:
- 1. Level A and Level B: The CEDMs are designed to function normally during and after exposure to these conditions.
- 2. Level D: For SSE plus Design Basis Pipe Breaks and/or pipe breaks not eliminated by LBB, the deflections of the CEDM pressure housing are limited to the clastic design limits of Article F-1330, Appendix F (defined above) so that the CEAs can be inserted after exposure to these conditions.
Notes: [1] References listed are taken from Section Ill of the ASME Boiler and Pressure Vessel Code. [2] Dynamic loads including SSE, pipe breaks not eliminated by LBB and Design Basis Pipe Breaks are combined by the SRSS method in accordance with the guidelines of NUREG4484. [3] Alternatively, a lower level of SSE motion may be used in accordance with Section 3.7.3.2.
%J 1
Pu.. : Dee&n Meteniel- Den &n of SSC Page 3.9-89
System 80+ Design ControlDocument Table 3.9-14 Stress Limits for Core Support and Internal Structures Design and Service Loads h Design Limits The core support and internal structures shall be designed to meet the Design Limits defined in NG-3221 of ASME Boiler and Pressure Vessel Code Section III Subsection NG for Design Imadings. Both structures are Safety Class 3, Seismic Category 1 and Quality Class 1 in accordance with ANSI /ANS-51.1-1983. Core Support Structures shall be constructed to the rules in accordance with the ASME Code Section III, Subsection NG-1100. Reactor internals other than core support structures shall meet the guidelines of NG-3000 and be constructed so as not to adversely affect the integrity of the core support structures. Under Level D Loadings, the maximum stress intensity will be obtained from principal stresses resulting from an SRSS combination of LOCA and SSE plus Normal Operating Dynamic and Static loading in accordance with NUREG-0484 Rev. 01. For other than level D loading conditions maximum stress intensity will be derived from an SRSS combination of dynamic loads in accordance with NUREG4484 Rev. 01 or a more conservative summation of stress intensities. Level A Service Limits The core support and internal structures shall be designed to meet the Level A Service Limits defined in NG-3222 of ibid for Level A Service Loadings. Level B Service Limits The core support and internal structures shall be designed to meet the level B Service Limits defined in NG-3223 of ibid for Level B Service loadings. Level C Service Limits The core support and internal structures shall be designed to meet the Level C Service Limits defined in NG-3224 of ibid for Level C Service Loadings. Level D Service Limits The core support structures shall be designed to meet the level D Service Limits defined in NG-3225 of ibid for clastic system analysis of Appendix F of Reference 3.1.2 using Level D Service Loadings. Maximum stress intensity will be obtained from principal stresses resulting from an SRSS combination of pipe break and SSE loadings plus normal operation loads in accordance with NUREG-0484 Rev. 01. O Approved Design Atatorial- Desrgn of SSC Page .1.9-90
Sy-tem 80 + Design controlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (h) (i) Safety Test Test Test Pump Class Parameter Freq Config. Figure No. l CCW Pump 1A 3 DP SPs,SPo,Qi,Q2,V 3 mo. 16 9.2.2-1, Sh 1 CCW Pump IB 3 DP SPs,SPo,Qi,Q2,V 3 mo. 16 9.2.2-1, Sh 1 CCW Pump 2A 3 DP.SPs,SPo,Qi,Q2,V 3 mo. 16 9.2.2-1, Sh 7 CCW Pump 2B 3 DP SPs,$Po,Qi,Q2,V 3 mo. 16 9.2.2-1, Sh 7 MD EFW Pump 1 3 DP,SPs,SPo,Q,V 3 mo. 21 10.4.9-1 Sh 1 TD EFW Pump 1 3 N,DP,SPs,SPo,Q,V 3 mo. 21 10.4.9-1, Sh 1 MD EFW Pump 2 3 DP,SPs,SPo,Q,V 3 mo. 21 10.4.9-1. Sh 1 TD EFW Pump 2 3 N,DP,SPs,SPo,Q,V 3 mo. 21 10.4.9-1, Sh 1 SIPump1 2 DP.SPs,SPo,Q,V (46) 3 mo. 18 6.3.2-1 A SI Pump 2 2 DP.SPs,SPo,Q,V (46) 3 mo. 18 6.3.2 1B SI Pump 3 2 DP.SPs,SPo,Q,V (46) 3 mo. 18 6.3.2-1 A SI Pump 4 2 DP,SPs,SPo,Q,V (46) 3 mo. 18 6.3.2-1B SC Pump i 2 DP.SPs,SPo,Q,V 3 mo. 19 6.3.2-1 A SC Pump 2 2 DP.SPs,SPo,Q,V 3 mo. 19 6.3.2-1B CS Pump 1- 2 DP,SPs,SPo,Q,V 3 mo. 19 6.3.2-1 A CS Pump 2 2 DP.SPs,SPo,Q,V 3 mo. 19 6.3.2-1B SSW Pump 1 A 3 DP.SP,,Q,V 3 mo. 17 9.2.2-1, Sh i SSW Pump IB 3 DP.SPc ,Q,V 3 mo. 17 9.2.2-1 Sh 1 SSW Pump 2A 3 DP,SPc ,Q,V 3 mo. 17 9.2.1-1, Sh 3 1 SSW Pump 2B 3 DP.SPc ,Q,V 3 mo. 17 9.2.1-1, Sh 3 ECW Pump 1A 3 DP,SPs,SPo,Q,V 3 mo, 20 9.2.9-1, Sh 1 j ECW Pump IB 3 DP.SPs.SPo,Q,V 3 mo. 20 9.2.9-1, Sh 1 ECW Pump 2A 3 DP SPs,SPo,Q,V 3 mo. 20 9.2.9-1, Sh 5 ECW Pump 2B 3 DP,SPs,SPo,Q,V 3 mo. 20 9.2.9-1, Sh 5 l DG Building Sump Pump 1 A 3 DP.SPc,Q,V 3 mo. 17 9.5.9-1 l 4 DG Building Sump Pump IB 3 DP,SPc,Q,V 3 mo. 17 9.5.9-1 DG Building Sump Pump 2A 3 DP,SPc,Q,V 3 mo. 17 9.5.9-1 DG Building Sump Pump 2B 3 DP.SPc,Q,V 3 mo. 17 9.5.9-1 RB Subsphere Quad A Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 1 Pump 1 RB Subsphere Quad A Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 1 Pump 2 RB Subsphere Quad B Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 2 ) lh. Pump 1
)
RB Subsphere Quad B Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 2 j Pump 2 l l 4 proved Design neanerial- Design of SSC Page 3.9-91 i
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd,) (h) (i) Safety Test Test Test Pump Class Parameter Freq Config. Figure No. RB Subsphere Quad C Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 1 Pump 1 RB Subsphere Quad C Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 1 Pump 2 RB Subsphere Quad D Sump 3 DP,SPc,Q,V 3 mo. 17 9.3.3-2, Sh 2 Pump i RB Subsphere Quad D Sump 3 DP,SPc,Q,V 3 mo. 9.3.3-2, Sh 2 Pump 2 Spent Fuel Pool Cooling Pump 1 3 DP,SPs,SPo,Q,V 3 mo. 20 9.1.3 Spent Fuel Pool Cooling Pump 2 3 DP,SPs,SPo,Q,V 3 mo. 20 9.1.3 DGE I Motor-driven 3 Note 43 Note 43 Note 43 9.5.4-1, Sh 1 Fuel Oil Booster Pump DGE 2 Motor-driven 3 Note 43 Note 43 Note 43 9.5.4-1, Sh 2 Fuel Oil Booster Pump DG 1 Engine-driven Fuel Oil Pump 3 Note 43 Note 43 Note 43 9.5.4-1, Sh 1 DG 2 Engine-driven Fuel Oil Pump 3 Note 43 Note 43 Note 43 9.5.41, Sh 2 DGE 1 Jacket Water Keep Warm 3 Note 43 Note 43 Note 43 9.5.5-1 Pump DGE 2 Jacket Water Keep Warm 3 Note 43 Note 43 Note 43 9.5.5-1 Pump DG 1 Engine-driven Jacket 3 Note 43 Note 43 Note 43 9.5.5-1 Circulation Pump DG 2 Engine-driven Jacket 3 Note 43 Note 43 Note 43 9.5.5-1 Circulation Pump DGE I Prelube Oil Pump 3 Note 43 Note 43 Note 43 9.5.71, Sh 1 DGE 2 Prelube Oil Pump 3 Note 43 Note 43 Note 43 9.5.7-1, Sh 2 DG 1 Engine-driven Lube Oil Pump 3 Note 43 Note 43 Note 43 9.5.7-1, Sh 1 DG 2 Engine-driven Lube Oil Pump 3 Note 43 Note 43 Note 43 9.5.7-1, Sh 2 O' Apnproved Desiger Motorial. Design of SSC Page 3.9-92
O O O System 80+ Design ControlDocumnt Table 3.9-15 Inservice Testing of Safety-Related Pu:nps and Valves (Cont'd.) (a) (b) (c) (d) (c) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq~ Config. Fig. No. CC-100 CCW IIX I A Bypass Control GL AD 3 B S 3 mo. - 9.2.2-1, Sh 1 MT 3 mo. - FS 3 mo. - LPV 2 yr. - CC-101 CCW llX IB Bypass Control GL AD 3 B S 3 mo. - 9.2.2-1, Sh 1 MT 3 mo. - FS 3 mo. - LPV 2 yr. - CC-102 Non. essential Supply IIcader i Isolation BF AD 3 B S CS(3) - 9.2.2-1, Sh 1 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-103 Non-essential Return Header 1 Isolation BF AD 3 B S CS(3) - 9.2.2-1. Sh 1 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-106 CCW liX 1 A Inlet Isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 1 MT 3 me. - LPV 2 yr. - CC-107 CCW HX IB Inlet isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 1 MT 3 mo. - LPV 2 yr. - CC-108 CCW llX 1 A Outlet Isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh ! MT 3 mo. - LPV 2 yr. - CC-109 CCW IIX IB Outlet isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 1 MT 3 mo. - LPV 2 yr. - Approved Design Material- Design of SSC P*90 3 9-93
System 80+ - _. _ . - . Design Control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (I) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. CC-Il0 SCS IIX 1 Contro! GL AD 3 B S 3 mo. - 9.2.2-i, Sh 2 h1T 3 rno. - FS 3 mo. - LPV 2 yr. - CC-Ill SCS IlX 1 Outlet isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 2 htT 3 mo. - LPV 2 yr. - CC-112 SFP llX 1 Control GL AD 3 B S 3 mo. - 9.2.2-1, Sh 2 h1T 3 mo. - FS 3 mo. - LPV 2 yr. - CC-113 SFP llX 1 Outlet Isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 2 MT 3 mo. i - LPV 2 yr. - CC-114 CS IlX 1 Outiet Isolation BF EL 3 B S 3 me. - 9.2.2-1, Sh 2 MT 3 mo. - LPV 2 yr. - CC-122 Non-essential Supply lleader 1 Isolation BF AD 3 B S CS(3) - 9.2.2-1, Sh 1 MT CS(3) - FS CS(3) - LFV 2 yr. - CC-123 Non-essential Return lleader 1 isolation BF AD 3 B S CS(3) - 9.2.2-1, Sh 1 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-130 CCW Supply to RCP 1 A,lB BF EL 2 A CIC S CS(l) - 9.2.2-1, Sh 5 MT CS(l) - LPV 2 yr. 2 LT 2 yr. 2 Approved Design Material- Design of SSC Page 3.9-94 O O O
O O O System 80+ Design ControlDocumnt Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freg Config. Fig. No. CC-1302 CCW Pump 1 A Discharge CK SA 3 C S 3 mo. 11 9.2.2-1, Sh 1 RF 3 mo. 9 CC-1303 CCW Pump IB Discharge CK SA 3 C S 3 mo. 11 9.2.2-1, Sh 1 RF 3 mo. 9 CC-1306 CCW Pump I A Surge Tank Sparger CK SA 3 C S 3 mo. - 9.2.2-1, Sh I RF 3 mo. 9 CC-1307 CCW Pump IB Surge Tank Sparger CK SA 3 C S 3 mo. - 9.2.2-1, Sh 1 RF 3 mo. 9 CC-131 CCW Supply to RCP 1 A,lB BF EL 2 A CIC S CS(l) - 9.2.2-1, Sh 5 MT CS(l) - LPV 2 yr. - LT 2 yr. 2 CC-1328 Makeup to CCW Surge Tank I from SSWS CK SA 3 C S 3 mo. 14 9.2.2-1, Sh 1 RF 3 mo. 9 CC-1331 SC IIX 1 IIeader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1337 SI Pump Motor Cooler 1 IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1334 SC Miniflow liX 1 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1350 SC Pump Motor Cooler i Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1. Sh 2 CC-1356 EFW Pump Motor Cooler 1 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-136 CCW Return from RCP 1A,lB BF EL 2 A CIC S CS(l) - 9.2.2-1, Sh 5 MT CS(l) - LPV 2 yr. - 1 LT 2 yr. 3 CC-1362 CS Pump Motor Cooler i licader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1368 CS Miniflow IlX 1 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1 Sh 2 Approved Design Material- Design of SSC Page 3.9-95
System 80+ Design contret Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (c) (1) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Ftrq Config. Fig. No. CC-137 CCW Return from RCP I A,lB BF EL 2 A CIC S CS(l) - 9.2.2-1, Sh 5 MT CS(l) - LPV 2 yr. - LT 2 yr. 3 CC-1374 SI Pump Motor Cooler 3 IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1380 SFP Cooling HX 1 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1384 SFP Cooling Pump Motor Cooler i Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1390 CS IlX 1 licader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 2 CC-1507 CCW Supply to RCP 1A,lB CK SA 2 A/C CIC S CS(2) 11 9.'2.2-1, Sh 5 LT 2 yr. 2 RF RO(2) 2 CC-1548 CCW Return from RCP 1 A,lB CK SA 2 A/C CIC S RO(2) 12 9.2.2-1, Sh 5 LT 2 yr. 3 RF CS(2) - CC-1591 CCW Pump Motor Cooler I A Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 3 CC-1597 CCW Pump Motor Cooler IB lleader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 3 CC-1603 DGE Jacket Water Cooler I Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 3 CC-1609 ECW Condenser i lleader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1. Sh 3 CC-1637 CilG Pump Motor Cooler I Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 4 CC-1643 CIIG Pump Miniflow IIX 1 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1 Sh 4 CC-1650 Instrument Air Comp 1 A IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 6 CC-1656 Instrument Air Comp 1B IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1 Sh 6 CC-1851 DGE Start Air Aftercooler I A IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1 Sh 3 CC-1857 DGE Start Air Aftercooler IB lleader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 3 CC-XXXX CCW Surge Tank i Vacuum Breaker RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh i Approved Design Material- Design of SSC Page 3.9-9G O O O
7 ~.3 s, h System 8gf _ _ _ _ _ _ _ w cony Docenent Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safdy Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. CC-200 CCW IIX 2A Bypass Control GL AD 3 B S 3 mo. - 9.2.2-1, Sh 7 MT 3 mo. - FS 3 mo. - LPV 2 yr. - CC-201 CCW HX 2B Bypass Control CL AD 3 B S 3 mo. - 9.2.2-1. Sh 7 MT 3 mo. - FS 3 mo. - LPV 2 yr. - CC-202 Non-essemial Supply lleader 2 Isolation BF AD 3 B S CS(3) - 9.2.2-1, Sh 7 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-203 Non-essential Return Header 2 Isolation BF AD 3 B S CS(3) - 9.2.21, Sh 7 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-206 CCW llX 2A Inlet Isolation BF BF EL 3 B S 3 mo. - 9.2.2-1, Sh 7 MT 3 mo. - LPV 2 yr. - ' CC-207 CCW 11X 2B Inlet Isolation BF BF EL 3 B S 3 mo. - 9.2.2-1, Sh 7 MT 3 mo. - LPV 2 yr. - i CC-208 CCW IIX 2A Outlet Isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 7 MT 3 mo. - LPV 2 yr. - CC-209 CCW IIX 2B Outlet Isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 7 MT 3 mo. - LPV 2 yr. - Approved Destgrr Material- Design of SSC P*9* 3 9*97
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. CC-210 SCS IIX 2 Control GL AD 3 B S 3 mo. - 9.2.2-1. Sh 8 MT 3 mo. - FS 3 mo. - LPV 2 yr. - CC-211 SCS IIX 2 Outlet Isolation BF BF EL 3 B S 3 mo. - 9.2.2-1 Sh 8 MT 3 mo. - LPV 2 yr. - CC-212 SFP IIX 2 Control GL AD 3 B S 3 mo. - 9.2.2-1, Sh 8 MT 3 mo. - FS 3 mo. - LPV 2 yr. - CC-213 SFP llX 2 Outlet isolation BF BF EL 3 B S 3 mo. - 9.2.2-1, Sh 8 MT 3 mo. - LPV 2 yr. - CC-214 CS IIX 2 Outlet Isolation BF EL 3 B S 3 mo. - 9.2.2-1, Sh 8 MT 3 mo. - LPV 2 yr. - CC-222 Non-essential Supply ifeader 2 Isolation BF AD 3 B S CS(3) - 9.2.2-1, Sh 7 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-223 Non-essential Return llender 2 Isolation BF AD 3 B S CS(3) - 9.2.2-1, Sh 7 MT CS(3) - FS CS(3) - LPV 2 yr. - CC-230 CCW Supply to RCP 2A,2B BF EL 2 A CIC S CS(l) - 9.2.2-1, Sh 11 MT CS(l) - LPV 2 yr. - LT 2 yr. 2 Approved Design Material- Design of SSC Page 3.9-98 O O O
i O O O 9.- - . _ _ _ _ _ _ - - Table 3.9-15 Inservice Testing of Safety-Related Pumps 'and Valves (Cont'd.) , (a) (b) (c). .(d) (e) (f) - (3) . (i) Valve Valve Valve Valve Safety Code Vs4ve Test Test Test ' N. scription Type Act Class Cat I%ect Reqd Fhq Config. Hg. N. CC-2302 CCW Pump 2A Dis.' arge CK SA 3 C S 3 mo. I1 9.2.2-1, Sh 7 [- RF 3 mo. 9 CC-2303 CCW Pump 2B Discharge CK SA 3 C S 3 mo. I1 9.2.2-1, Sh 7 RF 3 mo. 9
- CC-2306 - CCW Pump 2A Surge Tank Sparger CK SA 3 C S 3 mo. - 9.2.2-1, Sh 7 '
RF 3 mo. 9 CC-2307 CCW Pump 2B Surge Tank Sparger CK SA 3 C S 3 mo. - 9.2.2-1, Sh 7 RF 3 mo. 9 CC-231 CCW Supply to RCP 2A,2B BF EL 2 A CIC S CS(l) . - 9.2.2-1, Sh iI MT CS(l) - 9.2.2-1, Sh 11 LPV 2 yr. - LT 2 yr. 2
; CC-2328 Makeup to CCW Surge Tank 2 from SSWS CK SA 3 C S 3 mo. 14 9.2.2-1, Sh 7 -
RF 3 mo. 9 CC-2331 SC HX 2 Header Relief RV ' SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 CC-2337 St Pump Motor Cooler 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 CC-2344 SC Miniflow HX 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1 Sh 8 s CC-2350 SC Pump Motor Cooler 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 r CC-2356 EFW Pump Motor Cooler 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 CC-236 CCW Return from RCP 2A,2B BF- EL 2 A CIC S CS(l) - 9.2.2-1 Sh 11 MT CS(l) - LPV 2. yr. - LT 2 yr. 3 CC-2362 CS Pump Motor Cooler 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 - CC-2368 CS Miniflow HX 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 L. 2 Deefpn n9ererW Des 4pn of SSC Pese 3.9-99 j
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System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Fonct Reqd Freq Config. Fig. No. CC-237 CCW Return from RCP 2A,2B BF EL 2 A CIC S CS(l) - 9 2.2-1, Sh 1I MT CS(l) - LPV 2 yr. - LT 2 yr. 3 CC-2374 SI Pump Motor Cooler 4 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 CC-2380 SFP Cooling IIX 2 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1. Sh 8 CC-2384 SFP Cooling Pump Motor Cooler 2 lleader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 8 CC-2390 CS IIX 2 Ileader Relief RV SA 3 C RVT 10 yr. - 9/2.2-1, Sh 8 CC-240 CCW Supply to Letdown llX BF EL 2 A CIC S CS(4) - 9.2.2-1, Sh 14 MT CS(4) - LPV 2 yr. - LT 2 yr. 2 CC-241 CCW Supply to Letdown llX BF EL 2 A CIC S CS(4) - 9.2.2-1, Sh 14 MT CS(4) - LPV 2 yr. - LT 2 yr. 2 CC-2 02 CCW Return from Letdown IIX BF EL 2 A CIC S CS(4) - 9.2.2-1, Sh 14 MT CS(4) - LPV 2 yr. - LT 2 yr. 3 CC-243 CCW Return from Letdown IIX BF EL 2 A CIC S CS(4) - 9.2.2-1, Sh 14 MT CS(4) - LPV 2 yr. - LT 2 yr. 3 CC-2507 CCW Supply to RCP 2A,2B CK SA 2 A/C CIC S CS(2) I1 9.2.2-1, Sh 11 LT 2 yr. 2 RF RO(2) 2 Approved Design Material- Design of SSC Page 3.9-100 0 0 0
p f~% O b d V System 80+ Design ConnelDocumerrt Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funet Reqd Freq Config. Fig. No. CC-2548 CCW Return from RCP 2A,2B CK SA 2 A/C CIC S RO(2) 12 9.2.2-1 Sh 11 LT 2 yr. 3 RF CS(2) - CC-2591 CCW Pump Motor Cooler 2A IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 9 CC-2597 CCW Pump Motor Cooler 2B IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 9 CC-2603 DGE Jacket Water Cooler 2 IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1 Sh 9 CC-2609 ECW Condenser 2 Header Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 9 CC-2622 CCW Supply to Letdown IIX CK SA 2 A/C CIC S CS(5) 11 9.2.2-1, Sh 14 LT 2 yr. 2 RF RO(5) 2 CC-2628 CCW Return from Letdown IIX CK SA 2 A/C CIC S RO(5) 13 9.2.2-1, Sh 14 LT 2 yr. 3 RF CS(5) - CC-2637 CIIG Pun.p Motor Cooler 2 IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 10 CC-2643 CIIG Pump Miniflow llX 2 Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 10 CC-2650 Instrument Air Comp 2A Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 14 CC-2656 Instrument Air Comp 2B IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 14 CC-2851 DGE Start Air Aftercooler 2A Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 9 CC-2857 DGE Start Air Aftercooler 2B lleader Relief RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 9 CC-XXXX CCW Surge Tank 2 Vacuum Breaker RV SA 3 C RVT 10 yr. - 9.2.2-1, Sh 7 Cil-115 Volume Control Tank Relief RV SA 3 C RVT 10 yr. - 9.3.4-1 Sh 2 Cil-189 CVCS to IRWST Boron Recoveny Return CK SA 2 A/C CIC S 3 mo. - 9.3.4-1, Sh 2 LT 2 yr. - RF 3 mo. - Cil-199 Seal Injection Return IIcader Relief RV SA 2 C RVT 10 yr. - 9.3.4-1, Sh 2 Approved Design Material- Design of SSC Page 3.9-101
W#1** 80f- - __- - - _ - - - . - - . -_-
-__Desynfontrol Docsanent Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.)
(a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Valve Valve Safety Cook Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. Cil-255 Seal injection Containment Isolation GL EL 2 A CIC S CS(6) - 9.3.4-1, Sh 1 MT CS(6) - LPV 2 yr. - LT 2 yr. I Cil-304 Shutdown Purification to LD 11X CK SA 2 A/C CIC LT 2 yr. I 9.3.4-1 Sh 1 RF CS(27) - Cil-307 Shutdown Purification to LD HX GT M 2 A P,CIC LT 2 yr. I 9.3.4-1, Sh I Cil-354 Letdown to EDT Relief RV SA 3 C RVT 10 yr. - 9.3.4-1, Sh 1 Cil-432 Charging Line Bypass Relief RV SA 2 C RVT 10 yr. - 9.3.4-1 Sh I Cil-494 Resin Sluice Supply lleader to Reactor Drain CK SA 2 A/C CIC S 3 mo. 12 9.3.4-1, Sh 3 Tank LT 2 yr. I RF 3 mo. - Cil-505 RCP Seal Return GL AD 2 A CIC S CS(7) - 9.3.4-1, Sh 2 MT CS(7) - FS CS(7) - LPV 2 yr. - LT 2 yr. 4 Cll-506 RCP Seal Return GL AD 2 A CIC S CS(7) - 9.3.4-1, Sh 2 MT CS(7) - FS CS(7) - LPV 2 yr. - LT 2 yr. 4 Cil-509 CVCS IRWST r.oron Recovery Return GT EL 2 A CIC S 3 mo. - 9.3.4-1, Sh 2 MT 3 mo. - LPV 3 mo. - LT 2 yr. 1 Approved Design Material- Deshyn of SSC Page 3.9-102 O O O
snaaso+ _ __ me a~e-c : Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) - i t (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Hg.No. Imp 2B Letdown Isolation GL AD 9.3.4-1, Sh 1 CH-515 - 1 B S CS(8) - , MT CS(8) - FS CS(8) - l LPV 2 yr. - i CH-516 Imp 2B Letdown Isolation GL AD 1 A TIV S CS(8) - 9.3.4-1, Sh 1 MT cst;) - FS CS(8) - LPV 2 yr. - t LT (8) - I Cil-523 LP 2B Letdown Containment Isolation GL AD 2 A CIC S CS(8) - 9.3.4-1, Sh I ' MT CS(8) - FS CS(8) - LPV 2 yr. - LT 2 yr. 4 ; CH-524 CVCS Charging Line Isolation GL EL 2 A CIC S' CS(9) - 9.3.4-1, Sh 1 , MT CS(9) - LPV 2 yr. - LT 2 yr. 1 Cil-560 Reactor Drain Tank Discharge Isolation GL AD 2 A CIC S 3 mo. - 9.3.4-1, Sh 3 MT 3 mo. - ; 3 mo. FS - LPV 2 yr. - LT 2 yr. 4 Cil-561 Reactor Drain Tank Discharge Isolation GL AD 2 A CIC S 3 mo. - 9.3.4-1, Sh 3 - , MT 3 mo. - l FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 Approved Design Meteriel- Design of SSC . Pope 3.9-103
System 80+ _-_ _ - - - _ _ - - - - - _ - - - . oesign controlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. Cil-575 LP 2B letdown Containment isolation GL AD 2 A CIC S CS(8) - 9.3.4-1, Sh 1 MT CS(8) - FS CS(8) - LPV 2 yr. - LT 2 yr. 4 Cil-580 Resin Sluice Supply lleader to Reactor Drain GL AD 2 A CIC S 3 mo. - 9.3.4-1, Sh 3 Tank Isolation MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 1 Cil606 Dedicated Seal Injection Pump Recirculation RV SA 3 C RVT 10 yr. - 9.3.4-1, Sh 2 Relief Cil-607 Dedicated Seal Injection Pump Discharge Relief RV SA 3 C RVT 10 yr. - 9.3.4-1, Sh 2 Cil-657 Equipment Drain Tank Relief RV SA 3 C RVT 10 yr. - 9.3.4-1, Sh 3 Cil-717 Charging Pumps
- Mini-flow Line Relief RV SA 3 C RVT 10 yr. - 9.3.4-1, Sh 2 Cll-747 CVCS Charging Line CK SA 2 A/C CIC LT 2 yr. I 9.3.4-1, Sh I RF CS(10) -
CII-835 Seal Injection Containment Isolation CK SA 2 A/C CIC LT 2 yr. I 9.3.4-1, Sh 1 RF CS(ll) - Cil-865 Seal Injection IIcat Exchanger Outlet Relief RV SA 3 C RVT 10 yr. - 9.3.4-1. Sh 1 DF-130 FO Day Tank I Level Control GL AD 3 B S 3 mo. - 9.5.4-1, Sh 1 MT 3 mo. - FS 3 mo. - LPV 2 yr. - DF-230 FO Day Tank 2 Level Control GL AD 3 B S 3 mo. - 9.5.4-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - Approved Design Material- Design of SSC Page 3.9-104 , 9 O O
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%.J System 80+
oestan contna occument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. DS-Il0 Start Air Receiver I A Inlet CK SA 3 A/C RF 3 mo. 9 9.5.6-l. Sh I LT 2 yr, 1 DS-Il2 Start Air Receiver I A Outlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh I RF 3 mo.(35) - DS-Il3 SA Supply to Engine Control Panel I CK SA 3 C S 3 mo. 15 9.5.6-1, Sh I RF 3 mo. 9 DS-II5 DGE 1 SA Left Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 1 MT 3 mo. - FS 3 mo. - LPV 2 yr. - DS-116 DGE I SA Left Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 1 MT 3 mo. - FS 3 mo. - LPV 2 yr. - DS-117 DGE I SA Left Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh I RF 3 mo.(35) - DS-118 DGE I SA left Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh I ! RF 3 mo.(35) - DS-120 Start Air Receiver IB Inlet CK SA 3 A/C RF 3 mo. 9 9.5.6-1, Sh I LT 2 yr. I DS-122 Start Ai- Receiver IB Outlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1. Sh I RF 3 mo.(35) - DS-123 SA Supply to Engine Control Panel 1 CK SA 3 C S 3 mo. 15 9.5.6-1, Sh 1 RF 3 mo. 9 DS-125 DGE I SA Right Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 1 MT 3 mo. - FS 3 mo. - LPV 2 yr. - Approved Design Materiel- Deslyn of SSC Page 3.9-105
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. DS-126 DGE I SA Right Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 1 MT 3 mo. - 1 FS 3 mo. - LPV 2 yr. - DS-127 DGE I SA Right Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 1 RF 3 mo.(35) - DS-128 DGE I SA Right Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 1 RF 3 mo.(35) - DS-210 Start Air Receiver 2A Inlet CK SA 3 A/C RF 3 mo. 9 9.5.6-1, Sh I LT 2 yr. 1 DS-212 Start Air Receiver 2A Outlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 1 RF 3 mo.(35) - DS-213 SA Supply to Engine Control Panel 2 CK SA 3 C S 3 mo. 15 9.5.6-1, Sh 1 RF 3 mo. 9 DS-215 DGE 2 SA left Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 2 MT mo. - FS 3 mo. - LPV 2 yr. - DS-216 DGE 2 SA Left Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - DS-217 DGE 2 SA Left Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 2 RF 3 mo.(35) - DS-218 DGE 2 SA Left Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 2 RF 3 mo.(35) - l DS-220 Start Air Receiver 2B Inlet CK SA 3 C RF 3 mo. 9 9.5.6-1, Sh I LT 2 yr. 1 Approved Design Materia!- Design of SSC Page 3.9-106 O O O
A pJ v J System 80+ _ __ Deskn ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d). (e) (0 l (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. DS-222 Start Air Receiver 2B Outlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 1 RF 3 mo.(35) - DS-223 SA Supply to Engine Control Panel 2 CK SA 3 C S 3 mo. 15 9.5.6-1 Sh 1 RF 3 mo. 9 DS-225 DGE 2 SA Right Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - DS-226 DGE 2 SA Right Bank Inlet GT S 3 B S 3 mo. - 9.5.6-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - DS-227 DGE 2 SA Right Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 2 RF 3 mo.(35) - DS-228 DGE 2 SA Right Bank Inlet CK SA 3 C S 3 mo.(35) 15 9.5.6-1, Sh 2 RF 3 mo.(35) - EF-100 Steam-driven EFW Pump #1 SG isolation GT EL 2 A TIV, S 3 mo. - 10.4.9-1, Sh I CIN MT 3 mo. - LPV 2 yr. - LT 2 yr. (24,45) - 8 EF-101 Steam-driven EFW Pump #2 SG Isolation GT EL 2 A TIV, S 3 mo. - 10.4.9-1, Sh I CIN MT 3 mo. - LPV 2 yr. - LT 2 yr. (24,45) - 8 Approved Design Material- Design of SSC Page 3.9-107
l System 80+ _ - - _ - _ _ - - - - - - - - - - - Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. EF-102 Motor-driven EFW Pump #1 SG Isolation GT EL 2 A TIV, S 3 mo. - 10.4.9-1, Sh I CIN MT 3 mo. - LPV 2 yr. - LT 2 yr. (24,45) - 8 EF-103 Motor-driven EFW Pump #2 SG isolation GT EL 2 A TIV, S 3 mo. - 10.4.9-1, Sh I CIN MT 3 mo. - LPV 2 yr. - LT 2 yr. (24,45) - 8 EF-104 Steam-driven EFW Pump #1 FLOW Control GL EL 3 B S 3 mo. - 10.4.9-1, Sh 1 MT 3 mo. - LPV 2 yr. - EF-105 Steam-driven EFW Pump #2 Flow Control GL EL 3 B S 3 mo. - 10.4.9-1, Sh 1 MT 3 mo. - LPV 2 yr. - EF-106 Motor-driven EFW Pump #1 Flow Control GL EL 3 B S 3 mo. - 10.4.9-1, Sh 1 MT 3 mo. - LPV 2 yr. - EF-107 Motor-driven EFW Pump #2 Flow Control GL EL 3 B S 3 mo. - 10.4.9-1, Sh 1 MT 3 mo. - LPV 2 yr. - EF-108 EFW Pump Turbine #1 Steam Supply isolation GT AD 2 B S 3 mo. - 10.4.9-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - Approved Design Material- Design of SSC Page 3.9-108 O O O
~
V U' V' .System 80 + _ __ oesign contror Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd ) (a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Vahe Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. EF-109 EFW Pump Turbine #2 Steam Supply isolation GT AD 2 B S 3 mo. - 10.4.9-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - EF-110 EFW Pump #1 Continuous Steam-drain GT EL 2 B S 3 mo. - 10.4.9-1 Sh 2 MT 3 mo. - LPV 2 yr. - EF-111 EFW Pump #2 Continuous Steam-drain GT EL 2 B S 3 mo. - 10.4.9-1, Sh 2 MT 3 mo. - LPV 2 yr. - EF-il2 EFW Pump Turbine #1 Steam Supply Bypass GT AD 2 B S 3 mo. - 10.4.9-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - EF-113 EFW Pump Turbine #2 Steam Supply Bypass GT AD 2 B S 3 mo. - 10.4.9-1, Sh 2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - EF-200 Steam-driven EFW Pump #1 SG Isolation CK SA 2 C CIN S CS(12) 12 10.4.9-1, Sh 1 RF 3 mo.(12) 9 LT 2 yr.(45) 9 EF-201 Steam-drisen EFW Pump #2 SG Isolation CK SA 2 C CIN S CS(12) 12 10.4.9-1, Sh I RF 3 mo.(12) 9 LT 2 yr.(45) 9 EF-202 Mo'or-driven EFW Pump #1 SG Isolation CK SA 2 C CIN S CS(12) 12 10.4.9-1, Sh 1 RF 3 mo.(12) 9 LT 2 yr.(45) 9 EF-203 Motor-driven EFW Pump #2 SG isolation CK SA 2 C CIN S CS(12) 12 10.4.9-1, Sh I RF 3 mo.(12) 9 LT 2 yr.(45) 9 Approved Design Staterial- Design of SSC Page 3.9-109
System 80+ _ - - . - _ . --- _ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (s) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. EF-204 Steam-driven EFW Pump #1 Discharge CK SA 3 C S CS(12) 1I 10.4.9-1, Sh 1 RF 3 mo.(12) 9 EF-205 Steam-driven EFW Pump #2 Discharge CK SA 3 C S CS(12) 11 10.4.9-1, Sh 1 RF 3 mo.(12) 9 EF-206 Motor-driven EFW Pump #1 Discharge CK SA 3 C S CS(12) 11 10.4.9-1, Sh 1 RF 3 mo.(12) 9 EF-207 Motor-driven EFW Pump #2 Discharge CK SA 3 C S CS(12) 11 10.4.9 . Sh 1 RF 3 mo.(12) 9 EF-288 Steam-driven EFW Pump #1 Crossover Manual GT M 3 B S 3 mo. - 10.4.9-l. Sh ! j isolation Valve ! EF-289 Steam-driven EFW Pump #2 Crossover Manual GT M 3 B S 3 mo. - 10.4.9-1, Sh I Isolation Valve ! EF-290 Motor-driven EFW Pump #i Crossover Manual GT M 3 B S 3 mo. - 10.4.9-1. Sh 1 Isolation Valve EF-291 Motor-driven EFW Pump #2 Crossover Manual GT M 3 B S 3 mo. - 10.4.9-1, Sh 1 Isolation Valve EF-XXX Steam-driven EFW Pump #1 Turbine Trip and GL EL 2 B S(39) RO(39) - 10.4.9-1, Sh 2 l Throttle Valve t EF-XXX Steam-driven EFW Pump #2 Turbine Trip and GL EL 2 B S(39) RO(39) - 10.4.9-1, Sh 2 Throttle Valve EF-XXX Steam-driven EFW Pump #1 Turbine Governor GL SA 3 B S(39) 3 mo.(39) - 10.4.9-1, Sh 2 Valve l EF-XXX Steam-driven EFW Pump #2 Turbine Governor GL SA 3 B S(39) 3 mo.(39) - 10.4.9-1, Sh 2 i Vdve PC-206 SFPC Pump #1 Discharge Check Valve CK SA 3 C S 3 mo. 11 9.1-3 RF 3 mo. 9 PC-207 SFPC Pump #2 Discharge Check Valve CK SA 3 C S 3 mo. I1 9.1-3 l RF 3 mo. 9 Approved Design Material- Design of SSC Page 3.9-110 0 9 9
O O O System 80+ Deshyn ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (e) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. PC-257 Refueling Cavity Cleanup Sctn PK M 2 A P.CIC LT 2 yr. 4 9.1-3 PC-258 Refueling Cavity Cleanup Sctn PK M 2 A P.CIC LT 2 yr. 4 9.1-3 PC-291 Refueling Cavity Cleanup Sctn PK M 2 A P.CIC LT 2 yr. 4 9.1-3 PC-292 Refueling Cavity Cleanup Sctn PK M 2 A P,CIC LT 2 yr. 4 9.1-3 RC-100E Pressurizer Spray Contrcl Valve GL AD 1 B LPV 2 yr. - 5.1.2-3 RC-100F Pressurizer Spray Control Valve GL AD 1 B LPV 2 yr. - 5.1.2-3 RC-200 Pressurizer Safety Valve RV SA 1 C RVT 5 yr. - 5.1.2-3 RC-201 Pressurizer Safety Valve RV SA 1 C RVT 5 yr. - 5.1.2-3 RC-202 Pressurizer Safety Valve RV SA 1 C RVT 5 yr. - 5.1.2-3 RC-203 Pressurizer Safety Valve RV SA 1 C RVT 5 yr. - 5.1.2-3 RC-403 Reactor Vessel Closure llead leakoff Valve GL S 2 B LPV 2 yr. - 5.1.2-1 RC-406 Rapid Depressurization Valve GL EL 1 B S CS(13) - 5.1.2-3 MT CS(13) - LPV 2 yr. - RC-407 Rapid Depressurization Valve GL EL 1 B S CS(13) - 5.1.2-3 MT CS(13) - LPV 2 yr. - RC-408 Rapid Depressurization Valve GT EL 1 B S CS(13) - 5.1.2-3 MT CS(13) - LPV 2 yr. - RC-409 Rapid Deressurization Valve GT EL 1 B S CS(13) - 5.1.2-3 MT CS(13) - LPV 2 yr. - RC-410 Pressurizer Gas Vent Valve GL S I B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - LPV 2 yr. - Approved Design Meterle!- Desiges of SSC Page 3.9111
System 80 + _ Deshin contrar occument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (!) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. RC-411 Pressurizer Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - LPV 2 yr. - RC-412 Pressurizer Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - LPV 2 yr. - RC-413 Pressurizer Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - LPV 2 yr. - RC-414 Reactor Vessel Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(iQ - FS CS(14) - LPV 2 yr. RC-415 Reactor Vessel Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - l LPV 2 yr. - RC-416 Reactor Vessel Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - LPV 2 yr. - f RC-417 Reactor Vessel Gas Vent Valve GL S 1 B S CS(14) - 5.1.2-3 MT CS(14) - FS CS(14) - l LPV 2 yr. - I l Approved Design Material- Design of SSC Page 3.9-112 i t O O O
(m V System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) , l (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. RC-418 RCGV Discharge to Reactor Drain Tank GL S 2 B S 3 mo. - 5.1.2-3 MT 3 mo. - FS 3 mo. - LPV 2 yr. - RC-419 RCGV Discharge to IRWST GL S 2 B S 3 mo. - 5.1.2-3 MT 3 mo. - FS 3 mo. - LPV 2 yr. - RC-442 Pressurizer Spray Control Valve GT EL 1 B LPV 2 yr. - 5.1.2-3 RC-443 Pressurizer Spray Control Valve GT EL 1 B LPV 2 yr. - 5.1.2-3 RC-XXX SDS/ Safety Valve Sparger Line 1 Vacuum RV SA 2 C RVT 10 yr. - - Breaker RC-XXX SDS/ Safety Valve Sparger Line 2 Vacuum RV SA 2 C RVT 10 yr. - - Breaker SC-204 SG 1 Cold Leg Sample GL S 2 B CIN S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-211 SG 1 Ilot Ixg Sample GL S 2 B CIN S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-219 SG I Cold Leg Sample GL S 2 B CIN S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - l LT 2 yr.(45) - Approved Desigr* oeteterial- Desker of SSC Page 3.9-113
?'"80+--- - - - - - - - - - - - - --E*skn Cond Dwnnent Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.)
(a) I (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SC-220 SG I Downcomer Sample GL S 2 B CIN S 3 mo. - - hit 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) SC-221 SG 1 Downcomer Sample GL S 2 B CIN S 3 mo. - - h1T 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-222 SG 2 Cold Leg Sample GL S 2 B CIN S 3 mo. - - hit 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-223 SG 2 Cold leg Sample GL S 2 B CIN S 3 mo. - - hit 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-224 SG 2 Ilot 12g Sample GL S 2 B CIN S 3 mo. - - nit 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-225 SG 2 Hot Leg Sampic GL S 2 B CIN S 3 mo. - - hit 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - Approved Design Material- Design of SSC Page 3.9-114 O O O
eg. x ^%-
. bl .w) (G System 80+
. . _ D==& Control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.)
(a) (b; (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test lio. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SC-226 SG 2 Downcomer Sample GL S 2 B CIN S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-227 SG 2 Downcomer Sample GL S 2 B CIN S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SC-228 SG 1 Ilot Leg Sample GL S 2 B CIN S 3 mo. - - MT 3 mo. - FS 3 me. - LPV 2 yr. - LT 2 yr.(45) - SG-105 SG 1 ADV Isolation Valve GT EL 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - l LPV 2 yr. - LT 2 yr.(45) - SG-106 SG 1 ADV Isolation Valve GT EL 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SG-107 SG 2 ADV Isolation Valve GT EL 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - Approved Design Material- Design of SSC Po90 3.9-115
System 80+ . - . oesign controloccument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SG-108 SG 2 ADV isolation Valve GT EL 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SG-130 SG 1 Main FW Downcomer Isolation GT P 2 B CIN S CS(15) - 10.1-2 MT CS(15) - FS CS(15) - LPV 2 yr. - LT 2 yr.(45) 8 SG-132 SG 1 Main FW Economizer Isolation GT P 2 B CIN S CS(15) - 10.1-2 MT CS(15) - FS CS(15) - LPV 2 yr. - l LT 2 yr.(45) 8 SG-135 SG 2 Main FW Downcomer Isolation GT P 2 B CIN S CS(15) - 10.1-2 MT CS(15) - FS CS(15) - LPV 2 yr. - LT 2 yr.(45) 8 SG-137 SG 2 Main FW Economizer Isolation GT P 2 B CIN S CS(15) - 10.1-2 MT CS(15) - FS CS(15) - LPV 2 yr. - LT 2 yr.(45) 8 SG-140 SG 1 Main Steam Isolation GL P 2 B CIN S CS(16) - 10.1-2 MT CS(16) - FS CS(16) - LPV 2 yr. - LT 2 yr.(45) - Approved Design hinterial- Design of SSC Page 3.9116 O O O
EV****** 80 + . W W Documerrt i Tatde 3.9-15 Inservice Testing of Safety-Related Pumps _and Valves (Cont'd.) . . i , (k) (b) (c) (d) (e) - (f) _ (g) (i) Valve Valve Valve Valve Safety Code Vahe Test Test Test No.' Descdytion Type Act Class Cv. Funct Reqd Freg Config. Mg. No. SG-141 SG 2 Main Steam Isolation GL P 2 B CIN
- . S CS(16) -
10.1-2 MT CS(16). - l FS CS(16) . - i LPV 2 yr. - LT 2 yr.(45) - 1 SG-150 SG 1 Main Steam Isolation GL P 2 B CIN S CS(16) - 10.1-2 . . MT CS(16) - FS CS(16) - LPV 2 yr. - I LT - 2 yr.(45) - SG-151 SG 2 Main Steam Isolation GL P 2 B CIN S CS(16) - 10.1-2 MT_ CS(16) - FS CS(16) - LPV~ 2 yr. - LT 2 yr.(45) - 1 SG-168 SG 1 Main Steam Isolation Valve Bypass GT AD 2 B CIN S 3 mo. - 10.1-2 . MT 3 mo. - ! FS 3 me. - l LPV 2 yr. - LT 2 yr.(45) - i l SG.169 SG 1 Main Steam Isolation Valve Bypass GT AD 2 B CIN S 3 mo. - 10.1-2 ; MT 3 mo. - FS 3 mo. - I { LPV 2 yr. - l LT 2 yr.(45) - ! SG-172 SG 1 Main FW Downcomer Isolation GT P 2 B CIN S CS(15) - 10.1-2
- . MT CS(15) -
FS CS(15) - LPV 2 yr. - LT 2 yr.(45) 8 Approved Destpre nietaniel- D& of SSC P00* 3 9-11I l ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____ _ _ _ _____ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . . -_ _ _ _ _ . _ _ . . . . . . _ ~
System 80+ Design Control Document - Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (0 (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SG-174 SG 1 Main FW Economizer Isolation GT P 2 B CIN S CS(15) - 10.1-2 MT CS(15) - FS CS(15) - LPV 2 yr. - LT 2 yr.(45) 8 SG-175 SG 2 Main FW Downcomer Isolation GT P 2 B CIN S CS(15) - 10.1-2 MT CS(15) - l FS CS(15) - LPV 2 yr. - LT 2 yr.(45) 8 SG-177 SG 2 Main FW Economizer Isolation GT P 2 B CIN S CS(15) - 10.1-2 l MT CS(15) - [ FS CS(15) - l LPV 2 yr. - LT 2 yr.(45) 8 SG-178 SG 1 Atmospheric Dump Valve GL S 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SG.179 SG 2 Atmospheric Dump Valve GL S 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - FS 3 mo. - , LPV 2 yr. - l LT 2 yr.(45) - SG-182 SG 2 Main Steam Isolation Valve Bypass GT AD 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - Approved Design Material- Design of SSC Page 3.9-118 O O O
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System 80+ W ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SG-183 SG 2 Main Steam Isolation Valve Bypass GT AD 2 B CIN S 3 mo. - 10.1-2 , MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - 5G-184 SG 1 Atmospheric Dump Valve GL S 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SG-185 SG 2 Atmospheric Dump Valve GL S 2 B CIN S 3 mo. - 10.1-2 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr.(45) - SG-554 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-555 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-556 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-557 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-558 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-559 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-560 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - Approved Desigrr Meterial- Desiger of SSC Page 3.9-119
System 80+- - . - _ - - oesi.gn controloccument - Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SG-561 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-567 SG 1 Economizer Feedwater Line Check Valve CK SA 2 C RF CS(34) - 10.1-2 SG-572 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-573 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-574 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-575 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-576 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-577 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-578 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-579 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-598 SG 1 Economizer Feedwater Line Check Valve CK SA 2 C CIN RF CS(34) - 10.1-2 LT 2 yr.(45) 8 SG-599 SG 2 Economizer Feedwater Line Check Valve CK SA 2 C CIN RF CS(34) - 10.1-2 LT 2 yr.(45) 8 SG-612 SG 1 Economizer Feedwater Line Check Valve CK SA 2 C RF CS(34) - 10.1-2 SG-642 SG 1 Downcomer Feedwater Line Check Valve CK SA 2 C CIN RF CS(34) - 10.1-2 LT 2 yr.(45) 8 Approved Design Material- Design of SSC Page 3.9-120 0 0 9
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SG-643 SG 2 Downcomer Feedwater Line Check Valve CK SA 2 C CIN RF CS(34) - 10.1-2 LT 2 yr.(45) 8 SG-650 SG 1 Economizer Feedwater Line Check Valve CK SA 2 C RF CS(34) - 10.1-2 SG-651 SG 2 Economizer Feedwater Line Check Valve CK SA 2 C RF CS(34) - 10.1-2 SG-652 SG 1 Downcomer Feedwater Line Check Valve CK SA 2 C CIN RF CS(34) - 10.1-2 LT 2 yr.(45) 9 SG-653 SG 2 Downcomer Feedwater Line Check Valve CK SA 2 C CIN RF CS(34) - 10.1-2 LT 2 yr.(45) 9 SG-691 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-692 SG 1 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-694 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SG-695 SG 2 Main Steam Safety Valve RV SA 2 C CIN RVT 5 yr. - 10.1-2 LT 2 yr.(45) - SI-100 SIS Division 1 Miniflow-IRWST CK SA 2 C CIN S 3 mo. - 6.3.2-1 A RF 3 mo. - S1-101 SIS Division 2 Miniflow-IRWST CK SA 2 C CIN S 3 mo. - 6.3.2-1 B RF 3 mo. - SI-104 CS Pump #2 Suction Isolation GL M 2 B S 3 mo. - 6.3.2-1 B LPV 2 yr. - SI-105 CS Pump #1 Suction Isolation GL M 2 B S 3 mo. - 6.3.2-I A LPV 2 yr. - SI-l13 St Pump #4 Discharge CK SA 2 C S RO(18) 12 6.3.2-IC RF RO(18) - Approved Design Matwiel- Desigrs of SSC Page 3.9-121
System 80+ . - - - - - - _ - - . . Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) i l (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. , SI-123 SI Pump #2 Discharge CK SA 2 C S CS(19) 12 6.3.2-lC RF RO(19) - SI-133 St Pump #3 Discharge CK SA 2 C S RO(18) 12 6.3.2-IC RF RO(18) - SI-143 St Pump #1 Discharge CK SA 2 C S CS(19) 12 6.3.2-IC RF RO(19) - SI-157 CS Pump #1 Suction CK SA 2 C S 3 mo. - 6.3.2-1 A RF 3 mo. - SI-158 CS Pump #2 Suction CK SA 2 C S 3 mo. - 6.3.2-1 B RF 3 mo. - SI-161 CS Pump i Recirculation Line Relief RV SA 2 C RVT 10 yr. - 6.3.2-I A SI-164 CS Pump #1 Discharge CK SA 2 A/C CIC S RO(20) 12 6.3.2-lC LT 2 yr. 1 RF RO(20) - SI-165 CS Pump #2 Discharge CK SA 2 A/C CIC S RO(20) 12 6.3.2-IC LT 2 yr. 1 RF RO(20) - SI-166 St Hot Leg injection 2 Relief RV SA 2 C RVT 10 yr. - 6.3.2-IC SI-168 SCS Return Line 2 Check Valve CK SA 2 B S CS(19) - 6.3.2-IC l RF 3 mo. 9 l SI-169 SC Pump 2 Suction Relief RV SA 1 C RVT 5 yr. - 6.3.2-lC l l SI-178 SCS Return Line 1 Check Valve CK SA 2 B S CS(19) - 6.3.2-IC j RF 3 mo. 9 l SI-179 SC Pump 1 Suction Relief RV SA 2 A/C CIC RVT 10 yr. - 6.3.2-IC LT 2 yr. 4 SI-187 SC Line 1 Recirculation to IRWST Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 A Approved Design Afsterial- Design of SSC Page 3.9-122 O O O
O O O System 80+ Desen contrar Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. 51-188 SC Line 2 Recirculation to IRWST Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B SI-189 SC Pump 2 Suction Relief RV SA 2. A/C CIC RVT 10 yr. - 6.3.2-IC LT 2 yr. 4 SI-191 CS HX 2 Outlet Relief RV SA 2 C RVT 10 yr. - 6.3.2-1B SI-193 CS Pump 2 Recirculation Line Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B SI-194 CS HX 1 Outlet Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 A SI-211 SIT 4 Safety Valve RV SA 2 C RVT 10 yr. - 6.3.2-IC SI-215 SI Tank 4 Discharge CK SA 1 A/C PlV S RR(21) 14 6.3.2-IC LT 2 yr. 10 RF (44) - SI-217 DVI Nozzle 2B CK SA 1 A/C PlV S RR(22) 12 6.3.2-IC LT 2 yr. 10 RF (44) - SI-221 SIT 2 Safety Valve RV SA 2 C RVT 10 yr. - 6.3.2-IC SI-225 Si Tank 2 Discharge CK SA 1 A/C P!V S RR(21) 14 6.3.2-IC LT 2 yr. 10 RF (44) - SI-227 DVI Nozzle IB CK SA I A/C PlV S CS(22) 12 6.3.2-I C LT 2 yr. 10 j RF (44) - ! SI-231 SIT 3 Safety Valve RV SA 2 C RVT 10 yr. - 6.3.2-lC l SI-235 SI Tank 3 Discharge CK SA 1 A/C PlV S RR(21) 14 6.3.2-lC l LT 2 yr. 10 i RF- (44) - SI-237 DVI Nozzle 2A CK SA 1 A/C PlV S RR(22) 12 6.3.2-IC l LT 2 yr. 10 RF (44) - Approved Design Meterial- Design of ESC Page 3.9-123
System 80+ Design Control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Vahe Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Hg. No. SI-241 SIT 1 Safety Valve RV SA 2 C RVT 10 yr. - 6.3.2-!C SI-245 SI Tank i Discharge CK SA 1 A/C PlV S RR(21) 14 6.3.2-lC LT 2 yr. 10 RF (44) - SI-247 DVI Nozzle 1 A CK SA 1 A/C PlV S CS(22) 12 6.3.2-I C LT 2 yr. 10 RF (44) 1 SI-285 St Pump 1/3 Recirculation Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 A SI-286 St Pump 2/4 Recirculation Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B SI-287 SC Line 2 Recirculation to IRWST Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B SI-289 SC Line i Recirculation to IRWST Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 A SI-292 EDT/ Safety Injection Tank Relief RV SA 3 C RVT 10 yr. - 6.3.2-1B SI-293 SIT Fill Line Isolation GL M 2 A P,CIC LT 2 yr. 4 6.3.2-1 B SI-300 Division i SC/CS Pump Test Line GT EL 2 B CIN S 3 mo. - 6.3.2-1 A MT 3 mo. - LPV 2 yr. - SI-301 Division 2 SC/CS Pump Test Line GT EL 2 B CIN S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - SI-302 St Division 1 Miniflow Isolation GT EL 2 B CIN S RO(23) - 6.3.2-I A MT RO(23) - LPV 2 yr. - SI-303 SI Division 2 Miniflow Isolation GT EL 2 B CIN S RO(23) - 6.3.2-1 B MT RO(23) - LPV 2 yr. - SI-304 CS & St Pump 1 Sctn Isolation GT EL 2 B CIN S 3 mo. - 6.3.2-1 A MT 3 mo. - LPV 2 yr. - Approved Design Material- Design of SSC Page 3.9-124 O O O
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( System 80+ Deaien Conend Document , Tatde 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) '~ (a) (b) (c) (d) (e) (f) (g) - (i) Valve Valve Valve Valve Safety Code Valve Test - Test Test No. Description Type Act Class Cat Funct Reqd Freg Config. Hg. No. e. , S1-305 CS & S1 Pump 2 Sctn Isolation GT EL 2 B CIN S 3 mo. - 6.3.2-1B ) MT 3 mo. - LPV 2 yr. - i SI-308 S1 Pump 3 Sctn Isolation GT EL 2 B CIN S 3 mo. - 6.3.2-1 A 4 MT 3 mo. - ! LPV 2 yr. - SI-309 SI Pump 4 Sctn Isolation GT EL 2 B CIN S 3 mo. - 6.3.2-1 B '! MT 3 mo. -- I LPV 2 yr. - 1 SI-310 - SC HX #1 Outlet Isolation GL EL- 2 B S 3 mo. - 6.3.2-I A MT 3 mo. - LPV 2 yr. - l SI-311 SC HX #2 Outlet Isolation GL EL 2 B S 3 mo. - 6.3.2-1 B
, MT 3 mo. -
LPV 2 yr. - j SI-312 SC HX #1 Bypass GL EL 2 B S 3 mo. - 6.3.2-I A l MT 3 mo. - ! LPV 2 yr. - SI-313 SC HX #2 Bypass GL EL 2 B S 3 mo. - 6.3.2-1B MT 3 mo. - ! LPV 2 yr. - SI-314 SCS Division i 1RWST Recirculation Control GL EL 2 B S 3 mo. - 6.3.2-I A MT 3 mo. - 1 LPV 2 yr. - i SI-315 SCS Division 2 IRWST Recirculation Control GL EL 2 A S 3 mo. - 6.3.2-1 B MT 3 mo. - , LPV 2 yr. - Appred Desy nieterW- Design of SSC Page 3.9-125 f
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System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (c) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. F1g. No. 51-321 Ilot Leg injection LP 1 Isolation GL EL 2 B CI S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - LT 2 yr. I SI-322 Ilot Leg injection I Bleed Off isolation GL AD 1 A PlV S 3 mo. - 6.3.2-IC i MT 3 mo. - FS RO(42) - LPV 2 yr. - l LT 2 yr. 10 S1-331 Ilot Leg Injection LP 2 Isolation GL EL 2 A CIC S 3 mo. . 6.3.2-IC 1 MT 3 mo. - LPV 2 yr. - LT 2 yr. I SI-332 Ilot leg injection 2 Bleed Off Isolation GL AD 1 A PlV S 3 mo. - 6.3.2-IC 1 MT 3 mo. - l FS RO(42) - LPV 2 yr. - LT 2 yr. 10 l SI-340 Division i CS/SCS Suction Crossover GT EL 2 B S 3 mo. - 6.3.2-I A MT 3 mo. - LPV 2 yr. - l SI-341 Division 1 CS/SCS Discharge Crossover GT EL 2 B S 3 mo. - 6.3.2-1 A MT 3 mo. - LPV 2 yr. - SI-342 Division 2 CS/SCS Suction Crossover GT EL 2 B S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - Approved Design Material- Design of SSC Page 2.9-t26 O O O
L/ G %) System 80+ _ _ _ _ _ _ _ _ _ Design Control Docummt Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. 51-343 Division 2 CS/SCS Discharge Crossover GT EL 2 B S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - SI-390 IIVT Spillway Isolation GT EL 2 B S RO(40) - 6.8-4 MT RO(40) - LPV 2 yr. - SI-391 IIVT Spillway Isolation GT EL 2 B S RO(40) - 6.8-4 MT RO(40) - LPV 2 yr. - SI-392 IIVT Spillway Isolation GT EL 2 B S RO(40) - 6.8-4 MT RO(40) - LPV 2 yr. - SI-393 IIVT Spillway Isolation GT EL 2 B S RO(40) - 6.8-4 MT RO(40) - LPV 2 yr. - SI-394 Reactor Cavity Spillway Isolation GT EL 2 B S RO(40) - 6.8-4 MT RO(40) - LPV 2 yr. - SI-395 Reactor Cavity Spillway Isolation GT EL 2 B S RO(40) - 6.8-4 MT RO(40) - LPV 2 yr. - SI-3 % SI IRWST Boron Recovery Supply to CYCS GT EL 2 A CIC S 3 mo. - 6.8-3 Isolation MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 SI-397 SI IRWST Boron Recovery Supply to CVCS GT EL 2 A CIC S 3 mo. - 6.8-3 Isolation MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 Approved Design Material- Design of SSC Page 3.9-127
System 80+ Desi.gn ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SI404 SI Pump #1 Discharge CK SA 2 C S RO(18) - 6.3.2-1 A RF 3 mo.(18A) 9 SI-405 SI Pump #2 Discharge CK SA 2 C S RO(18) - 6.3.2-1 B RF 3 mo.(18A) 9 SI-409 SI Pump 2 Discharge Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B SI-417 SI Pump 1 Discharge Relief RV SA 2 C RVT 10 yr. - 6.3.2-I A SI-422 SC IlX 1 Outlet Relief RV SA 2 C RVT 10 yr. - 6.3.2-I A SI-423 SC llX 2 Outlet Relief RV SA 2 C RVT 10 yr. - 6.3.2-1B SI-424 St Pump #1 Miniflow CK SA 2 C S 3 mo. 11 6.3.2-1 A RF 3 mo.(25) 9 SI-426 St Pump #2 Miniflow CK SA 2 C S 3 mo. I1 6.3.2-1 B RF 3 mo.(25) 9 SI-434 SI Pump #3 Discharge CK SA 2 C S RO(18) - 6.3.2-1 A RF 3 mo.(18A) 9 SI-439 SI Pump 3 Discharge Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 A SI.446 SI Pump #4 Discharge CK SA 2 C S RO(18) - 6.3.2-1B RF 3 mo.(18A) 9 SI-448 St Pump #4 Miniflow CK SA 2 C S 3 mo. I1 6.3.2-1 B RF 3 mo.(25) 9 SI-449 St Pump 4 Discharge Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B SI-451 SI Pump #3 Miniflow CK SA 2 C S 3 mo. 11 6.3.2-I A [ RF 3 mo.(25) 9 l SI-466 SC Return Line i Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 A I SI-467 SC Return Line 2 Relief RV SA 2 C RVT 10 yr. - 6.3.2-1 B 1 l SI-468 SI Ilot Leg Injection i Relief RV SA 2 C RVT 10 yr. - 6.3.2-I C l SI469 SC Pump 1 Suction Relief RV SA 1 C RVT 5 yr. - 6.3.2-IC i Approved Design Material- Design of SSC Page 3.9-128 l O O O
O O O , System 80+ _ _ _ _ _ Design Control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Mg. No. SI-473 SIT Drain Line Relief RV SA 2 C RVT 10 yr. - 6.3.2-IC SI-474 SIT Fill Line Relief RV SA 2 A/C CIC RVT 10 yr. - 6.3.2-IC LT 2 yr. 4 SI-484 CS Pump #2 Discharge CK SA 2 C S 3 mo. 12 6.3.2-1 B RF 3 mo.(17) 9 SI-485 CS Pump #1 Discharge CK SA 2 C S 3 mo. 12 6.3.2-I A RF 3 mo.(17) 9 SI-522 Hot txg injection LPl CK SA 1 A/C PlV S RO(18) 12 6.3.2-lC LT 2 yr. 10 RF (44) - SI-523 Hot Leg Injection LP1 CK SA 1 A/C CIC, S RO(18) 12 6.3.2-1C PIV LT 2 yr. 1 RF (44) - SI-532 Hot leg injection LP2 CK SA 1 A/C PlV S RO(18) 12 6.3.2-lC LT 2 yr. 10 RF (44) - SI-533 Hot Leg Injection LP2 CK SA 1 A/C CIC, S RO(18) 12 6.3.2-IC PlV LT 2 yr. 1 RF (44) - SI-540 SI Pump #4 Discharge CK SA 1 A/C PlV S RO(IL) 12 6.3.2-lC LT 2 yr. 10 RF (44) - SI-541 S1 Pump #2 Discharge CK SA 1 A/C PlV S CS(19) 12 6.3.2-IC LT 2 yr. 10 , RF (44) - SI-542 St Pump #3 Discharge CK SA. 1 A/C PlV S RO(18) 12 6.3.2-IC LT 2 yr. 10 RF (44) - Approve <9 Design neatoriel- Design of SSC Page 3.9-129
System 80+ Design Control Document
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Table 3.9-15 Inservice Testing of Safety-Rela (ed Pumps and Valves (Cont'd.) (a) (b) (c) (d) (c) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. S1-543 SI Pump #1 Discharge CK SA 1 A/C PlV S CS(19) 12 6.3.2-lC LT 2 yr. 10 RF (44) - SI-568 SC Pump #1 Discharge CK SA 2 C S 3 mo. I1 6.3.2-I A RF 3 mo.(17) 9 SI-569 SC Pump #2 Discharge CK SA 2 C S 3 mo. 11 6.3.2-1 B RF 3 mo.(17) 9 SI-600 SCS Train 2 Discharge GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-601 SCS Train 1 Discharge GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-602 SI Line 2 Throttle GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-603 SI Line 1 Throttle GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-604 Ilot Leg injection LPI Isolation GT EL 2 B S 3 mo. - 6.3.2-I A i MT 3 mo. - l LPV 2 yr. - SI-605 SIT 4 Atmospheric Vent Isolation GL S 2 B S CS(33) - 6.3.2-IC MT CS(33) - FS CS(33) - LPV 2 yr. - Approved Design Material- Design of SSC Page 3.9-130 0 0 - 0 -
- System 80 + o-& contrar Document Table 3.9-15 Imervice Testing of Safety-Related Pumps and Valves (Cont'd.)
(a) (b) (c) (d) (c) (f) (g) (i)
- Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No.
51-606 SIT 2 Atmospheric Vent Isolation GL S 2 B S CS(33) - 6.3.2-IC MT CS(33) - FS CS(33) - LPV 2 yr. - SI-607 SIT 3 Atmospheric Vent Isolation GL S 2 B S CS(33) - 6.3.2-IC MT CS(33) - FS CS(33) - LPV 2 yr. -
- SI-608 SIT I Atmospheric Vent Isolation GL S 2 B S CS(33) -
6.3.2-lC MT CS(33) - FS CS(33) - LPV 2 yr. - SI-609 Ilot Leg Injection LP2 Isolation GT EL 2 B S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - S1-611 SIT #4 Fill Line GL AD 2 B S 3 mo. - 6.3.2-lC MT 3 mo. - FS RO(42) - LPV 2 yr. - SI-612 SIT 4 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-IC , FS RO(41) - LPV 2 yr. - St.613 SIT 4 Atmospheric Vent Isolation GL S 2 B S CS(33) - 6.3.2-lC MT CS(33) - FS CS(33) - LPV 2 yr. - SI-614 SIT #4 Discharge Isolation GT EL 1 B S CS(28) - 6.3.2-IC MT CS(28) - LPV 2 yr. - l l Approveet Destyrs Material- Design of SSC P*9* 3 9-13I
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System 80+ oesign control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. 51-616 SI Pump #4 Discharge GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-618 St Line 4 Leakage Return GL AD 1 A PlV S 3 mo. - 6.3.2-lC MT 3 mo. - FS RO(42) - LPV 2 yr. - LT 2 yr. 10 51-619 SIT 4 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-IC FS RO(41) - LPV 2 yr. - l 51-621 SIT #2 Fill Line GL AD 2 B S 3 mo. - 6.3.2-IC MT 3 mo. - FS RO(42) - LPV 2 yr. - ! SI-622 SIT 2 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-IC FS RO(41) - LPV 2 yr. - SI-623 SIT 2 Atmospheric Vent isolation GL S 2 B S CS(33) - 6.3.2-IC MT CS(33) - FS CS(33) - LPV 2 yr. - SI-624 SIT #2 Discharge Isolation GT EL 1 B S CS(28) - 6.3.2-IC MT CS(28) - LPV 2 yr. - SI-626 St Pump #2 Discharge GL EL 2 B' CIN S 3 mo. - 6.3.2-lC MT 3 mo. - LPV 2 yr. - Approved Design Materist - Design of SSC Page 3.9-132 O O O
a v v System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (c) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SI-628 SI Line 2 Leakage Return GL AD 1 A PIV S 3 mo. - 6.3.2-IC MT 3 mo. - FS RO(42) - LPV 2 yr. - LT 2 yr. 10 SI-629 SIT 2 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-IC FS RO(41) - LPV 2 yr. - SI-631 SIT #3 Fill Line GL AD 2 B S 3 mo. - 6.3.2-lC MT 3 mo. - FS RO(42) - LPV 2 yr. - SI-632 SIT 3 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-lC FS RO(41) - LPV 2 yr. - SI-633 SIT 3 Atmospheric Vent isolation GL S 2 B S CS(33) - 6.3.2-IC MT CS(33) - FS CS(33) - LPV 2 yr. - S1 634 SIT #3 Discharge Isolation GT EL 1 B S CS(28) - 6.3.2-I C MT CS(28) - LPV 2 yr. - 51-636 S1 Pump #3 Discharge GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-638 St Line 3 Leakage Return GL AD 1 A PlV S 3 mo. - 6.3.2-IC MT 3 mo. - FS RO(42) - LPV 2 yr. - LT 2 yr. 10 Approwd Design Meterial- Design of SSC P*0* 3 9-133
-----_-a- - . ._ _ . _- - - . --------___..----..----.__.--.---_._-_-__--__.._.--.----__.--_-._-__--.__.-__-____-__._.____-----.----_m- - .<.=e-_e- _.-_ --- - -e 4 -e- ., w+~ , v-
System 80+ Design ControlDocument l Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) l (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. 51-639 SIT 3 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-IC FS RO(41) - LPV 2 yr. - S1-641 SIT #1 Fill Line GL AD 1 B S 3 mo. - 6.3.2-lC MT 3 mo. - FS RO(42) - LPV 2 yr. - SI-642 SIT I Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-IC FS RO(41) - LPV 2 yr. - SI-643 SIT I Atmospheric Vent Isolation GL S 2 B S CS(33) - 6.3.2-lC MT CS(33) - FS CS(33) - LPV 2 yr. - SI 644 SIT #1 Discharge Isolation GT EL 1 B S CS(28) - 6.3.2-IC MT CS(28) - LPV 2 yr. - SI-646 S1 Pump #1 Discharge GL EL 2 B CIN S 3 mo. - 6.3.2-IC MT 3 mo. - LPV 2 yr. - SI-648 51 Line 1 Leakage Return GL AD 1 A PlV S 3 mo. - 6.3.2-lC MT 3 mo. - FS RO(42) - LPV 2 yr. - LT 2 yr. 10 SI-649 SIT 1 Nitrogen Pressure Control GL AD 2 B S El(41) - 6.3.2-lC FS RO(41) - LPV 2 yr. - Approved Desigts Material- Dessgrr of SSC Page 3.9-134 O O O
O O O System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat funct Reqd Freq Config. Fig. No. 51-651 SC Pump #1 Suetin GT EL 1 A PlV S CS(29) - 6.3.2-IC MT CS(29) - LPV 2 yr. - LT 2 yr. 10 SI-652 SC Pump #2 Suction GT EL i A PIV S CS(29) - 6.3.2-lC MT CS(29) - LPV 2 yr. - LT 2 yr. 10 SI-653 SC Pump #1 Suction GT EL 1 A PlV,CI S CS(29) - 6.3.2-I C C MT CS(29) . - LPV 2 yr. - LT 2 yr. 10 LT 2 yr. .4 SI-654 SC Pump #2 Suction GT EL 1 A PlV,Cl S CS(.'91 - 6.3.2-IC C MT CS(29) - LPV 2 yr. - LT 2 yr. 10 LT 2 yr. 4 SI-655 SC Pump #1 Suction GT EL 2 B CIC S CS(29) - 6.3.2-IC MT CS(29) - LPV 2 yr. - LT 2 yr. 4
- SI-656 SC Pump #2 Suction GT EL 2 B CIC S CS(29) -
6.3.2-I C MT CS(29) - LPV 2 yr. - LT 2 yr. 4 SI-657 CSS Division 1 IRWST Recirculation Control GL EL 2 B S 3 mo. - 6.3.2-1 A MT 3 mo. - LPV 2 yr. - Approved Design ntatorial- Design of SSC . Page 3.9-135
System 80+ Design control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (0 (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SI-658 CSS Division 2 IRWST Recirculation Control GL EL 2 B S 3 mo. - 6.3.2-IB MT 3 mo. - LPV 2 yr. - SI-661 SIT Drain to RDT Isolation GL AD 2 B S 3 mo. - 6.3.2-IC FS RO(42) - LPV 2 yr. - SI-670 SIT Drain to IRWST Isolation GL AD 2 B S 3 mo. - 6.3.2-I C FS RO(42) - LPV 2 yr. - SI-671 CS Pump #2 Discharge GT EL 2 A CIC S 3 mo. - 6.3.2-I B MT 3 mo. - LPV 2 yr. - LT 2 yr. I SI-672 CS Pump #1 Discharge GT EL 2 A CIC S 3 mo. - 6.3.2-I A MT 3 mo. - LPV 2 yr. - LT 2 yr. I SI-682 SIT Fill Line Containment Isolation GL AD 2 A CIC S 3 mo. - 6.3.2-I C MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SI-686 CS IIX 1 to IRWST Isolation GT EL 2 B S 3 mo. - 6.3.2-I A MT 3 mo. - LPV 2 yr. - 51-687 Division 1 CS Ileader Block Isolation GT EL 2 3 S 3 mo. - 6.3.2-I A MT 3 mo. - LPV 2 yr. - Approved Design Materls!- Destyn of SSC Page 3.9-136 O O O
g s s U V v System 80+-- Design Control Document Tahic 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Hg. No. SI-688 SCS Division 1 1RWST Recirculation isolation GT EL 2 B S 3 mo. - 6.3.2-I A MT 3 mo. - LPV 2 yr. - SI-693 SCS Division 2 IRWST Recirculation isolation GT EL 2 B S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - 51-695 Division 2 CS Ileader Block Isolation GT EL 2 B S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - SI-6% CS IlX 2 to IRWST Isolation GT EL 2 B S 3 mo. - 6.3.2-1 B MT 3 mo. - LPV 2 yr. - SS-200 llot Ixg Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SS-201 Pressurizer Liquid Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SS-202 Pressurizer Steam Space Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 Approved Design Meterial- Design of SSC Page 3.9-137
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (c) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SS-203 flot Leg Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 3T. 4 SS-204 Pressurizer Liquid Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SS-205 Pressurizer Steam Space Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SS-208 lloldup Volume Tank Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - i FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SS-210 lloidup Volume Tank Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 SS-211 Iloidup Volume Tank Sample GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 Approved Design Msterial- Design of SSC Page 3.9-138 O O O
x r^s (vI U svaran eg+_ __ _ _ wn" co""* -t Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig No. , SW-100 SSW Strainer I A Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-101 SSW Strainer IB Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-102 SSW Strainer I A Backwash PL EL 3 B S 3 mo. - 9.2.1-1. Sh 1 MT 3 mo. - LPV 2 yr. - SW-103 SSW Strainer IB Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-104 SSW Strainer IA Backwash PL EL 3 B S 3 mo. - 9.2.1-1. Sh 1 MT 3 mo. - LPV 2 yr. - SW-105 SSW Strainer IB Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-106 SSW Strainer l A Backwash PL EL 3 B S 3 mo. - 9.2.1-1. Sh 1 MT 3 mo. - LPV 2 yr. - SW-107 SSW Strainer IB Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-108 SSW Strainer I A Backwash PL EL 3 B S 3 mo. - 9.2.1-1. Sh 1 MT 3 mo. - LPV 2 yr. - Approveef Design Material- Design of SSC Page 3.9-139
System 80 + oesign contrar Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SW-109 SSW Strainer IB Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-110 SSW Strainer I A Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 1 MT 3 mo. - LPV 2 yr. - SW-Ill SSW Strainer IB Backwash PL EL 3 B S 3 mo. - 9.2.1-1 Sh 1 MT 3 mo. - LPV 2 yr. - SW-120 CCW IlX 1 A Inlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 2 MT 3 mo. - LPV 2 yr. - SW-121 CCW IIX IB Inlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 2 MT 3 mo. - LPV 2 yr. - SW-122 CCW IlX 1 A Outlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 2 MT 3 mo. - LPV 2 yr. - SW-123 CCW HX IB Outlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1. Sh 2 MT 3 mo. - LPV 2 yr. - SW-1302 SSW Pump 1 A Discharge CK SA 3 C S 3 mo. 11 9.2.1-1, Sh 1 RF 3 mo. 9 SW-1303 SSW Pump IB Discharge CK SA 3 C S 3 mo. 11 9.2.1-1, Sh 1 RF 3 mo. 9 SW-1350 CCW IIX 1 A Ileader Relief RV SA 3 C RVT 10 yr. - 9.2.1-1 Sh 2 SW-1351 CCW llX IB lleader Relief RV SA 3 C RVT 10 yr. - 9.2.1-1, Sh 2 Approved Design Materist - Design of SSC Page 3.9-140 0 - 0 0 -
System 80+ __ Desien w oocannent j Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) . , (a) ' (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test ; No. Description Type Act Class Cat 1%ect Reqd Freq Config. Hg. No. ;
- SW-200 SSW Strainer 2A Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3
+ MT 3 mo. - -! LPV 2 yr. - SW-201 SSW Strainer 2B Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 , MT 3 mo. -
- j. LPV 2 yr. -
f SW-202 SSW Strainer 2A Backwash PL- EL 3 B S 3 me. - 9.2.1-1, Sh 3 MT 3 mo. - . j LPV 2 yr. - SW-203 ~ SSW Strainer 2B Backwash PL EL 3 B S 3 mo.- - 9.2.1-1, Sh 3 : MT 3 mo. - LPV 2 yr. - SW-204 SSW Strainer 2A Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 ! MT 3 mo. LPV 2 yr. - L SW-205 SSW Strainer 2B Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 , MT 3 mo. - LPV 2 yr. -
)
SW-2% SSW Strainer 2A Backwash PL EL 3 B S 3 mo. - 9.2.1-1 Sh 3 ; MT 3 m. - [ LPV 2 yr. - SW-207 SSW Strainer 2B Backwash PL EL 3 B S 3 mo. - 9.2.1-1 Sh 3 j MT 3 mo. - LPV 2 yr. - SW-208 SSW Strainer 2A Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 l MT 3 mo. - LPV' 2 yr. - .i l 1 6 Apprend Duinn nietwM- Des > of SSC Pese 3.9-141 1
System 80 +_____ _ _ - _ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. SW-209 SSW Strainer 2B Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 MT 3 nm. - LPV 2 yr. - SW-210 SSW Strainer 2A Bxkwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 MT 3 mo. - LPV 2 yr. - SW-211 SSW Strainer 2B Backwash PL EL 3 B S 3 mo. - 9.2.1-1, Sh 3 MT 3 mo. - LPV 2 yr. - SW-220 CCW llX 2A Inlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 4 MT 3 mo. - LPV 2 yr. - SW-221 CCW llX 2B Inlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 4 MT 3 mo. - LPV 2 yr. - SW-222 CCW llX 2A Outlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 4 MT 3 mo. - LPV 2 yr. - SW-223 CCW IIX 2B Outlet Isolation BF EL 3 B S 3 mo. - 9.2.1-1, Sh 4 MT 3 mo. - LPV 2 yr. - [ SW-2302 SSW Pump 2A Discharge CK SA 3 C S 3 mo. I1 9.2.1-1, Sh 3 ( RF 3 mo. 9 SW-2303 SSW Pump 2B Discharge CK SA 3 C S 3 mo. 11 9.2.1-1, Sh 3 RF 3 mo. 9 , SW-2350 CCW HX 2A IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.1-1, Sh 4 i i SW-2331 CCW 11X 2B IIcader Relief RV SA 3 C RVT 10 yr. - 9.2.1-1, Sh 4 I i { Approved Design Material- Design of SSC Page 3.9-142 t O O O
p
.I.e3 [m v< tJ s -).
psp _Q+ W NW ht Tabte 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fgt. No. XX-001 Breathing Air Supply GT EL 2 A CIC S 3 mo. - -~ MT 3 mo. - LPV 2 yr. - LT 2 yr. 1 XX-002 Breathing Air Supply CK SA 2 A/C CIC S 3 mo. 14 - LT 2 yr. I RF 3 mo. - XX4)03 Station Air Supply GT EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - LT 2 yr. 1 XX-004 Station Air Supply CK SA 2 A/C CIC S 3 mo. 14 - LT 2 yr. 1 RF 3 mo. - XX-005 Division i Instrumentation Air Supply GT EL 2 A CIC S 3 mo. - 9.3.1-1, Sh 2 MT 3 mo. - LPV 2 yr. - LT 2 yr. 1 XX-006 Division 1 Instrumentation Air Supply CK SA 2 A/C CIC S 3 mo. 14 ' 9.3.1-1, Sh 2 LT 2 yr. I RF 3 mo. - XX-007 Division 2 Instrumentation Air Supply GT EL 2 A CIC S 3 mo. - 9.3.1-1, Sh 2 MT 3 mo. - LPV 2 vr. - LT 2., t XX-008 Division 2 Instrumentation Air Supply CK SA 2 A/C CIC S 3 mo. 14 9.3.1-1 Sh 2 LT 2 yr. 1 RF 3 mo. - Approved Deslyn Material- Design of SSC Page 3 9-143
System 80+ _ ___ _ _ _ _ _ _ _ oesign contrar Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (I) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-010 !!igh Volume Containment Purge Supply 1 BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - LPV 2 yr. - LT 2 yr. 5 XX-011 High Volume Containment Purge Supply 1 BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - LPV 2 yr. - LT 2 yr. 5 XX-012 liigh Volume Containment Purge Supply 2 BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - LPV 2 yr. - LT 2 yr. 5 XX-013 liigh Volume Containment Purge Supply 2 BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - ! LPV 2 yr. - LT 2 yr. 5 l XX-014 fligh Volume Containment Purge Exhaust i BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - LPV 2 yr. - LT 2 yr. 5 l XX-015 liigh Volume Containment Purge Exhaust i BF AD 2 A CIC S CS(30) - 9.4.6
- MT CS(30) -
1 FS CSGG) - LPV 2 yr. - LT l2 yr. 5 Approved Design Material- Design of SSC Page 3.9-144 O O O
A._/ s v System 80+ Deshyn ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX 016 liigh Volume Containment Purge Exhaust 2 BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - LPV 2 yr. - LT 2 yr. 5 , XX-017 Ifigh Volume Containment Purge Exhaust 2 BF AD 2 A CIC S CS(30) - 9.4.6 MT CS(30) - FS CS(30) - LPV 2 yr. - LT 2 yr. S XX-018 IAw Volume Containment Purge Supply BF AD 2 A CIC S 3 mo. - 9.4.6 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. I XX-019 Imw Volume Containment Purge Supply CK SA 2 A/C CIC S 3 mo. 12 9.4.6 LT 2 yr. 1 RF 3 mo. - XX-020 tow Volume Containment Purge Exhaust BF AD 2 A CIC S 3 mo. - 9.4.6 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4 XX-021 Low Volume Containment Purge Exhaust BF AD 2 A CIC S 3 mo. - 9.4.6 MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4
#pproved Design Material- Design of SSC Page 3.9145
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-030 SG I Combined Blowdown GT EL 2 B CIN S 3 mo. - 10.4.8-1 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - XX-031 SG 1 Combined Blowdown GT EL 2 B CIN S 3 mo. - 10.4.8-1 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - XX-032 SG 1 Combined Blowdown CK SA 2 C CIN S CS(26) - 10.4.8-1 RF 3 mo. - LT 2 yr.(45) - XX-033 SG 2 Combined Blowdown GT EL 2 B CIN S 3 mo. - 10.4.8-1 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - XX-034 SG 2 Combined Blowdown GT EL 2 B CIN S 3 mo. - 10.4.8-1 MT 3 mo. - LPV 2 yr. - LT 2 yr.(45) - XX-035 SG 2 Combined Blowdown CK SA 2 C CIN S CS(26) - 10.4.8-1 RF 3 mo. - LT 2 yr.(45) - l XX-040 Fire Water Supply 1 GT EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - LT 2 yr. 1 XX-041 Fire Water Supply 1 CK SA 2 A/C CIC LT 2 yr. 1 - RF RO(36) - l Approved Design Material- Design of SSC Page 3.9-146 O O O
. m - . - . . . _ . .._ _ . . _ - _. - .. -_ _ _ _. . .. - _ . ._ . _. _ . _ . . . _ . . . ._.
I w' SVstern 80+ Design Coned Docenment Table 3.9-15 la vke Testing of Safety-Related Pumps and Valves (Cont'd.)
- j (a) (b) (c) (d) - (e) - (f) - (g) : (i) -
~ Valve : Valve Valve Valve Safety Code Valve Test : Test : . Test .
No.- Description Type Act Class Cat Funct Reqd Freq Config. 71g. No. 4 XX4)42 Fire Water Supply 2 GT EL 2 A CIC S- 3 mo. - - - MT 3 mo - -- j LPV 2 yr. - , LT 2 yr. 1' ' XX-043 Fire Water Supply 2 CK SA 2 A/C CIC LT 2 yr. .1 - i RF RO(36) .- .t XX-050 - Containment Radiation Monitor Sampic Inlet - GL S 2 A CIC S 3 mo. - - MT 3 mo. - FS 3 mo. - LPV 2 yr. - LT 2 yr. 4~ XX-051 Containment Radiation Monitor Sample Inlet GL S 2 A CIC S 3 me. - - MT 3 mo. - t , FS 3 mo. - l l LPV 2 yr. - l^^ LT 2 yr. 4 XX-052 Containment Radiation Monitor Sampic Outlet GL S 2 A CIC S 3 mo. - - I' MT 3 mo. - , FS 3 mo. - LPV 2 yr. - LT 2 yr. 4-l XX-053 Containment Radiation Monitor Sample Outlet GL S 2 A- CIC S 3 me. - - l MT 3 mo. - FS 3 me. - l LPV 2 yr. - LT 2 yr. 4 XX-060. ILRT Pressure Sensing GL M 2 A P,CIC LT 2 yr. 4- - t XX-061 - ILRT Pressure Sensing GL M 2 A P,CIC LT 2 yr. 4 - 1 AppremiDes/pr Afsferief-Des 4pn of SSC Page 3.9-747 i
. . _ _ _ -. ._ _ _ _ _ _- - _ - _ __ _ . . . _ . . _ . ~ - , _ _ _ . - - . _ , _ . - . _ - - _ ~ -. .. ___..-._ __.._-... ..-. _ ,_.....-.. . . _ _
l System 80+ Design ControlDocument l Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) l (a) (b) (c) (d) (e) (f) (g) (i) Va!ic Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. l XX-070 Demineralized Water Supply GT EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - LT 2 yr. - XX-071 Demineralized Water Supply CK SA 2 A/C CIC S RO(37) 15 - LT 2 yr. 1 RF RO(37) - XX-080 Nitrogen Supply GT EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - l LT 2 yr. 1 XX-081 Nitrogen Supply CK SA 2 A/C CIC S 3 mo. 14 - LT 2 yr. I RF 3 mo. - XX-090 ILRT Pressurization Line GL M 2 A P,CIC LT 2 yr. 6 - XX-100 RCP Oil Fill Line GT EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 XX-101 RCP Oil Fill Line GT EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 XX-110 Containment Sump Pumps Discharge GT AD 2 A CIC S 3 mo. - 9.3.3-1 MT 3 mo. - LPV 2 yr. - LT 2 yr. 3 Approved Design Material- Design of SSC Page 3.9-148 O O O
. . . . . _ - .. -.. . - . . - - -. - . _ . . . .~ . . .. - , .- . . . . -
y Sv* tem 80+ Design ConeelDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) ! (a) (b) (c) - (d) (e) (f) (g) (i) ! Valve Valve Valve Valve Safety Code Valve Test - Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Hg. No. XX-Ill Containment Sump Pumps Discharge GT AD 2 A CIC S 3 mo. - 9.3.3-1 l MT 3 mo. - 1 LPV 2 yr. -
'j LT 2 yr. 3 ;
XX-120 Reactor Drain Tank Gas Space to GWMS GL EL. 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 , XX-121 Reactor Drain Tank Gas Space to GWMS GL EL 2 A CIC S 3 mo. - - MT 3 mo. - LPV 2 yr. - [ LT 2 yr. 4 XX-130 Decontamination Line GL M 2 A P.CIC LT 2 yr. 4 - XX-131 Decontamination Line GL M '2 A P,CIC LT 2 yr. 4 - XX-150 SG 1 Wet Layup Recirculation CK SA 2 C P,CIN RF CS(31) 9 - XX-151 SG 2 Wet Layup Recirculation CK SA 2 C P.CIN RF CS(31) 9 - XX-152 SG I Wet Layup Recirculation GT M 2 B P.CIN - - - - XX-153 SG 2 Wet Layup Recirculation GT M 2 B P.CIN - - - - j XX-160 Containment Ventilation Units Drain Header GT EL 2 A CIC S 3 mo. - - MT 3 mo. -
-l LPV 2 yr. -
LT 2 yr. 3 f XX-161 Containment Ventilation Units Drain Header GT EL 2 A CIC S 3 mo. - - I MT 3 mo. - ; LPV 2 yr. - LT 2 yr. 3 XX-162 Containment Ventilation Units Drain Header CK SA 2 A/C CIC S RO(26) 13 - LT 2 yr. 3 : RF 3 mo.(26) - Approwd Dwign MeNrW- Dnipn of SSC Pege 3.9149
System 80+ _ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (0 (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-170 PAL 1 Equalization Line CK SA 2 A/C CIC S 3 mo. 15 - LT 2 yr. 7 RF 3 mo. - XX-171 PAL 1 Equalization Line CK SA 2 A/C CIC S 3 mo. 15 - LT 2 yr. 7 RF 3 mo. - XX-172 PAL 2 Equalization Line CK SA 2 A/C CIC S 3 mo. 15 - LT 2 yr. 7 RF 3 mo. - XX-173 PAL 2 Equalization Line CK SA 2 A/C CIC S 3 mo. 15 - LT 2 yr. 7 RF 3 mo. - XX-180 ECW Exp:_ en Tank i Nitrogen Supply CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh I XX-181 ECW Expansion Tank i Nitrogen Supply CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh I XX-182 ECW Expansion Tank i SSWS Make-up CK SA 3 C S 3 mo. 14 9.2.9-1, Sh 1 RF 3 mo. 9 XX-183 ECW Expansion Tank 1 SSWS Make-up CK SA 3 C S 3 mo. 14 9.2.9-1, Sh 1 RF 3 mo. 9 XX-184 ECW Expansion Tank 1 DWMS Make-up CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh I XX-185 ECW Expansion Tank 1 DWMS Make-up CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh I XX-186 ECW Pump 1A Discharge CK SA 3 C S 3 mo. I1 9.2.9-1, Sh 1 RF 3 mo. 9 XX-187 ECW Pump IB Discharge CK SA 3 C S 3 mo. 11 9.2.9-1. Sh 1 RF 3 mo. 9 XX-188 ECW Expansion Tank 2 Nitrogen Supply CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh 5 XX-189 ECW Expansion Tank 2 Nitrogen Supply CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh 5 Approved Design Material- Design of SSC Page 3.9-150 0 0 0
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System 80+ _ Q control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freg Conrg. Fig. No. XX-190 ECW Expansion Tank 2 SSWS Make-up CK SA 3 C S 3 mo. 14 9.2.9-1, Sh 5 RF 3 mo. 9 XX-191 ECW Expansion Tank 2 SSWS Make-up CK SA 3 C S 3 mo. 14 9.2.9-1, Sh 5 RF 3 mo. 9 XX-192 ECW Expansion Tank 2 DWMS Make-up CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh 5 XX-193 ECW Expansion Tank 2 DWMS Make-up .CK SA 3 C RF 3 mo. 9 9.2.9-1, Sh 5 XX-194 ECW Pump 2A Discharge CK SA 3 C S 3 mo. I1 9.2.9-1, Sh 5 RF 3 mo. 9 XX-195 ECW Pump 2B Discharge CK SA 3 C S 3 mo. I1 9.2.9-1, Sh 5 RF 3 mo. 9 XX-196 NCW Containment Supply Division 1 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 12 MT 3 mo.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 1 XX-197 NCW Containment Supply Division 1 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 2 MT 3 mo.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 3 XX-198 NCW Containment Retum Division 1 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 12 MT 3 mo.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 3 Approved Design Material- Desigrr of SSC !*9*3 9-151
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System 80+- _ _ - - Design Control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Vahe Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-199 NCW Containment Return Division 1 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1. Sh 12 MT 3 me.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 3 XX-200 NCW Containment Supply Division 2 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 16 MT 3 me.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 1 XX-201 NCW Containment Supply Division 2 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 16 MT 3 mo.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 3 XX-202 NCW Containment Return Division 2 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 16 MT 3 me.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 3 XX-203 NCW Containment Return Division 2 BF AD 2 A CIC S 3 mo.(32) - 9.2.9-1, Sh 16 MT 3 mo.(32) - FS 3 mo.(32) - LPV 2 yr. - LT 2 yr. 3 XX-2040 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Ilandling Unit I A Control Valve MT 3 mo. - FS 3 mo. - Approved Design Material- Design of SSC Page 3.9-152 O O O
u) v System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) - (a) (b) (c) (d) (c) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test . No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-2041 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 IIandling Unit IB Centrol Valve MT 3 mo. - FS 3 mo. - XX-2042 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Handling Unit IC Control Valve MT 3 mo. - FS 3 mo. - XX-2043 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Handling Unit ID Control Valve MT 3 mo. - FS 3 mo. - XX-2044 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Handling Unit 2A Control Valve MT 3 mo. - FS 3 mo. - XX-2045 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Ilandling Unit 2B Control Valve MT 3 mo. - FS 3 mo. - XX-2046 Channel Electrical Equipment Recirculatien Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Handling Unit 2C Control Valve MT 3 mo. - FS 3 mo. - XX-2047 Channel Electrical Equipment Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Ilandling Unit 2D Control Valve MT 3 mo. - FS 3 mo. - XX-2048 Division i Essential CW Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2049 Division 2 Essential CW Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Handling Unit Control Valve MT 3 mo. - FS 3 mo. - Approved Desigrr Meterial- Desigre of SSC Page 3.9-153
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (I) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-2050 Remote Shutdown Panet Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Ilandling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2051 Division 1 Motor-driven EFW Pump Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Recirculation Air liandling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2052 Division 2 Motor-driven EFW Pump Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Recirculation Air liandling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2053 Division 1 TD EFW Pump Room Recirculation GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Air Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2054 Division 2 TD EFW Pump Room Recirculation GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Air llandling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2055 Division 1 CS IlX Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Ilandling Unit Control Valve MT 3 mo. - FS 3 mo. - ! XX-2056 Division 2 CS IIX Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 i llandling Unit Control Valve MT 3 mo. - I FS 3 mo. - XX-2057 Division i SCS IlX Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 11andling Unit Control Valve MT 3 mo. - l FS 3 mo. - XX-2058 Division 2 SCS 11X Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Ilandling Unit Control Valve MT 3 mo. - FS 3 mo. - l l Approved Design Material- Design of SSC Page 3.9-154 l O O O
System 80+ Desinn conend Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Hg. No. XX-2059 S1 Pump Room 1 Recirculation Air Handling Unit GL .AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Control Valve MT 3 mo. - i FS 3 mo. - XX-2060 SI Pump Room 2 Recirculation Air Handling Unit GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Control Valve MT 3 mo. - FS 3 mo. - XX-2061 51 Pump Room 3 Recirculation Air llandling Unit GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Control Valve MT 3 mo. - FS 3 mo. - XX-2062 SI Pump Room 4 Recirculation Air Handling Unit GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Control Valve MT 3 mo. - FS 3 mo. - XX-2063 Division 1 CS Pump Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2064 Division 2 CS Pump Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2065 Division 1 SCS Pump Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 - Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2066 Division 2 SCS Pump Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2067 CCW Pump I A Room Recirculation Air llandling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Unit Control Valve MT 3 mo. - FS 3 mo. - Approved Design Meterfel- Design of SSC Page 3.9-155
System G0+ - - - - - - - oesign control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (c) (f) (g) (I) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-2068 CCW Pump IB Room Recirculation Air llandling GL AD 3 B S 3 mo. - 9.2.9-1 Sh 2 Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2069 CCW Pump 2A Room Recirculation Air Handling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Unit Control Valve MT 3ano. - FS 3 mo. - XX-2070 CCW Pump 2B Room Recirculation Air llandling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2071 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Channel A Recirculation Air llandling Unit 1 MT 3 mo. - Control Valve FS 3 mo. - XX-2072 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Channel A Recirculation Air Handling Unit 2 MT 3 mo. - Control Valve FS 3 mo. - XX-2073 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Channel B Recirculation Air IIandling Unit 1 MT 3 mo. - Control Valve FS 3 mo. - XX-2074 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Channel B Recirculation Air flandling Unit 2 MT 3 mo. - Control Valve FS 3 mo. - XX-2075 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1 Sh 2 Channel C Recirculation Air llandling Unit 1 MT 3 mo. - Control Valve FS 3 no. - XX-2076 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 2 Channel C Recirculation Air Handling Unit 2 MT 3 mo. - Control Valve FS 3 mo. - Approved Design Material- Design of SSC Page 3.9-156 O O O
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System 80+ _ Des 4pn ControlDocument Table 3.9-15 Inservice Testing of Safety-RcLated Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-2077 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Channel D Recirculation Air llandling Unit 1 MT 3 mo. - Control Valve FS 3 mo. - XX-2078 Vital Electrical and Instrumentation Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 6 Channel D Recirculation Air Handling Unit 2 MT 3 mo. - Control Valve FS 3 mo. - XX-2079 Penetration Room A Recirculation Air Handling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 3 Unit 1 Control Valve MT 3 mo. - FS 3 mo. - XX-2080 Penetration Room A Recirculation Air llandiing GL AD 3 B S 3 mo. - 9.2.9-), Sh 3 Unit 2 Control Valve MT 3 mo. - FS 3 mo. - XX-2081 Penetration Room B Recirculation Air llandling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 7 Unit 1 Control Valve MT 3 mo. - FS 3 mo. - XX-2082 Penetration Room B Recirculation Air Handling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 7 Unit 2 Control Valve MT 3 mo. - FS 3 mo. - XX-2083 Penetration Room C Recirculation Air Handling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 3 Unit 1 Control Valve MT 3 mo. - FS 3 mo. - XX-2084 Penetration Room C Recirculation Air Handling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 3 Unit 2 Control Valve MT 3 mo. - FS 3 mo. - XX-2085 Penetration Room D Recirculation Air llandling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 7. Unit I Control Valve MT 3 mo. - FS 3 mo. - i Approved Design Material- Design of SSC Page 3.9-157
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-2086 Penetration Room D Recirculation Air Handling GL AD 3 B S 3 mo. - 9.2.9-1, Sh 7 Unit 2 Control Valve MT 3 mo. - FS 3 mo. - XX-2087 Division i Fuel PoolIIX Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1. Sh 3 llandling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2088 Division 2 Fuel Pool IIX Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 7 IIandling Ur? Control Vahe MT 3 mo. - FS 3 mo. - XX-2089 Division I CR Mechanical Equipment Room GL AD 3 B S 3 me. - 9.2.9-1, Sh 4 ! Recirculation Air Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2090 Division 2 CR Mechanical Equipment Room GL AD 3 B S 3 mo. - 9.2.9-1, Sh 8 Recirculation Air IIandling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-2091 Division I Control Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 4 IIandling Uni' Control Valve MT 3 mo. - FS 3 mo. - ( XX-2092 Division 2 Control Room Recirculation Air GL AD 3 B S 3 mo. - 9.2.9-1, Sh 8 l Handling Unit Control Valve MT 3 mo. - FS 3 mo. - XX-210 CS Pump Room i Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh I RF 3 mo. PA XX-211 CS IIX Room 1 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh ! RF 3 mo. 9A XX-212 SI Pump Room i Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 RF 3 mo. 9A XX-213 SC HX Room i Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 RF 3 mo. 9A Approved Desiger Material- Desker of SSC Page 3.9-168 L O _ O O
7-) J G G System 80+ Deslan controlDocumnt Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-214 SC Pump Room i Backwater Valve CK SA 2 C S 3 mo. 14 9.3.3-2, Sh 1 RF 3 mo. 9A XX 215 SI Pump Room 3 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 RF 3 mo. 9A XX-216 CS Pump Room 2 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-217 CS IIX Room 2 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-218 St Pump Room 2 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-219 SC IIX Room 2 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-220 SC Pump Room 2 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-221 SI Pump Room 4 Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-222 Division 1 EFW Pump Rooms Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh I RF 3 mo. 9A XX-222 Division 1 CCW Pump Rooms Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 RF 3 mo. 9A
.tX-224 Division 2 EFW Pump Rooms Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-225 Division 2 CCW Pump Rooms Backwater Valve CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 RF 3 mo. 9A XX-230 RB Subsphere Quadrant A Sump Pump 1 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 Discharge RF 3 mo. 9 XX-231 RB Subsphere Quadrant A Sump Pump 2 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 Discharge RF 3 mo. 9 Approved Design Material- Desigre o,' SSC Page 3.9-1S9
System 80+ _ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-232 RB Subsphere Quadrant B Sump Pump 1 CK SA 3 C S 3 mo. 14 9.3.3-2 Sh 2 Discharge RF 3 mo. 9 XX-233 RB Subsphere Quadrant B Sump Pump 2 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 Discharge RF 3 mo. 9 XX-234 RB Subsphere Quadrant C Sump Pump 1 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 Discharge RF 3 mo. 9 XX-235 RB Subsphere Quadrant C Sump Pump 2 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 1 Discharge RF 3 mo. 9 XX-236 RB Subsphere Quadrant D Sump Pump 1 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 Discharge RF 3 mo. 9 XX-237 RB Subsphere Quadrant D Sump Pump 2 CK SA 3 C S 3 mo. 14 9.3.3-2, Sh 2 Discharge RF 3 mo. 9 XX-240 DG Building Sump Pump 1 A Discharge CK SA 3 C S 3 mo. 14 9.5.9-1 RF 3 mo. 9 XX-241 DG Building Sump Pump IB Discharge CK SA 3 C S 3 mo. 14 9.5.9-1 RF 3 mo. 9 XX-242 DG Building Sump Pump 2A Discharge CK SA 3 C S 3 mo. 14 9.5.9-1 RF 3 mo. 9 XX-243 DG Building Sump Pump 2B Discharge CK SA 3 C S 3 mo. 14 9.5.9-1 RF 3 mo. 9 X.X-250 Division 1 CIIRS Suction GL EL 2 A CIC S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 XX-251 Division 1 CHRS Suction GL EL 2 . A CIC S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 Approved Design Material- Design of SSC Page 3.9-160 0 - - - - - 0 - 0- - - -
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.' N System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Relat d Pumps and Valves (Cont'd.)
(a) (b) (c) (d) (e) (f) (g) (i) Velve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No, t XX-252 Division 2 CIIRS Suction GL EL 2 A CIC S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 XX-253 Division 2 CIIRS Suction GL EL 2 A CIC S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - LT 2 yr. 4 XX-254 Division 1 CllRS Discharge GL EL 2 A CIC S 3 mo. - 6.2.5-1 , MT 3 mo. - LPV 2 yr. - LT 2 yr. 1 XX.255 Division I CilRS Discharge CK SA 7 A/C CIC S 3 mo. 14 6.2.5-1 , LT 2 yr. ' 1 RF 3 mo. - XX-256 Division 2 CIIRS Discharge GL EL 2 A CIC S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - LT 2 yr. 1 XX-257 Division 2 CIIRS Discharge CK SA 2 A/C CIC S 3 mo. 14 6.2.5-1 LT 2 yr. 1 RF 3 mo. - XX-258 Division 1 CilRS Individual Suction GL EL 2 B S 3 mo. - 6,2.5-1 MT 3 mo. - LPV 2 yr. - XX-259 Division 1 CIIRS Individual Suction GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - Approved Desigru Meterial- Desiger of SSC Page 3.9-161 __m ____ _ _ ___ _ _ _ __ . _ _ _ _ - _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ < - _ . _
System 80+ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'dd (a) (b) (c) (d) (e) (f) (g) (1) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-260 Division I CIIRS Individual Suction GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-261 Division 2 CIIRS Individual Suction GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-262 Division 2 CIIRS Individual Suction GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-263 Division 2 CIIRS Individual Suction GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-264 Division 1 Recombiner Inlet isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-265 Division 1 Recombiner Inlet Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-266 Division 2 Recombiner Inlet Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - l XX-267 Division 2 Recombiner Inlet Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-268 Division i Recombiner Outlet Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - Approved Design Material- Design of SSC Page 3.9-162 O O O
g s m
) 4 System 80+ Deedsn ControlDocannent 4
Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) . (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test .
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No. Description Type Act Class Cat Funct Reqd Freq Config. Hg. No. XX-269 Division 1 Recombiner Outlet CK SA 2 C. S 3 mo. 14 6.2.5-1 RF' 3 mo. 9A XX-270 Division 2 Recombiner Outlet Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - i XX-271 Division 2 Recombiner Outlet CK SA 2 C S 3 mo. 14 6.2.5-1 RF 3 mo. '9A
.GL XX-272 Division 1 Hydrogen Calibration Supply EL .2 'B S 3 mo. -
6.2.5-1 MT 3 mo. - ! LPV 2 yr. -
- _ XX-273 Division 2 Hydrogen Calibration Supply GL EL 2 B S 3 mo. -
6.2.5-1 , i MT 3 mo. - [ LPV 2 yr. - 4 XX-274 Division 1 Nitrogen Supply . GL S 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - 1- XX-275 Division 2 Nitrogen Supply GL S 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - i XX-276 Division 1 Analyzer Inlet from Recombiner GL S 2 B S 3 mo. - 6.2.5-1 Supply MT 3 mo. - LPV 2 yr. - XX-277 Division 2 Analyzer Inlet from Recombiner GL S 2 B S 3 mo. - 6.2.5-1 Supply MT 3 mo. - , LPV 2 yr. - i GL XX-278 Division 1 Analyzer Inlet from Recombiner S 2 B S 3 mo. - 6.2.5 1 Discharge MT 3 ma. - LPV 2 yr. - ' _w:: Design ntetedel- Design of SSC Page 3.9163 _ _ _ _ _ _ _ _ _ _ _ _.___~_ __ _. _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _ . - . . . . __ . _ _ _ _ . _ _ _ _ . . _ . _ _ _
System 80+ Design Control Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-279 Division 2 Analyzer Inlet fmm Recombiner GL S 2 B S 3 mo. - 6.2.5-1 Discharge MT 3 mo. - LPV 2 yr. - XX-280 Division 1 Analyzer Outlet GL S 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - lXX-281 Division 2 Analyzer Outlet GL S 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-282 Division 1 CHRS Purge to Annulus GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-283 Division 2 CIIRS Perge w Annulus GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-284 Division 1 CilRS Bypass Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-285 Division 2 CilRS Bypass Isolation GL EL 2 B S 3 mo. - 6.2.5-1 MT 3 mo. - LPV 2 yr. - XX-286 Prelube Oil Pump 1 Discharge Relief RV SA 3 C RVT 10 yr. - 9.5.7-1, Sh I XX-287 Prelube Oil Pump 2 Discharge Relief RV SA 3 C RVT 10 yr. - 9.5.7-1, Sh I XX-288 Air Receiver I A Relief RV SA 3 C RVT 10 yr. - 9.5.6-1, Sh I XX-289 Air Receiver IB Relief RV SA 3 C RVT 10 yr. - 9.5.6-1, Sh I XX-290 Air Receiver 2A Relief RV SA 3 C RVT 10 yr. - 9.5.6-1, Sh 2 XX-291 Air Receiver 2B Relief RV SA 3 C RVT 10 yr. - 9.5.6-1 Sh 2 XX-292 ECW Expansion Tank i Relief RV SA 3 C RVT 10 yr. - 9.2.9-1, Sh 1 Approved Design Material- Design of SSC Page 3.9-164 O O O
V iQ. fy w [V System 80+ Design Contred Document Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) (a) (b) (c) (d) (e) (f) (g) (i) Valve Valve Valve Valve Safety Code Valve Test Test Test No. Description Type Act Class Cat Funct Reqd Freq Config. Fig. No. XX-293 ECW Expansion Tank 2 Relief RV SA 3 C RVT 10 yr. - 9.2.9-1, Sh 5 XX-294 Engine-driven FO Pump 1 Discharge Relief RV SA 3 C RVT 10 yr. - 9.5.4-1, Sh I XX-295 Engine-driven FO Pump 2 Discharge Relief RV SA 3 C RVT 10 yr. - 9.5.4-1, Sh 2 XX-296 Motor-driven FO Booster Pump i Discharge RV SA 3 C RVT 10 yr. - 9.5.4-1, Sh I Relief XX-297 Motor-driven FO Booster Pump 2 Discharge RV SA 3 C RVT 10 yr. - 9.5.4-1. Sh 2 Relief XX-298 DG Engine DFO Relief 1 RV SA 3 C RVT 10 yr. - 9.5.4-1, Sh I XX-299 DG Engine DFO Relief 2 RV SA 3 C RVT 10 yr. - 9.5.4-1, Sh 2 Notes: (a) Valve Type: GL- Globe BK- Butterfly GT- Gate PK - Packless CK- Check PL- Plug RV- Relief (b) Valve Actuator: EL- Electric motor S- Solenoid SA - Self actuating EII - Electro-hydraulic AD- Air diaphragm P- Piston M- Manual (c) ASME Safety Classification as defined in Subsection 3.2.2. Approved Design Material- Design of SSC Page 3.9-165 _ _ - - - _ _ . _ . _ - - - - - _ - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ - - _ _ - - - - - _ - - _ _ _ _ - _ . _ _ = - _ _ - _ _ - _ _ - - _ _ ._ - . _ _ - _ - - - _ _ _ _ - - .
System 80+_ _ _ Design ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) s Notes: (Cont'd.) (d) Valve ASME Code Category A, B, C, or D as defined in ASME OM-10. OMa 1988, Subsection 1.4 (e) Valve Function: CIC - Containment Isolation Vaive as listed in Table 6.2.4-1, which is Type-C leakage rate tested in accordance with 10 CFR 50, Appendix J. (Section 3.9.6.2.4). CIN - Containment Isolation Valve as listed in Table 6.2.4-1, which is not Type-C leakage rate tested in accordance with 10 CFR 50, Appendix J (Section 3.9.6.2.4). PlV - Pressure Isolation Valve (Section 3.9.6.2.4). TIV - Temperature Isolation Valve (Section 3.9.6.2.4). P- Passive valves as defined by ASME OM-10. OMa 1988, Subsection 1.3 are denoted by a P in this column. All other valves are active valves. (f) Required Valve Tests per ASME OM-10 and OM-1, OMa 1988: and additional required testing: LT - Valve Leakage Rate Test (per ASME OM-10): Subsections 4.2.2.2,4.2.2.3(e) and (f) for valves with function CIC in (c) above. Subsection 4.2.2.3 for valves with function PlV in (e) above. Reactor Coolant System PlVs are leakage rate tested in accordance with Technical Specifications Surveillance Requirement 3.4.13.1. Subsection 4.2.2.3 for other Category A valves, except for valves which have a TIV function. T!V valves are tested in accordance with Subsection 4.2.2.1 and with the directives contained in their explanatory notes. LPV - Valve Position Verification (ASME OM-10, Subsection 4.1) i S- Valve Stroke Exercise in the forward flow direction: f Category A or B (ASME OM-10, Subsection 4.2.1) ! Category C (ASME OM.10. Subsection 4.3) l RF - Reverse Flow Exercise for A/C and C valves (ASME OM-10 Subsection 4.3). Where the " Test Config." column of Table 3.9-15 (see Note i) lists a Figure 3.9-16 test configuration for a valve's "RF" test, establishment of reverse flow to perform the test is deemed practical. Where a (-) appears instead. l l establishment of reverse flow is impractical, and the COI applicant will determine the alternative positive means described in OM-10 Subsection 4.3.2.4 to be
- employed to verify that the check valve obturator has stroked to its reverse, or seated position. *RF" testing is performed at the same testing frequency as the l corresponding "S" test, unless otherwise described.
MT- Valve Stroke Time Test of Category A or B power operated valves (ASME OM-10 Subsection 4.2.1.4) FS - Valves Test for fail-safe actuation of Category A or B valves (ASME OM-10 Subsection 4.2.1.6) RVT- Relief Valve Test (ASME OM-1) (g) Pump or valve test exclusions, alternatives, and frequency per ASME Code OM-6 and OM-10. For valves whose test frequency exceed the normal frequency, i see the note (as indicated in parenthesis beside the test frequency) for additional information/ justification. 1 l Approved Design Material- Design of SSC Page 3.9-166 O O O
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(v) (V ) System 80+ Design controlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) Notes: (Cont'd.) CS - Cold Shutdown The following condition applies for all testing performed during cold shutdown: Cold shutdown testing in accordance with the requirements of OM-10 Subsection 4.2.1.2(f) and (g), and Subsection 4.3.2.2(f) and (g). Valve exercising during cold shutdown shall commence until all testing is complete or the plant is ready to return to power. A completion of all valve testing is _not a prerequisite to return to power. Any testing not completed by the end of one cold shutdown will be performed during subsequent cold shutdowns, starting from the last test performed at the previous cold shutdown. In case of frequent shutdowns, testing will not be performed more often than once every three months. For extended outages, testing need not be commenced in 48 hours provided all valves required to be tested during cold shutdown will be tested prior to plant startup. RO- Refueling Outage. All Refueling Outage valve testing shall be completed prior to returning the plant to operation. RR- Partially stroke valve at or when proceeding to/ starting up froir cold shutdown. Fully stroke valve during each refueling outage. Some valves may require mechanical exercising or disassembly during each refueling outage to verify operability. All RR testing measures must be completed prior to returning the plant to operation. QC- Partially stroke valve every three months. Fully stroke valve during cold shutdown. El - Valve operates in the course of plant operation at a frequency which satisfies test requirements. Additional exercising not required provided the test parameters are analyzed and recorded at an operational interval not exceeding the test interval requirement. Category A or B (ASME OM-10, Subsection 4.2.1.5) Category C (ASME OM-10, Subsection 4.3.2.3) (h) Pump Test Parameters as defined in ASME OM-6. Subsection 5.2: N- St ad V- Vibration DP- Differential Pressure SP,- Static Suction Pressure Q- Flow Rate SP,- Operating Suction Pressure SP,- Calculated Suction Pressure (vertical wet pit pumps) Note: If OM-6 pump tests cannot be performed on the CCW or SSW pumps due to inability to repeat pump tests single point flow conditions, pump curve testing will be used to assess pump degradation, as described in Section 3.9.6.l. (i) Typical test configurations for pumps and valves requiring special valve arrangements and/or test connections are shown in Figure 3.9-16. When referenced. these typical test configurations constitute design requirements for the affected pump / valve to be reflected in affected documentation by the COL applicant during later detailed design. (1) Valves: CC-130, CC-131 CC-136, CC-137, CC-230. CC-231 CC-236. CC-237 During normal operations, these valves are open to supply / return cooling water to/from the Reactor Coolant Pump (RCP) support coolers. Failure of these valves in the closed position could lead to pump damage or failure and force a unit shutdown. Therefore, these valves will be tested during cold shutdown when the RCP's are not operating. Approved Design Material- Design of SSC Page 3.9-167
System 80+ Design ControlDocument Table 3,9-15 Inservice Testing of Safety-Related Puntps and Valves (Cont'd.) Notes: (Cont'd.) , (2) Valves: CC-1507. CC-1548. CC-2507. CC-2548
- These valves provide containment isolation and overpressure protection for the Component Cooling Water (CCW) supply and return lines to/from the Reactor Coolant Pumps (re: Figures 9.2.2-1.5. Sh 5 and 9.2.2-1 Sh 1I). Since these CCW lines must remain in service during plant operation, it is impractical to perform S or RF testing on the valves on a quarterly test frequency.
Valves CC-1507 and CC-2507 are forward stroke tested during Cold Shutdown by isolating CC-131/CC-231 while keeping CC-130/CC-230 open to allow a CCW header pressure to stroke CC-1507/CC-2507. The reverse stroke (RF) test of CC-1507/CC-2507, however, is impractical to perform without isolating CC-131/CC-231, CC-1509/CC-2509 and CC-1505/CC-2505, and then pressurizing against the check valve seat in the reverse Dow direction via test connection CC-1508/CC-2508. The resultant leakage is then measured through test connection CC-1506/CC-2506. Since this method of testing requires access to areas of high radiation and contamination, a test of this type can be performed only during refueling. This method of testing is the same as will be employed for the 10 CFR 50 Appendix J Type-C leakage rate test. Therefore, LT testing accomplishes and satisfies the reverse flow testing requirements for CC-1507 ar41 CC-2507. Valves CC-1548 and CC-2548 are reverse flow stroke tested during Cold shutdown. With CCW RCP Containment Supply header supply and return lines in service CC-136/CC-236 and CC-1550/CC-2550 are isolated, thus backseating check valve CC-1548/CC-2548 may be then measured via test connection CC-1579/CC-2549. The forward stroke (S) of CC-2628, however, is impractical to perform without isolating CC-136/CC-236, CC-1546/CC-2546, and CC-1550/CC-2550, while keeping CC-137/CC-237 open and injecting a test flow through test connection CC-1549/CC-2549. The resultant outleakage is then measured at test connection CC-1547/CC-2547. Since this valve testing methodology requires containment entries to areas of high radiation and contamination, the forward stroke testing of CC-1548 and CC-2548 will be performed during refueling. (3) Valves: CC-102 CC-103, CC-122. CC-123 CC-202, CC-203, CC-222 CC-223 These valves close on receipt of a Safety injection Actuation Signal to isolate the non-essential component cooling water (CCW) loops. The non-essential cooling loops provide cooling of the Normal Chillers. As described in (32), normally both divisions of Normal Chilled Water (NCW) must operate in order to maintain containment temperature within the Technical Specificat6n limit of Il0*F by the use of three of the four containment coolers. There may be periods during the year, however, when 2 out of 4 containment cooler operatien (one NCW division operating, one NCW division secured) provides su!Yicient cooling to maintain containment temperature within the Technical Specifications limit, due to less severe site climate and heat sink characteristics (e.g., non-summer months). When Division 1 NCW is secured for testing of Division 1 NCW valves, the Instrument Air Compressors on the Division I CCW non-essential header may be secured, and Division I non-essential header CCW valves CC-102, CC-103, CC-122 CC-123 may then be stroke tested. These valves will use the same test frequency as the NCW valves, as described in (32). The Division 2 non-essential CCW header services the letdown heat exchanger in addition to the Division 2 Normal Chillers and Instrument Air Compressors. Closing the Division 2 non-essential CCW header valves during plant operation could result in unnecessary Reactor Coolant System transients. Also, failure to cool the high temperature letdown flow leaving the tegenerative heat exchanger can lead to cavitation at the letdown orifices, which has been known to cause line failure. Therefore, valves CC-202, CC-203, CC-222, and CC-223 will be tested during cold shutdown. (4) Valves: CC-240, CC-241, CC-242, CC-243 These valvn isolate cooling water to/from the letdown heat exchanger and close on a Containment Isolation Actuation Signal. For reasons stated in (3) above, testing these valves during normal operations is not practical. Therefore, these valves will be tested during cold shutdown. Approved Design Material- Design of SSC - Page .3.9-168 O O O
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System 80+ Design control Document Table 3.9-15 Inservice Testing of Safety-Related Putnps and Valves (Cont'd.) Notes: (Cont'd.) (5) Valves: CC-2622 CC-2628 These valves provide containment isolation and overpressure protection for the Component Cooling Water (CCW) supply and return lines to/from the letdown IIcat Exchanger (re: Figure 9.2.2-1, Sh 14). Since these CCW lines must remain in service during plant operation, it is impractical to perform S or RF testing on the valves on a quarterly test frequency. Valve CC-2622 is forward stroke tested during Cc,ld Shutdown by isolating CC-241 whik keping CC-240 open to allow a CCW header pressure to stroke CC-2622. 'Ihc reverse stroke (RF) test of CC-2622, however, is impractical to perform wcw isolating CC-241, CC-2624 and CC-2620, and then pressurizing against the check salve seat in the reverse direction via test connection CC-2623. De resultant leakage is then measured through test connection CC-2621. Since this method of testing requires access to areas of high radiation and contamination, a test of this type can be performed only during refueling. His method of testing is the sarnt as will be employed for the 10 CFR 50 Appendix J Type-C leakage rate test. Therefore LT test lng accomplishes and satisfies the reverse flow testing requiremere for CC-2622. Valve CC-2628 is reverse flow stroke tested during Cold Shutdown. With CCW letdown supply and return lines in service, CC-242 and CC-2630 are isolated, thus backseating check valve CC-2628 against CCW header pressure. Any leakage past CC-2628 may then be measured via test connection CC-2629. The forward stroke (S) of CC-2628, however, is impractical to perform without isolating CC-242, CC-2630, and CC-2626, while keeping CC-243 open and injecting a test flow through test connection CC-2629. The resultant outleakage is then measured at test connection CC-2627. Since this valve testing methodology requires containment entries to areas of high radiation and contamination, the forward stroke testing of CC-2628 will be performed during refueling. (6) Valves: Cil-255 This valve isolates seal injection water to the RCP seats. Valve closure during normal operations with the RCP's operating would result in damage to pump seats. Therefore, this valve will be tested during cold shutdown when the RCP's are not operating. (7) Valves: Cll-505, Cll-506 These valves close on receipt of a Contsinment Spray Actuation Signal to isolate the RCP seal return line. During normal operations, these valves are open to maintain seal injection flow across the RCP seals. Closure of these valves during normal operations would inhibit seal water flow across the RCP seals which would result in damage to the pump seats. Therefore, these valves will be tested during cold shutdown when the RCP's are not operating. (8) Valves: Cll-515, Cll-516, Cll-523, Cll-575 These valves are normally open to pass letdown flow from tlw RCS to the Chemical and Volume Control System (CVCS). Stroking these valves during normal operations could result in unnecessary RCS transients. In addition, these valves experience high stresses when cycled due to the high pressure environment in which they operate. Repeated cycling of the valves at this high pressure could severely affect valve integrity over the expected operating life of the valves. In addition, failure of these valves in the closed position could result in a loss of pressurizer level control forcing a unit shutdown. Therefore, these valves will be tested during cold shutdown when the effects of valve operation are minimized. Globe valve CII-516 performs a Temperature Isolation Valve (TIV) function. This valve isolates the letdown line on a high temperature, as sensed downstream of the letdown lleat Exchanger by dedicated temperature monitors (Re: Figure 9.3.4-1 Sh 1). The setpoints of these temperature monitors and associated valve isolation actuation circuitry are such that the design temperature limits of the interfacing CVCS piping and components will not be exceeded prior to the closure of Ch-516. Temperature monitors are also used to evaluate the integrity of Cll-516 in this closed position. Each refueling outage, an integrity evaluation of Cil-516 shall be performed by isolating the letdown line using Cll-
$16 and then subjecting the valve to Reactor Coolant System pressure and temperature and analyzing the resultant temperature differential across the valve over time. RCS pressure and temperature may be actually lower than plant at-power RCS pressure and temperature levels to avoid valve duty stress, provided these parameters are analyzed and extrapolated to full RCS pressure and temperature.
Approved Design Materini- Design of SSC Page 3.9-169
System 80+ _.__ __ Design contror Document _ Table 3.9-15 Inservice Testir:g of Safety-Related Pumps aml Valves (Cont'd.) Notes: (Cont'd.) (9) Valves: Cil524 nis valve functions as a containment isolation valve and isolates charging flow to the RCS. During normal operations, this charging flow is used to cool the letdown flow in the regenerative heat exchanger and to provide makeup to the RCS. For reasons stated in (3), it is not practical to test this valve during normal operations. In additiorc, failure of this valve in the closed position could result in a loss of pressurizer level control forcing a unit shutdown. Therefore. this valve will be tested during old shutdown. (10) Valves: Cll-747 This valve functions as a containment isolation valve. Testing requires that charging flow be isolated. As stated in (9) above, this is not practical during normal operations. Therefore, this valve will be tested during cold shutdown. (11) Valves: CH-835 This valve functions as a containment isolation valve. Testing requires that seal injection to the RCP's be isolated. As stated in (6), this is not practical during normal operations. Therefore, this valve will be tested during cold shutdown. (12) Valves: EF-200, EF-201, EF-202. EF-203 EF-204, EF-205 EF-206. EF-207 When Emergency Feedwater System (EFWS) operation is required, these valves murt open to provide flow to the steam generators (SG's). Testing of these valves requires EFW injection into the SG's which is not practical during normal operations due to the effects of thermal shock to the SG feedwater nozzles and l potential overcooling of the RCS. Testing during cold shutdown is not desirable due to the SG's being in wet layup conditions. Therefore, these valves will be tested following coki shutdown prior to entering mode 2 which allows normal SG water levels to be established and the system aligned for standby readiness. Reverse flow for reverse flow testing of valves EF-204, EF-205 EF-206, and EF-207 is obtained from operating the momr driven EFW pump in the opposite division as the tested valve and manipulating manual crossover valves EF-288. EF-289. EF-290, and EF-291, as appropriate, to align flow from the motor driven EFW pump to the valve under test. Reverse flow for reverse flow testing of valves EF-200, EF-201 EF-202, ami EF-203 is obtained by opening EFW Pump to Steam Generator Isolation Valves EF-100. EF-101, EF-102, and EF-103, respectively. l (13) Valves: RC-406, RC-407. RC-408, RC-409 The safety-related function of these valves is to remain closed in order to maintain the integrity of the Reactor Coolant Pressure Boundary (RCPB). These valves are open only for a beyond design basis event of a Total Loss of Feedwater (TLOFW). These valves are Class I valves and the RCPB class break is I downstream of the second valve in each SDS line in accordance with the 10CFR50.2 and the ANSI /ANS 51.1 definitions of RCPB. The definition of RCPB in l these two documents require two valves to be normally closed during normal reactor operation. l ANSI /ANS 51.1 identifies numerous combinations of valve arrangements that may be used to comply with these criteria. The Rapid Depressurization (RD) valves' design complies with these requirements. If one of the RD valves was opened, for example during testing, the RCPB is not mainta4 ed according to the 10CFR50.2 and tl.e ANSI /ANS 51.1 definitions during the testing of the valves. Since these valves are opened only for low probability events (TLOFW and severe accidents) and it is essential to maintain the integrity of the RCPB, these valves are stroke tested during a shutdown period when it would be more acceptable to compromise the RCPB Herefore, these valves should be tested during plant shutdown periods and not during reactor operation. Approved Design A*eterial- Design of SSC Pege 3.9-170 0 0 0
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b d System 80+ : Design ControlDocameent Table 3.9-15 Inservice Testing of Safety-Related Purnps and Valves (Cont'd.) Notes: (Cont'd.) , in addition, the Rapid Depressurization valve configeration is similar to the Shutdown Cooling System (SCS) suction line valve configirration (i.e., there are two normally closed motor-operated valves between the ECS and the connecting system). ABB-CE has specified only cold shutdown testing of the SCS suction line 4 valves (SI-651, 51-652. 51-653 and 51-654), to prevect placing the operating plant within one valve of an intersystem LOCA challenge. Similar logic applied to , the RD valves suggests these valves should be tested only at cold shutdown to prevent a LOCA. challenge. These valves will be tested as Cold Shutdown valves. (14) Valves: RC410. RC41I, RC 412. RC-413. RC-414 RC-415 RC-416. RC-417 Dese valves are closed during normal plant operating to maintain the Reactor Coolant Pressure Boundary (RCPB). These valves are active valves and are designed to be used during a safety grade cooldown of the RCS. Opening these valves during normal op-ration leaves only one Class I valve, which does not maintain the RCPB according to the 10CFR50.2 and ANSI /ANS 51.1 dermitions. While there is a third valve downstream of the two Reactor Coolant Gas Vent
- System (RCGVS) valves, the piping and the third valve are Class 2. In order to maintain the integrity of the RCPB, these valves should be tested during plant shutdown periods only and not during reactor operation.
These valves will be tested r- Cold Shutdown valves. (15) Valves: SG-130. SG-132. SG-135, SG-137. SG-172 SG-174, SG-175 SG-177 These valves isolate main feedwater to the SG's upon receipt of a Main Steam Isolation Signal (MSIS). Closure of these valves during normal operations would , isolate feedwater to the SG's which may result in a severe transient in the SG and a unit trip. Therefore, these valves will be tested during cold shutdown. (16) Valves: SG-140 SG-141, SG-150 SG-151 These valves are Main Steam Isolation Valves (MSI"s), which isolate the main steam lines upon receipt of a MSIS. Performance of either a full stroke or partial stroke test during normal operations may cauw severe transients in the main steam lines and result in a unit trip. He valves will therefore be full stroke tested on a cold shutdown frequency basis, but with .1 unit in MODE 3 and at operating temperature and pressure, so as to replicate design conditions under which valve closure must be achieved. (17) Valves: SI-484. 51-485. 51-568. 51-569 These valves must close to prevent reverse flow when either the SC pumps are use to provide containment spray or the CS pumps are used to provide shutdown cooling flow. These valves are tested by operating either the SC or CS pump in a division, cpening the discharge crossover isolation valve between the two sys ;ms, and isolating the suction of the off-line pump. Closure of the check valve on reverse flow in the discharge of the off-line pump is verified by i monitoring pressure increase upstream of the valve. (18) Valves: SI-113. SI-133, SI-404, 51-405, SI-434, SI-446, 51-522. SI-523, $1-532, 51-533, SI-540, 51-542 l Check valves 51-113,51-133. 51-404, S1-405 S1-434,51-446, SI-540, and S1-542 must be provided with sufficient flow from the Safety injection Pumps to i stroke to their full open position. The flow for stroke testing these valves passes through the Direct VesselInjection (DVI) nozzles and into the Reactor Coolant System (RCS). The Safety Injection Pump's discharge pressure is not sufficient to overcome normal RCS operating pressure. In addition, any flow from Safety Injection through these valves and into the RCS during power operations would produce an undesirable temperature transient at the DVI nozzles. The valves are also not full or partial stroked during Cold Shutdown, since this may result in low temperature overpressurization of the RCS, Since it is impractical to full 4 stroke test these check valves during plant operation or to perform full / partial stroke test during Cold Shutdown conditions, these valves are full stroke tested each refueling outage. l Approwd Destgrr A0eteriet - Desigre of SSC Pope 3.9-171
System 80+ Design Control Document _ Table 3,9-15 Inservice Testing of Safety-Related Putnps and Valves (Cont'd.) Notes: (Cont'd.) Check valves SI-522. SI-523, SI-532, and SI-533 nmt be provided with sufficient flow from the Safety Injection Purnps to stroke to their fu!I open position. The flow for stroke testing these valves passes through the Reactor Coolant System (RCS) Hot Legs (Hot Leg Injecten). The Safety Injection Pump's discharge pressure is not sufficient to overcome normal RCS operating pressure. in addition, any flow from Safety injection through these valves and into the RCS during power operations would produce an undesirable temperature transient at the Shutdown Cooling line connections to the Hot Legs. The valves are also not full or partial stroked during Cold Shutdown, since this may resuh in low temperature overpressurization of the RCS. Since it is impractical to full stroke test these check valves during plant operation or to perform full / partial stroke test during Cold Shutdown conditions, these valves are full stroke tested each refireling outage. Check valves SI-113 and SI-133 are not reverse flow tested quarterly, since testing of these valves during power operations would require containment entries by plant personnel to high radiation and airborne contamination areas. These valves are not reverse flow tested every Cold Shutdown, because of the extensive test equipment setup which could extend the Cold Shutdown. These valves are reverse flow tested during refueling. (18A) Valves: SI-404. 51-405, SI-434, S1-446 These valves are reverse flow tested by pressurizing the volume of piping t'etween these valves and their respective SI pump discharge maintenance isolation valve (S1476,51-478, SI-435, and SI-447) with water, and using either pressure decay or volumetric analysis to determine valve reverse seating function. (19) Valves: SI-123, SI-143, SI-541, SI-543, 51-168, SI-178 Check valves SI-123, SI-143, SI-541, SI-543, SI-168, and SI-178 must be provided with sufficient flow to stroke to their full open position. This test flow ultimately passes through the Direct Vessel injection (DVI) nozzles and into the Reactor Coolant System (RCS). Neither the Safety injection nor the Shutdown Cooling Pumps' discharge pressures are sufliciem to overcome normal RCS operating pressure in order to establish the flow required to perform a partial or full stroke test of these valves. In addition, any flow from Safety injection or Shutdown Cooling to the RCS during power operations would produce an undesirabic temperature transient at the DVI nozzles. During Cold Shutdown, the Safety Injection Fu ops may not be used for stroke testing these valves, because this could result in low temperature overpressurization of the reactor vessel. A full flow stroke test of these valves during Cold Shutdown is achievable by use of the Shutdown Cooling Pumps. l Valves SI-123 and SI-143 are not reverse flow tested quarterly, since testing of these vz!ves during power operations would require containment entries by testing personnel to high radiation and airborne contamination areas. These valves are not reverse flow tested every Cold Shutdown, because of the extensive test equipment setup which could extend the Cold Shutdown. These valves are reverse flow tested during refueling. (20) Valves: SI-164, SI-165 Tly;se valves are required to open to pass flow from the Containment Spray (CS) pumps to the containment atmosphere. These valves cannot be stroked open with CS flow, since this would result in spraying down containment. These valves will be equipped with external means to exercise the valve obturator and to measure the force required to exercise the valve open and closed (this performs both the S and RF test). Since the valves M located in a containment area subject to moderate to high radiation and contamination levels, the valves will be exercised each refueling outage, instead of during Cold Shutdown or plant operation (each 3 months). l Approved Design Matraint - Design of SSC Page 3.9-172 O O O
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System 80+ Deskn ControlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) Notes: (Cont'd.) (21) Valves: 51-215, 51-225 SI-235 SI-245 These SI Tank outlet check valves must be provided with sufficient flow from the SI tanks to the Reactor Coolant System (RCS) to stroke to their full open position. During normal operations, the SI Tanks are not capable of providing flow to the RCS, due to RCS pressure and tank pressure limitations. Also, providing flow to the Direct Vessel Injection (DVI) nozzles during plant operations would cause undesirable temperature transients at the DVI nozzles. The SI Tanks may be used, however, to provide flow to partially stroke these valves, with minimal temperature transient impact to the DVI nozzles, when proceeding to or starting up from Cold Shutdown. In this configuration a full flow stroke test is impractical, due to significant inventory additions to the RCS should SI Tank level become tco low. During refueling and with the reactor head removed, full flow testing of these valves is practical. In this condition, SI Tank flow may be obtained which is sufficient to full stroke the SI Tank outlet check valve, with minimal risk of injecting nitrogen into the RCS. Therefore, these valves will be partially stroke tested during Cold Shutdown and full stroke tested during refueling. (22) Valves: SI-217, 51-227, SI-237, SI-247 Providing flow to the Direct VesselInjection (DVI) nozzles in order to stroke these check valves during plant operation is not practicable, since Reactor Coolant System (RCS) pressure during normal operations is significantly higher than the discharge pressures of the Safety Injection Pumps, Shutdown Cooling Pumps, or Safety Injection Tanks. In addition, any flow from these sources to the RCS during power operations would produce an undesirable temperature transient at the DVI nozzles. Check Valves SI-217 and SI-237: Full stroke testing of these check valves is not practical at Cold Shutdown for several reasons. First, Safety Injectio i Pumps 4 and 3 are not capable of providing sufficient flow to full stroke their respective DVI check valve (SI-217/SI-237). Secondly, such use of Safety Injection pumps during Cold Shutdown condition is not practical, since it could result in low temperature overpressurization of the reactor vessel. Thirdly, use of water inventory from their respective Safety Injection Tanks (SITS) to full stroke test these check valves during Cold Shutdown when RCS pressure is low is impractical, because of the risk of injecting Nitrogen into the RCS, and because the RCS is not capable of accepting the added SIT inventory from a full stroke test. Ilowever, a partial stroke test of these check valves may be achieved with minimal temperature transient impact to their respective DVI nozzles, when proceeding to or starting up from Cold Shutdown, by use of water inventory from their respective SITS to establish flow through these valves. During refueling and with the reactor head removed, full flow testing of these valves is practical. In this condition, SI Tank flow may be obtained which is sufreient to full stroke the respective DVI check valve, with minimal risk of injecting nitrogen into the RCS. Therefore, these valves will be partially stroke tested during Cold Shutdown and full stroke tested during refueling. Check Valves SI-227 and SI-247: These check valves have the same testing limitations as SI-217 and S1-237, above, except that a full stroke test of the check valves is practical during Cold Shutdown by operating their respective Shutdown Cooling Pump. SI-227 and SI-247 are tested as Cold Shutdown valves. (23) Valves: SI-302,SI-303 Closing of these valves to perform stroke testing renders both Safety Injection pumps in the valves's respective division inoperable, which is in violation of Technical Specification 3.5.2 and 3.5.3. Technical Specification 3.5.3, however, allows inoperability of both a division's Safety injection pumps during refueling subject to prescribed Reactor Coolant System parameters. Approved Design Meterial- Design of SSC Page 3.9-173
System 80+ Design ControlDocument Table 3,9-15 Inservice Testing of Safety-Related Punips and Valves (Cont'd.) Notes: (Cont'd.) (24) Valves: EF-100. EF-101, EF-102 EF-lG3 Gate valves EF-100. EF-101 EF-102, and F,F-103 similarly provide the highliow temperature interfaces between Main Feedwater System piping (Design Temperature: 575') and the Emergency Feedwater System pipmg (Design Temperature: 140'), as illustrated in Figure 10.4.9-1. Sh 1. These valves are maintained closed during plant operation to prevent backleakage into the Emergency Feedwater System piping, which can result in exceeding piping design temperature and steam binding of the EmergencyFeedwater Pumps (Re: Response to Generic Safety issue 093). Dedicated temperature monitors, which alarm to the control room, are located upstream of these valves and are stacd to detect any high temperature backleakage from the Main Feedwater System. The setpoints of these temperature monitors and associated alarms will be such that the design temperature limit of the interfacing Emergency Feedwater System piping will
' tot be exceeded prior to the initiation of the alarm.
(25) Valves: 424. SI-426, SI-448 SI-451 Each of these valves is reverse flow tested by isolating its associated pump and operating the other divisional Si pump in miniflow to provide reverse flow against the tested valve. In this test alignment, the operating pump must remain in miniflow condition, fully capable of supplying design basis accident flow. Procedural measures are implemented thus so that only one divisional Si pump will be inoperable (i.e. the isolated SI pump). (26) Valves: XX432, XX-035, XX-162 Valves XX-032 and XX-035 serve as both containment isolation valves and as thermal relief valves for the steam generator blowdown lines. They are effectively RF tested by closing the inside steam generator blowdown valve while leaving the outside steam generator blowdown valve open. (re: Figure 10.4.8-1). In this manner XX-032/XX-035 are backseated by steam generator pressure. The forward stroke test of XX-032 and XX435, however, is impractical to perform without first depressurizing the steam generator blowdown line upstream of XX-032/XX-035, so that test flow may stroke the valves in their relieving direction (in the direction of the steam generators). Thus a quarterly test frequency, which assumes plant operation conditions, is impractical. Since the steam generators are depressurized during Cold Shutdown, the forward stroke test will be performed at that frequency. Valve XX-162 serves as both a containment isolation valve and as a thermal relief valve for the containment ventilation units drain header. This valve is j efTectively RF rested by closing the inside containment ventilation units drain header containment isolation valve while leaving the corresponding outside i containment isolation valve open. In this manner valve XX-162 is backseated against the ventilation units' drain header static pressure. Such a test could be l performed quarterly, while the unit is at power. The forward stroke of XX-162, however, is impractical to perform without isolating and removing from service l the ventilation units drain header to enable test flow to be injected into the line itself in the containment direction. During plant operations, these ventilation l units condense considerable condensate, and the drain header line must remain open to prevent condensate backup into the ventilation units. Also, it is l impractical to perform the forward stroke test of this valve during operations and Cold Shutdown since this valve and applicable test connections are located in areas of high radiation and airborne contamination during power operations and Ccid Shutdown. 'the stroke test "S" for XX-162 will be performed on a refueling outage basis for ALARA purposes. (27) Valves: Cll-304 This check valve functions as a containment isolation valve and isolates the shutdown cooling purification line. During normal operation, Cll-304 and manual valve Cll-307 are in the closed position to istolate this line. When shutdown cooling purification is used during cold shutdown, these valves are open. To ensure eperability of CH-304 for its containment isolation function, this r?ve is reverse flow tested. Reverse flow testing of CH-304 is not practical quarterly, since during unit operation, opening of Cll-307 or other venting path could result in an inter-system LOCA. The appropriate interval for such testing is during cold shutdown when the shutdown purification line is secured prior to unit startup. Approved Design Ataterial- Design of SSC Page 3.9-174 O O O
. _ _ _ _ _ _ . _ _ . _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ m ___ _ _ _ _ _ _ _
r - g U v x.J System 80+ Design ControlDocument Table 3,9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd,) Notes: (Cont'd.) (28) Valves: S1414, S1424, 51-634, 51444 These valves must be open to pass flow from the Si tanks to the RCS. Technical Specifications do not permit testing these valves during normal operation since all four Si tanks must be operable. Normal shutdown /startup procedures require these valves to be closed when proceeding to cold shutdown and to be opened when starting up from cold shutdown. Testing of these valves will be performed at this time.
. (29) Valves: SI-651, SI452, S1453, SI-654, SI-655, 51-656 These valves must open to align the SC pump suction to the RCS. These valves are interlocked such that they cannot be opened when RCS pressure is above the operating pressure of the SCS. Therefore, these valves cannot be tested during normal operations. Testing will be performed during cold shutdown when valves can be manipulated.
(30) Valves: XX-010, XX411, XX-012, XX-013, XX-014, XX-015, XX-016, XX-017 These valves must close on receipt of a CIAS to perform their containment isolation function. During normal operations, these valves are closed and Technical Specifications do not permit opening. Therefore, these valves will be tested during cold shutdown. (31) Valves: XX-150, XX-151 These valves must close on icverse flow in the SG wet layup recirculation line to perform their containment isolation function. The recirculation liaes are isolated during normal operations and are only used when the SG's are in wet layup conditions such as during cold shutdown. Therefore, these valves will be tested during cold shutdown when the recirculation system is stopped. (32) Valves: XX-196, XX-197, XX-198, XX-199, XX-200, XX-201, XX-202, XX-203 During normal operations, these valves are open to proved Normal Chilled Water (NCW) to the containment and CEDM coolers. Stroke testing of any one of these valves will require interruption of at least one division of NCW to the containment and CEDM coolers. To maintain containment air temperatures within the 110'F Technical Specifications limit year-round, three of the containment coolers are required to operate with the fourth unit in standby. There may be periods during the year, however, when 2 out of 4 containment cooler operations (one NCW division operating, one NCW division secured) provides sufficient cooling to maintain containment temperature within the Technical Specifications limit, due to less severe site climate and heat sink characteristics (e.g., non-summer months). For these periods of the year, the valves will be quarterly stroke tested. For other periods of the year during which at least 3 of the 4 containment coolers must be kept in operation to maintain containment temperature within the II0'F Technical Specification limit, the valves will be tested as cold shutdown valves. (33) Valves: SI-605, SI-606, 51-607, SI-608, SI-613, SI-623, S1433, S1443 These solenoid valves are both stroke tested (S) and fail-safe tested (FS) during cold shutdown because opening any of these valves will result in depressurizing the affected Safety injection Tank (SIT),thus causing the SIT to be inoperable. These valves cannot be tested during plant operations, since plant Technical Specification 3.5.1 requires all SITS to remain operable in Mode I (Power Operations). Technical Specification LCO 3.5.1 (Required Action) for inoperability of any SIT requires restoration of that SIT to operable status in one hour, or commence unit shutdown. Since this LCO is too stringent to allow valve stroke or fail-safe testing of these valves during plant operations, this testing will be performed during Cold Shutdown. Approved Design Material- Design of SSC Page 3.9-175
System 80+~ ~._-_. -.~.-- -.---~_---.- -- Design contrer Document l l l- Table 3,9-15 Inservice Testing of Safety-Related Purnps and Valves (Cont'd.) Notes: (Cont'd.) (34) Valves: SG-567 SG-598 SG-599, SG-612. SG442, SG443, SG450, SG451, SG452, SG-653 These Mahi Feedwater System check valves are located on the feedwater inlet lines to the steam generators. These check valves have only a safety function to j close. Since these valves must remain open during power operations to maintain steam generator level and prevent reactor trip and plant shutdown, quarterly reverse flow testing is impractical. As described in Section 10.4.7 and Figure 10.4.7-3, the feedwater split between the economirer feedwater lines amt the downcomer feedwater line always maintains some flow through the downcomer feedwater line, even though the two economizer feedwater lines are sized to l collectively provide 100% required flow to the steam generator they service. The 10% of required steam generator flow which passes through the downcomer line at full power operation is used to maintain the downcomer line at a constant temperature to protect the line from thermal transient (waterhammer) damage.
'Ihus, the downcomer feedwater line may not be isolated during power operations in order to perform a reverse flow test on its feedwater check valves. Flow I
testing of these valves is performed on a cold shutdown (CS) frequency basis while the plant is in Mode 3 (llot Standby), at which condition adequate steam ( generator pressure exists to perform reverse flow tests on the valves. (35) Valves: DS-I12, DS-117. DS-118, DS-122, DS-127. DS-128, DS-212. DS-217, DS-218. DS-222, DS-227, DS-228 These valves are tested by isolating one of the starting air receiver tanks and starting the diesel generator using the remaining operational tank to direct full flow through the tested valves. In addition to the diesel generator starting using the one air receiver tank, positive means in accordance with ASME OM-10 are l provided to verify conclusively that check valves DS-117, DS-118 DS-127 LS-128, DS-217, DS-218, DS-227, and DS-228 stroke open. Positive means are ! also employed to verify reverse flow seating in accordance wi:h ASME OM-10 for check valves DS-112. DS-Il7, DS-IIS, DS-122, CS- 27 DS-128, DS-212, l DS-217. DS-218, DS-222 DS-227, and DS-228. (36) Valves: XX 041, XX-043 The safety function of valves XX-041 and XX-043 in the forward stroke direction is to relieve thermal pressure to the containment Fire Water Supply piping and thus prevent damage to the containment penetration as a result of containment heatup following a LOCA. It is impractical to perform a forward stroke test for check valves XX-041 and XX-043 during power operations or Cold Shutdown for several reasons: significant radiation and contamination exposure to test personnel in consainment, the necessity of disabling the sprinkler system within containment to perform the test which jeopardizes the system's response to containment fire, and the extensive restoration / draining of the fire supply headers inside of containment post-testing to their normal
- dry" status which would result in extending the Cold Shutdown.
The reverse flow safety function is containment isolation. Reverse flow testing of these check valves is impractical during power operations or Cold Shutdown for several reasons: significant radiation and contamination exposure to test personnel in containment, the necessity of disabling the sprinkler systems to fill the
" dry" Fire Water Supply piping in the reverse flow test volume in order to establish backpressure on the check valve seat which jeopardizes the system's response to containment fire, and the extensive restoration / draining of the fire supply headers inside of containment post-testing to their normal
- dry" status which would result in extending the Cold Shutdown.
Approved Design Afsterial- Design of SSC Page 3.9-176 O O O
p p O O O System 80+ -- Deskn controlDocument Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) Notes: (Cont'd.) (37) Valves: XX-071 The safety function of valve XX-071 in the forward stroke direction is to relieve thermal pressure to the containment Demineralized Water piping as a result of containment heatup following a LOCA. Verification of this safety function requires forward stroke testing, and use of demineralized water within containment. Ilowever, during power operations and Cold Shutdown, there are no users of demineralized water within containment to establish this flow, without necessitating containment entry to areas on high radiation dose and airborne contamination present during power operations and Cold Shutdown to manipulate manual valves at decontamination sinks, etc. The forward stroke test will be performed during refueling for ALARA purposes. Similarly, the RF test will require containment entry to areas of high radiation dose and airborne contamination present during power operations and Cold Shutdown. For ALARA purposes, this test will be performed during refucting. (38) Note Deleted (39) Valves EF-XXX (Steam Driven EFW Pumps I and 2 Turbine Trip and Thrcttle Valves and Governor Valves) Each Steam Driven EFW Pump Trip and Throttle (T&T) Valve is tested during each refuelirag outage. Testing involves uncoupling the turbine drive shaft from the EFW pump and operating the turbine by means of auxiliary steam to achieve the T&T Valve overspeed trip setpoint. The T&T Valve overspeed tripping function is verified along with stroking the T&T Valve from its tripped position to its reset (open) position. Due to the involved maintenance procedures required to perform this testing (i.e., securing / tagging out the pump and turbine lines for maintenance work, uncoupling the pump from the turbine, performing the overspeed trip testing, adjusting the trip point setting if necessary, recoupling the pump to the turbine, coupling balancing, and formal retest of the pump), this testing is impractical to perform during Technical Specifications LCO 3.7.4 in Modes 1-3. Due to plant availability concerns it is also impractical to perform such maintenance related testing during Mode 4 (llot Shutdown) and Mcde 5 (Cold Shutdown). Therefore, this testing will be performed on a refueling outage frequency. Each Steam Driven cFW Pump Governor Valve is tested for functionability by the quarterly ASME OM-6 test performed on each Steam Driven EFW Pump. Testing consists of monitoring valve performance during the initial start of the pump to verify that the initially open governor valve controls the turbine speed and prevents a turbine overspeed trip. Once the pump has started, pump speed is incrementally adjusted throughout the RPM operating range of the pump, and RPM measurements are taken e demonstrate pump speed control settings correspond with the actual pump RPM. (40) Valves: SI-390 SI-391, SI-392,51-393, SI-394, SI-395 S!-390, SI-391, SI-392, and SI-393 are motor-operated holdup volume tank (IIVT) flooding valves; SI-394 and SI-395 are reactor cavity flooding valves. The valves are normally closed and remain closed thmughout the recovery period of any design basis accident. The valves are opened only for a severe accident which requires flooding of the reactor cavity in the event of the reactor vessel breach. The opening of the valves allows water to flow from the IRWST to the reactor cavity to cover core debris. Operability of the valves is not required for shutting down the reactor, maintaining cold shutdown, or mitigating the consequences of any design basis accident. Testing of the IIVT flooding valves requires that the manual valves located upstream be closed to prevent the flow of water from the IRWST to the llVT. Closing the manual valves is not practical during operations at power because containment entry would be required. The reactor cavity flooding valves are not tested during operations at power because, in the event of a design basis accident, the failure of the valves in the open position would provide sn open flow path from the llVT to the reactor cavity. This would adversely affect the operability of the IRWST in mitigating the consequences of a design basis accident; operability of the IRWST is required for all plant operating modes from power operation through, and including, cold shutdown. The IIVT flooding valves and the reactor cavity flooding valves will be tested during each refueling outage. This will limit personnel radiation exposure and minimize the potential impact on the operability of the IRWST. Approved Design Material Design of SSC Page 3.9-177
System 80 + Design ControlDocument ~ Table 3,9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) Notes: (Cont'd.) i (41) Valves: 51412 SI422 , S!432 , S1442 , 51419. 51429, 51439, SI-649 These air. operated SIT Nitrogen Pressure Control Valves are stroked in the course of plant operation as a matter of normal operation and pressure control of the SITS at a frequency which satisfies test requirements of quarterly testing. Fail-safe actuation (FS) on a 3 month basis, however, is impractical during plant operations (quarterly test frequency) or cold shutdown because such testing involves entries to containment to proximity of the SITS (high radiation dose and airborne contamination area) to fail air to the air diaphragm valve actuators. Therefore, the FS test for these valves will be performed on a refueling outage l basis for ALARA purposes. l (42) Valves: SI-322. 51-332, S1411, 3I418, S1421, SI428, SI431, SI438, SI-641, SI448, SI461, SI470 l These air-operated valves are stroked on a quarterly frequency. Fail-safe (FS) actuation testing on a 3 month basis, however ,is impractical during plant ! operations (quarterly test frequency) or cold shutdown because such testing involves entries to containment to proximity of the SITS (high radiation dose and f airborne contamination area) to fail air to the air diaphragm valve actuators. Therefore, the FS test for these valves will be performed on a refueling outage basis for ALARA purposes. (43) Although these Emergency Diesel Generator support system components are Safety Class 3, they are procured, tested and maintaincd as part of the Emergency l Diesel Generators themselves, which are tested for operability and reliability by the plant Technical Specifications. Therefore, these components are tested by l Technical Specifications Survei!!ance Requirements of Technical Specification Section 3.8. (44) Pressure Isolation Valves (PIVs) are not reverse flow tested quarterly, since testing of these valves during power operation would require containment entries to l high radiation and airbome contamination areas. PIVs are not reverse flow tested every Cold Shutdown, because of the extensive test equipment setup which could extend the Cold Shutdown. The RF function is verified, however, by leakage testing each valve in the reverse flow direction during unit startup for the testing frequency outlined in Technical Specification Surveillance Requirement 3.4.13.1. 'this surveillance requirement states that leakage testing of these valves is required every 18 months AND prior to entering Mode 2 whenever the plant has been in Mode 5 (Cold Shutdown) for 7 days or more, if leakage testing has not been performed in the previous 9 months AND within 24 hours following valve actuation due to automatic or manual action or flow through the valve (s). (45) Inservice Testing / Monitoring for Valves on Pipinf', Connected to the Steam Generators
- Secondary Side l
( Steam Generator (SG) Main Steam Isolation Valves and SG Main Steam Isolation Dypass Valves are tested for gross leakage each refueling outage. Testing is l performed by isolating these valves with the steam generators under steam pressures created by normal startup/ shutdown and measuring downstream steam header pressure and temperature. SG Main Steam Safety Valves are tested each refuehng outage for gross leakage by means of walkdown/ temperature / acoustic monitoring with main steam lines l pressurized. ! SG Atmospheric Dump Valves are tested for gross leakage cach refueling outage by temperature / acoustic monitoring of the ADV lines downstream of the ADVs I w;th main steam lines pressurized. Steam Generator Blowdown Valves are tested for gross leakage each refueling outage. Testing is performed by isolating these valves individually against steam generator pressure and then monitoring the steam generator blowdown tank for an increase in tank level which would be indicative of gross valve leakage. Steam Generator Sampling Line Valves are tested for gross leakage each refueling outage. Testing is performed by isolating these valves individually against ! steam generator pressure and then monitoring sample line flow for gross valve leakage. l Approved Design Material- Design of SSC Page 3.9-178
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System 80+ DeaTan coneet occammat Table 3.9-15 Inservice Testing of Safety-Related Pumps and Valves (Cont'd.) Notes: (Cont'd.) Main Feedwater Containment Isolation Valves are tested for gross leakage each refueling outage. Testing is performed by individually subjecting these valves to steam generator pressure experienced during unit startup/ shutdown and then measuring resultant valve leakage through the provided test connection. The Startup . Feedwater Pump may be used for establishing and maintaining steam generator inventory for this gross leakage test. Emergency Feedwater (EFW) Containment Isolation Valves are tested for gross leakage each refueling outage. The EFW Containment Isolation Check Valves are leakage tested by individually subjecting these valves to steam generator pressures experienced during unit startup/ shutdown and then measuring resultant . valve leakage through the provided test connection. The outside<ontainment EFW Containment Isolation Valves are leakage tested by pressurizing the piping between these valves and their inside<ontainment Containment Isolation Check Valves while the steam generators are at startup! shutdown pressures. Valve leakage is then measured through the provided test connection. These EFW valves also employ installed temperature instrumentation to detect leakage past these-valves. (46) Safety Injection System For inservice testing of the safety injection 3. umps during refaeling outages, a walkdown visual examination of safety injection system piping and components outside Containment will be conducted to verify the leak tight integrity of the system. 4 i 1 F T Approvat Desip MatwM - Desig of SSC Page 3.9-179
i System 80+ Design control Document l l Table 3.9-16 Pressure Isolation Valves { Note: The following is a listing of pressure isolation valves associated with the Reactor Coolant l System (RCS). Detailed testing information is contained in Table 3.9-15, " Inservice j Testing of Safety-Related Pumps and Valves," and Technical Specification Surveillance Requirement 3.4.13.1. Valve Description l SI-215 SI Tank 4 Discharge Check SI-217 DVI Nozzle 2B Check Valve SI-225 SI Tank 2 Discharge Check SI 227 DVI Nozzle IB Check Valve SI 235 SI Tank 3 Discharge Check SI-237 DVI Nozzle 2A Check Valve SI-245 SI Tank 1 Discharge Check SI-247 DVI Nozz!c 1 A Check Valve SI-322 Hot Leg Injection 1 Bleed Off Isolation SI-332 Hot Leg injection 2 Bleed Off Isolation SI 522 Hot Leg Injection Loop 1 SI-523 Hot Leg Injection Loop 1 SI-532 Hot Ixg injection Loop 2 SI 533 Hot Leg injection Loop 2 SI-540 SI Pump #4 Discharge SI-541 SI Pump #2 Discharge SI-542 SI Pump #3 Discharge SI-543 SI Pump #1 Discharge SI 618 SI Line 4 Leakage Return SI428 SI Line 2 Leakage Return SI-638 SI Line 3 leakage Return SI-648 SI Line 1 Leakage Return SI-651 SC Pump #1 Suction SI-652 SC Pump #2 Suction SI-653 SC Pump #1 Suction SI-654 SC Pump #2 Suction O Astuswd Desiers MaterW - Desigrs of SSC Page 3.9-180
.- - . . - .. - .- . , - ._ . - . ~ . - .- . - . - - . - - . - - . - -..- ..- .
J System 80+ De-ion ControlDocumart :. IO-- ' Table .,.9-17 Reactor Vessel Internals Arrangement Comparison i Palo Verde System 80+ ; i' Core Support Barrel i' . A. Right Circular cylinder,3 sections, A. Right Circular cylinder,3 sections, ;
- supported by heavy external flange top supponed by heavy external flange top i end, heavy internal flange bottom end. end, heavy internal flange bottom end. ~l 1
B. Two outlet nozzles through barrel. B. Two outlet nozzles through barrel. l l C. Top flange seats on reactor vessel ledge, C. Top flange seats on reactor vessel ledge, and has (4) alignment keys attached, and has (4) alignment keys attached. D. Bottom flange suppons lower support D. Bottom flange suppons lower support
- structure, fuel and core shroud. structure, fuel and core shroud.
E. Top flange suppons upper guide structure E. Top flange suppons upper guide structure assembly. assembly. F. (6) amplitude limitidg devices (snubbers) F. (6) amplitude limiting devices (snubbers) l [ ' attached to lower barrel. attached to lower barrel. l 1 Lower Support Structure A A. Made up of interlocked grid beams with A. Made up of interlocked grid beams with jQ surrounding short cylinder and perforated bottom plates attached to the bottoms of surrounding shon cylinder and perforated bottom plates attached to the bottom of the the beams. beams B. Fuel suppon pins are attached to the top B. Fuel support pins are attached to the top end of the grid beams, end of the grid beams. C. The core shroud assembly is attached to C. The core shroud assembly is attached to the top of the LSS cylinder, the top of the LSS cylinder. l D. The LSS cylinder rests on and is attached D. The LSS cylinder rests on and is attached to the CSB bottom (intemal) flange. to the CSB bottom (internal) flange. Upper Guide Stiveture A. Right circular cylinder supponed by a A. Right circular cylinder supponed by a heavy external flange type end, and heavy heavy external flange top end, and heavy plate attached to bottom end. plate attached to bottom end. 4 B .~ Heavy plate in "A" is perforated with flow B. Heavy plate in "A" is perforated with flow holes and guide tubes. holes and guide tubes.
=
C. A second heavy plate suppons the bottom C. A second heavy plate supports the bottom , ends of the guide tubes, is also perforated. ends of the guide tubes, is also perforated. j D. This second heavy plate "C" engages guide D This second heavy plate "C" engages lugs on the core shroud. guide lugs on the core shroud. E. - The guide tubes are welded to the two E. The guide tubes are welded to the two heavy plates above. heavy plates above. i Anment Deekn Aseewief Deedpn et SSC _ Page 3.9181 b
System SO+ oesign control Document Table 3.9-17 Reactor Vessel Internals Arrangement Comparison (Cont'd.) Palo Verde System 80+ l 1 Control Element Assembly Shroud Assembly A. A series of large diameter tubes are A. A series of large diameter tubes are connected together by full length webs. connected together by full length webs. ! B. All welded constructed. B. All welded construction. C. Blank C. The shroud tube and web assembly is connected to an external cylinder. D. The tube and web assembly is supponed D. The tube, web, and cylinder assembly is by the UGS support plate via tie rods. supported by the UGS upper flange. Incorporates four interlocking snubbers to - the upper UGS. E. Material - Austenitic Stainless Steel E. Material - Austenitic Stainless Steel O O Approved Deelgrr Notenlet Desiers of SSC page 3,y.ggy
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- Table 3.9-18 Nominal Dimensional Comparison Reactor Pressure Vessel Internals Component Palo Verde System 80+
Core Support Barrel: Length in. 383-1/4 383-1/4 Diameter (ID) in. 157 157 Thickness Upper in. 3 3 Thickness Middle in. 2-5/8 2-5/8 Thickness Lower in. 3 3 ! Outlet Nozzles (Qty) 2 2 Diameter (ID) 46-5/8 46-5/8 f l l Snubbers (Qty) 6 6 Lower Support Structure: Cylinder Height in. 16-1/4 16-1/4 r 156-5/16 Cylinder Diameter (OD) in. 156-5/16 (v) Main Beams (Qty) 16 16 Beam Thickness in. 1-3/4 1-3/4 Beam Height in. 26-3/8 26-3/8 i UGS Support Barrel Assembly: Length in. 193-7/8 193-7/8 Diameter Flange (OD) in. 179-1/2 179-1/2 Diameter Barrel (OD) in. 156 156 Barrel Thickness in. 3 3 { i CEA Guide Tubes (Qty) 804 820 Fuel Alignment Plate Diameter in. 156 156 Plate Thickness in. 4-1/2 4-1/2 L) Attwoved Desirs Waterial- Desivt of SSC Page 3.9-183
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